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 |

    Energy Savers [EERE]

    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

    Office of Environmental Management (EM)

    American Recovery and Reinvestment Act workers achieved a significant milestone in the decommissioning of a Cold War reactor at the Sa- vannah River Site this month after they safely re- moved its rusty, orange, 75-foot-tall dome. With the help of a 660-ton crane and lifting lugs, the work- ers pulled the 174,000-pound dome off the Heavy Water Components Test Reactor, capping more than 16 months of preparations. Workers will cut the dome into smaller pieces for disposal. Removal of the dome

  4. 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...

  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. Heavy Water Components Test Reactor Decommissioning - Major Component Removal

    SciTech Connect (OSTI)

    Austin, W.; Brinkley, D.

    2010-05-05

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

  8. 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

  9. 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.

  10. Massive Hanford Test Reactor Removed- Plutonium Recycle Test Reactor removed from Hanford’s 300 Area

    Broader source: Energy.gov [DOE]

    RICHLAND, WA – 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.

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

    DOE Patents [OSTI]

    Hunsbedt, Anstein N. (Los Gatos, CA)

    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.

  12. 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 accountability, back up or fill in work, work packages, condensation control, radiological contamination control, and organization of the waste stream.

  13. 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.

  14. 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. [Korea Hydro and Nuclear Power Co., Ltd., 260 Naa-ri Yangnam-myeon Gyeongju-si, Gyeonasangbuk-do, 780-815 (Korea, Republic of)

    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)

  15. 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

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

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

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

  17. System for fuel rod removal from a reactor module

    DOE Patents [OSTI]

    Matchett, Richard L. (Bethel Park, PA); Roof, David R. (North Huntingdon, PA); Kikta, Thomas J. (Pittsburgh, PA); Wilczynski, Rosemarie (McKees Rocks, PA); Nilsen, Roy J. (Pittsburgh, PA); Bacvinskas, William S. (Bethel Park, PA); Fodor, George (Pittsburgh, PA)

    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.

  18. 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.

  19. 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

  20. 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.

  1. 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,...

  2. 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. In brief, the Path Forward was developed to reconsider potential open air demolition areas; characterize to determine if any zircaloy exists, evaluate existing concrete data to determine additional characterization needs, size the new building to accommodate human machine interface and tooling, consider bucket thumb and use ofshape-charges in design, and finally to utilize complex-wide and industry explosive demolition lessons learned in the design approach. Appendix B documents these results from the team's use ofValue Engineering process tools entitled Weighted Analysis Alternative Matrix, Matrix Conclusions, Evaluation Criteria, and Alternative Advantages and Disadvantages. These results were further supported with the team's validation of parking-lot information sheets: memories (potential ideas to consider), issues/concerns, and assumptions, contained in Appendix C. Appendix C also includes the recorded workshop flipchart notes taken from the SAR Alternatives and Project Overview presentations. The SAR workshop presentations, including a 3-D graphic illustration demonstration video have been retained in the CHPRC project file, and were not included in this report due to size limitations. The workshop concluded with a round robin close-out where each member was engaged for any last minute items and meeting utility. In summary, the team felt the session was value added and looked forward to proceeding with the recommended actions and conceptual design.

  3. 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.

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

    DOE Patents [OSTI]

    Cooper, Martin H. (Monroeville, PA)

    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.

  5. Packed-Bed Reactor Study of NETL Sample 196c for the Removal of Carbon

    Office of Scientific and Technical Information (OSTI)

    Dioxide from Simulated Flue Gas Mixture (Technical Report) | SciTech Connect Technical Report: Packed-Bed Reactor Study of NETL Sample 196c for the Removal of Carbon Dioxide from Simulated Flue Gas Mixture Citation Details In-Document Search Title: Packed-Bed Reactor Study of NETL Sample 196c for the Removal of Carbon Dioxide from Simulated Flue Gas Mixture An amine-based solid sorbent process to remove CO2 from flue gas has been investigated. The sorbent consists of polyethylenimine (PEI)

  6. 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.

  7. 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.

  8. 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...

  9. 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.

  10. 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.

  11. 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)

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

    DOE Patents [OSTI]

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

    1978-01-01

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

  13. 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 Conversion of the Budapest Research Reactor and Removal of All Fresh HEU in Hungary | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations

  14. 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.

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

    DOE Patents [OSTI]

    Frost, Albert C. (Congers, NY); Yang, Chang-Lee (Spring Valley, NY)

    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.

  16. 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

  17. 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...

  18. 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

  19. 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

  20. 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.

  1. 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.

  2. 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...

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

    DOE Patents [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.

  4. HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

    SciTech Connect (OSTI)

    Austin, W.; Brinkley, D.

    2011-10-13

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

  5. 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)

  6. Reactor pressure vessel vented head

    DOE Patents [OSTI]

    Sawabe, James K. (San Jose, CA)

    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.

  7. Energy Department Sells Historic Teapot Dome Oilfield | Department of

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

    Energy Sells Historic Teapot Dome Oilfield Energy Department Sells Historic Teapot Dome Oilfield January 30, 2015 - 11:00am Addthis News Media Contact 202 586 4940 DOENews@hq.doe.gov Energy Department Sells Historic Teapot Dome Oilfield Sale Ends Department Ownership of Naval Petroleum Reserves WASHINGTON - Today, the Energy Department finalized the sale of the historic Teapot Dome Oilfield located 35 miles north of Casper, Wyoming to Stranded Oil Resources Corporation, a subsidiary of

  8. 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.

  9. 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)

  10. 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.

  11. 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.

  12. 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:...

  13. 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.

  14. 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.

  15. 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/circulators would be kept low by maintaining a pressurized system coolant back pressure of {approx}7-8 bars through the design of the guard containment for such a design pressure. This approach is termed the medium pressure approach by both CEA and the US. Such a containment design pressure is in the range of the LWR experience, both PWRs and BWRs. Both metal containments and concrete guard containments are possible in this pressure range. This approach is then a time-at-risk approach as the power requirements should be low enough that battery/fuel cell banks without diesel generator start-up failure rate issues should be capable of providing the necessary power. Compressed gas sources are another possibility. A companion PRA study is being conducted to survey the reliability of such systems.

  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 the uranium deposit removal system and equipment. After removal a series of NDA measurements was performed to determine the amount of uranium material remaining in the ACB, the amount of uranium material removed from the ACB, and the amount of uranium material remaining in the uranium removal equipment due to removal activities.

  17. 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...

  18. 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.

  19. 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.

  20. 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.

  1. 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 information supporting job planning and understanding of the conditions was the data obtained from the gross gamma meter that was inserted into each casing to provide a rough estimate of dose rates in the tubes. No added value was realized in attempting to quantify the source term and/or associate the isotopic activity with a particular actual waste form (e.g., sludge). Implementing the WRM system allowed monitoring of worker and boundary exposure rates from a distance, maintaining compliance with ALARA principles. This system also provided the project team early knowledge of items being removed that had high exposure rates associated with them, thus creating an efficient method of acknowledging an issue and arriving at a solution prior to having an upset condition. An electronic dosimeter with telemetry capability replaced the excavator mounted AMP-100 system approximately half way through remediation of the silos. Much higher connectivity efficiency was derived from this configuration. Increasing the data feed efficiency additionally led to less interruption of the remediation effort. Early in system testing process a process handicap on the excavator operator was acknowledged. A loss of depth perception resulted when maneuvering the excavator and bucket using the camera feed to an in-cab monitor. Considerable practice and mock-up testing allowed this handicap to be overcome. The most significant equipment failures involved the cable connection to the camera mounted between the clamshell bucket jaws and the video splitter in the excavator cab. Rotation of the clamshell bucket was identified as the cause of cable connection failures because of the cyclic twisting motion and continuous mechanical jarring of the connection. In-cab vibration was identified as the culprit in causing connection failures of the video splitter. While these failures were repaired, substantial production time was lost. Ultimately, the decision was made to purchase a second cable and higher quality video splitter eliminate the down time. An engineering improvement for future operations would be i

  2. 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...

  3. 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

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

    Office of Legacy Management (LM)

    Tatum Salt Dome Test Site - MS 01 FUSRAP Considered Sites Site: Tatum Salt Dome Test Site (MS.01) Designated Name: Alternate Name: Location: Evaluation Year: Site Operations: Site Disposition: Radioactive Materials Handled: Primary Radioactive Materials Handled: Radiological Survey(s): Site Status: Also see Documents Related to Tatum Salt Dome Test Site

  5. 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.

  6. 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

  7. 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

  8. Carbon Sequestration in New Mexico's Bravo Dome | U.S. DOE Office...

    Office of Science (SC) Website

    Sequestration in New Mexico's Bravo Dome Basic Energy Sciences (BES) BES Home About ... provides new clues for the long-term viability of geological carbon dioxide storage. ...

  9. 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.

  10. material removal

    National Nuclear Security Administration (NNSA)

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

    Page...

  11. material removal

    National Nuclear Security Administration (NNSA)

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

    Pag...

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

    Office of Environmental Management (EM)

    Deep Dig Removes Contaminated Soil Hanford Deep Dig Removes Contaminated Soil March 11, 2013 - 12:00pm Addthis An aerial view of Hanford’s 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 contamination. An aerial view of Hanford's 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 contamination. Workers remove soil

  13. Brookhaven Lab Completes Decommissioning of Graphite Research Reactor:

    Energy Savers [EERE]

    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

  14. 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)

  15. 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.

  16. Actinide removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C. (Pleasanton, CA); von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Adamson, Martyn G. (Danville, CA)

    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.

  17. Metals removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C. (Pleasanton, CA); Von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Brummond, William A. (Livermore, CA); Adamson, Martyn G. (Danville, CA)

    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.

  18. 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.

  19. 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...

  20. 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...

  1. 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...

  2. 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.

  3. Nuclear reactor reflector

    DOE Patents [OSTI]

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

    1994-01-01

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

  4. 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.

  5. 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.``

  6. 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.

  7. Selective purge for hydrogenation reactor recycle loop

    DOE Patents [OSTI]

    Baker, Richard W. (Palo Alto, CA); Lokhandwala, Kaaeid A. (Union City, CA)

    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.

  8. 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.

  9. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.; Marasco, J.A.

    1996-02-27

    A fluidized bed reactor system is described which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves. 3 figs.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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 of a waste tank is expected to depend on two main parameters: carbonate is expected to slow olivine degradation rates, whereas hydroxide is expected to enhance olivine dissolution rates. Which of these two competing dissolution drivers will have a larger impact on the dissolution rate in the specific environment of a waste tank is currently not identifiable. In general, cancrinite is much smaller and less hard than either olivine or garnet, so would be expected to be less erosive to processing equipment. Complete degradation of either garnet or olivine prior to being processed at the Waste Treatment and Immobilization Plant cannot be confirmed, however.

  15. Liquid metal cooled nuclear reactor plant system

    DOE Patents [OSTI]

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

    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.

  16. 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.

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

    DOE Patents [OSTI]

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

    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.

  18. 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.

  19. 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.

  20. 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...

  1. 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."

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Liquid metal cooled nuclear reactors with passive cooling system

    DOE Patents [OSTI]

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

    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.

  7. 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.

  8. 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.

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

    DOE Patents [OSTI]

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

    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.

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

    DOE Patents [OSTI]

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

    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.

  11. Packed fluidized bed blanket for fusion reactor

    DOE Patents [OSTI]

    Chi, John W. H. (Mt. Lebanon, PA)

    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.

  12. 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

  13. 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 resulted in the delivery of radionuclidecontaminated brine to the surface, where a portion was diverted to culinary salt for direct ingestion by the existing population. Consequence analyses indicated calculated human doses that would be highly deleterious. Additional analyses indicated that doses well above background would occur from such a scenario t even if it occurred a million years into the future. The way to preclude such an intrusion is for continued control over the repository sitet either through direct institutional control or through the effective passive transfer of information. A secondary aspect of the specific human intrusion scenario involved a breach through the side of the salt dome t through which radionuclides migrated via the ground-water system to the accessible environment. This provided a demonstration of the geotransport methodology that AEGIS can use in actual site evaluations, as well as the WRIT program's capabilities with respect to defining the source term and retardation rates of the radionuclides in the repository. This reference site analysis was initially published as a Working Document in December 1979. That version was distributed for a formal peer review by individuals and organizations not involved in its development. The present report represents a revisiont based in part on the responses received from the external reviewers. Summaries of the comments from the reviewers and responses to these comments by the AEGIS staff are presented. The exercise of the AEGIS methodology was successful in demonstrating the methodologyt and thus t in providing a basis for substantive peer review, in terms of further development of the AEGIS site-applications capability and in terms of providing insight into the potential for consequential human intrusion into a salt dome repository.

  14. 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....

  15. Reactor core isolation cooling system

    DOE Patents [OSTI]

    Cooke, Franklin E. (San Jose, CA)

    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.

  16. 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.

  17. 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.

  18. Heat dissipating nuclear reactor with metal liner

    DOE Patents [OSTI]

    Gluekler, Emil L. (San Jose, CA); Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

    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.

  19. 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.

  20. 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.

  1. 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.

  2. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, James E. (Greensburg, PA); Meuschke, Robert E. (Pittsburgh, PA)

    1995-01-01

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul (Pittsburgh, PA)

    1981-01-01

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

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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. The reflector dip calculated from these images were used to further constrain the depth images from the less well sampled VSP data. In spite of the above technical success, we were less than pleased with our final VSP images. Since no extra sources are used, simultaneous recording of the surface and VSP data were believed to provide a cost-effective means to acquire 3-D VSP data. However, the subsurface sampling associated with the VSP is quite different from that associated with the surface seismic data. After our analysis, we find that considerable shot infill would result in a better, unaliased subsurface image. We were able to ascertain that the subsurface illuminated by the VSP was extremely small, with the PS image being even smaller than the PP image. One-way wave equation extrapolators do not work well for the VSP geometry, where we wish to extrapolate energy sideways (from the VSP well towards and away from the salt dome) as well as vertically (away from the shots on the earth surface). Merging separately-generated images proved to be both cumbersome and error-prone. Alternative, advanced multiarrival traveltime calculations that we obtained from research colleagues at other institutions could not be easily modified to image rays that had an upgoing component. In the end, we used a simpler first-arrival Eikonal-based traveltime algorithm with its well-known limitations. While the surface acquisition using radial receiver lines and concentric shot lines provided good illumination of the salt dome, this unconventional geometry proved to be particularly difficult to process using commercial software, where the lack of ''shot lines'' and ''receiver lines'' necessary for dip filtering, residual statics, and residual velocity analysis proved to be nearly intractable. We also learned that while commercial software available at UH works well for a bootstrapped velocity model computed from the seismic data alone, it was severely limited in its ability to include the dense well control available at Vinton Dome. To more accurately estimate velocities, we develope

  13. 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.

  14. 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.

  15. 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.

  16. Phenol removal pretreatment process

    DOE Patents [OSTI]

    Hames, Bonnie R. (Westminster, CO)

    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.

  17. 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.

  18. 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.

  19. Neutronic reactor

    DOE Patents [OSTI]

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

    1983-01-01

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

  20. 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.

  1. 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.

  2. 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.

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

    DOE Patents [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.

  4. Graphitic packing removal tool

    DOE Patents [OSTI]

    Meyers, K.E.; Kolsun, G.J.

    1997-11-11

    Graphitic packing removal tools for removal of the seal rings in one piece are disclosed. The 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. 5 figs.

  5. Graphitic packing removal tool

    DOE Patents [OSTI]

    Meyers, Kurt Edward (Avella, PA); Kolsun, George J. (Pittsburgh, PA)

    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.

  6. EIS-0021: Strategic Petroleum Reserve, Seaway Group Salt Domes, Brazoria County, Texas (also see EIS-0075-S and EIS-0029)

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy's Strategic Petroleum Reserve Office developed this statement to analyze the environmental impacts which would occur during site preparation and operation of oil storage facilities at each of five proposed candidate sites in the Seaway Group of salt domes.

  7. 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.

    1996-06-01

    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 U.S. 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.

  8. Characterizing the Weeks Island Salt Dome drilling of and seismic measurements from boreholes

    SciTech Connect (OSTI)

    Sattler, A.R.; Harding, R.S.; Jacobson, R.D.; Finger, J.T.; Keefe, R.; Neal, J.T.

    1996-10-01

    A sinkhole 36 ft across, 30 ft deep was first observed in the alluvium over the Weeks Island Salt Dome (salt mine converted for oil storage by US Strategic Petroleum Reserve) May 1992. Four vertical, two slanted boreholes were drilled for diagnostics. Crosswell seismic data were generated; the velocity images suggest that the sinkhole collapse is complicated, not a simple vertical structure. The coring operation was moderately difficult; limited core was obtained through the alluvium, and the quality of the salt core from the first two vertical wells was poor. Core quality improved with better bit selection, mud, and drilling method. The drilling fluid program provided fairly stable holes allowing open hole logs to be run. All holes were cemented successfully (although it took 3 attempts in one case).

  9. PHOTOMETRY OF VARIABLE STARS FROM DOME A, ANTARCTICA: RESULTS FROM THE 2010 OBSERVING SEASON

    SciTech Connect (OSTI)

    Wang, Lingzhi; Zhu, Zonghong; Macri, Lucas M.; Wang, Lifan; Ashley, Michael C. B.; Lawrence, Jon S.; Luong-Van, Daniel; Storey, John W. V.; Cui, Xiangqun; Feng, Long-Long; Gong, Xuefei; Liu, Qiang; Shang, Zhaohui; Yang, Huigen; Yang, Ji; Yuan, Xiangyan; Zhou, Xu; Zhu, Zhenxi; Pennypacker, Carl R.; York, Donald G.

    2013-12-01

    We present results from a season of observations with the Chinese Small Telescope ARray, obtained over 183 days of the 2010 Antarctic winter. We carried out high-cadence time-series aperture photometry of 9125 stars with i ?< 15.3 mag located in a 23 deg{sup 2} region centered on the south celestial pole. We identified 188 variable stars, including 67 new objects relative to our 2008 observations, thanks to broader synoptic coverage, a deeper magnitude limit, and a larger field of view. We used the photometric data set to derive site statistics from Dome A. Based on two years of observations, we find that extinction due to clouds at this site is less than 0.1 and 0.4 mag during 45% and 75% of the dark time, respectively.

  10. Dome – like variation of the superconducting gap anisotropy in Fe-based superconductors

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

    Prozorov, R.; Cho, K.; Kim, H.; Tanatar, M. A.

    2013-07-17

    Experiments performed on different iron-based superconductors suggest a variety of possible structures of the superconducting energy gap, both nodeless and nodal. To understand the pairing mechanisms, it is important to identify common features in the behavior of different materials. Measurements of the temperature - dependent London penetration depth provide important information on the structure of the superconducting gap. We show that despite significant differences between different iron - based superconductors, there is a universal trend: the gap is least anisotropic at the optimal doping and its anisotropy increases upon the departure towards underdoped and overdoped ends of the ''superconducting dome''.more » As a result, this trend is not related to the presence of the long-range magnetic order in the underdoped state.« less

  11. 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).

  12. Recovery Act Workers Clear Reactor Shields from Brookhaven Lab

    Broader source: Energy.gov [DOE]

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

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Recovery Act Workers Clear Reactor Shields from Brookhaven Lab

    Office of Environmental Management (EM)

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

  18. Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites |

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

    Department of Energy Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites In January 2013, the Department of Energy issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. 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

  19. 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.

  20. Device for removing blackheads

    DOE Patents [OSTI]

    Berkovich, Tamara (116 N. Wetherly Dr., Suite 115, Los Angeles, CA)

    1995-03-07

    A device for removing blackheads from pores in the skin having a elongated handle with a spoon shaped portion mounted on one end thereof, the spoon having multiple small holes piercing therethrough. Also covered is method for using the device to remove blackheads.

  1. 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.

  2. Chris Bergren Director, Environment Compliance & Area Completion Projects

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

    Deactivation & Decommissioning at SRS Chris Bergren Director, Environment Compliance & Area Completion Projects DOE Office of Environmental Management Robotics Team Visit to SRS Tuesday, December 8, 2015 Tony Long Acting Manager, Area Completion Projects T Area Completion Area Completions Then Now M Area Completion Now Then Now 2 In Situ Decommissioning of the Heavy Water Components Test Reactor (HWCTR) Before Reactor Dome Removal Reactor Vessel Removal Demolition of Dome After 3 K-Area

  3. 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.

  4. Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

    SciTech Connect (OSTI)

    A. B. Culp

    2007-01-26

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

  5. Passive shut-down heat removal system

    DOE Patents [OSTI]

    Hundal, Rolv (Greensburg, PA); Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

    1988-01-01

    An improved shut-down heat removal system for a liquid metal nuclear reactor of the type having a vessel for holding hot and cold pools of liquid sodium is disclosed herein. Generally, the improved system comprises a redan or barrier within the reactor vessel which allows an auxiliary heat exchanger to become immersed in liquid sodium from the hot pool whenever the reactor pump fails to generate a metal-circulating pressure differential between the hot and cold pools of sodium. This redan also defines an alternative circulation path between the hot and cold pools of sodium in order to equilibrate the distribution of the decay heat from the reactor core. The invention may take the form of a redan or barrier that circumscribes the inner wall of the reactor vessel, thereby defining an annular space therebetween. In this embodiment, the bottom of the annular space communicates with the cold pool of sodium, and the auxiliary heat exchanger is placed in this annular space just above the drawn-down level that the liquid sodium assumes during normal operating conditions. Alternatively, the redan of the invention may include a pair of vertically oriented, concentrically disposed standpipes having a piston member disposed between them that operates somewhat like a pressure-sensitive valve. In both embodiments, the cessation of the pressure differential that is normally created by the reactor pump causes the auxiliary heat exchanger to be immersed in liquid sodium from the hot pool. Additionally, the redan in both embodiments forms a circulation flow path between the hot and cold pools so that the decay heat from the nuclear core is uniformly distributed within the vessel.

  6. Ceramic membrane reactor with two reactant gases at different pressures

    DOE Patents [OSTI]

    Balachandran, Uthamalingam (Hinsdale, IL); Mieville, Rodney L. (Glen Ellyn, IL)

    2001-01-01

    The invention is a ceramic membrane reactor for syngas production having a reaction chamber, an inlet in the reactor for natural gas intake, a plurality of oxygen permeating ceramic slabs inside the reaction chamber with each slab having a plurality of passages paralleling the gas flow for transporting air through the reaction chamber, a manifold affixed to one end of the reaction chamber for intake of air connected to the slabs, a second manifold affixed to the reactor for removing the oxygen depleted air, and an outlet in the reaction chamber for removing syngas.

  7. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

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

    2011-03-01

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

  8. EIS-0024: Strategic Petroleum Reserve, Capline Group Salt Domes, Iberia, Napoleonville, Weeks Island Expansion, Bayou Choctaw Expansion, Chacahoula- Iberia, Iberville, and Lafourche Parishes, Louisiana

    Broader source: Energy.gov [DOE]

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

  9. RIVERTON DOME GAS EXPLORATION AND STIMULATION TECHNOLOGY DEMONSTRATION, WIND RIVER BASIN, WYOMING

    SciTech Connect (OSTI)

    Ronald C. Surdam; Zunsheng Jiao; Nicholas K. Boyd

    1999-11-01

    The new exploration technology for basin center gas accumulations developed by R.C. Surdam and Associates at the Institute for Energy Research, University of Wyoming, was applied to the Riverton Dome 3-D seismic area. Application of the technology resulted in the development of important new exploration leads in the Frontier, Muddy, and Nugget formations. The new leads are adjacent to a major north-south trending fault, which is downdip from the crest of the major structure in the area. In a blind test, the drilling results from six new Muddy test wells were accurately predicted. The initial production values, IP, for the six test wells ranged from < one mmcf/day to four mmcf/day. The three wells with the highest IP values (i.e., three to four mmcf/day) were drilled into an intense velocity anomaly (i.e., anomalously slow velocities). The well drilled at the end of the velocity anomaly had an IP value of one mmcf/day, and the two wells drilled outside of the velocity anomaly had IP values of < one mmcf/day and are presently shut in. Based on these test results, it is concluded that the new IER exploration strategy for detecting and delineating commercial, anomalously pressured gas accumulation is valid in the southwestern portions of the Wind River Basin, and can be utilized to significantly reduce exploration risk and to increase profitability of so-called basin center gas accumulations.

  10. Discovery of multiple pulsations in the new ? Scuti star HD 92277: Asteroseismology from Dome A, Antarctica

    SciTech Connect (OSTI)

    Zong, Weikai; Fu, Jian-Ning; Niu, Jia-Shu; Zhu, Zonghong; Charpinet, S.; Vauclair, G.; Ashley, Michael C. B.; Lawrence, Jon S.; Luong-Van, Daniel; Cui, Xiangqun; Gong, Xuefei; Feng, Longlong; Wang, Lifan; Yuan, Xiangyan; Zhu, Zhenxi; Liu, Qiang; Wang, Lingzhi; Zhou, Xu; Pennypacker, Carl R.; York, Donald G.

    2015-02-01

    We report the discovery of low-amplitude oscillations in the star HD 92277 from long, continuous observations in the r and g bands using the CSTAR telescopes in Antarctica. A total of more than 1950 hours of high-quality light curves were used to categorize HD 92277 as a new member of the ? Scuti class. We have detected 21 (20 frequencies are independent and one is the linear combination) and 14 (13 frequencies are independent and one is the linear combination) pulsation frequencies in the r and g bands, respectively, indicating a multi-periodic pulsation behavior. The primary frequency f{sub 1} = 10.810 days{sup ?1} corresponds to a period of 0.0925 days and is an l = 1 mode. We estimate a B ? V index of 0.39 and derive an effective temperature of 6800 K for HD 92277. We conclude that long, continuous and uninterrupted time-series photometry can be performed from Dome A, Antarctica, and that this is especially valuable for asteroseismology where multi-color observations (often not available from space-based telescopes) assist with mode identification.

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

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

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

  12. Small Modular Reactors - SRSCRO

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

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

  13. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Stewart, H.B.

    1958-12-23

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

  14. Continuous sulfur removal process

    DOE Patents [OSTI]

    Jalan, V.; Ryu, J.

    1994-04-26

    A continuous process for the removal of hydrogen sulfide from a gas stream using a membrane comprising a metal oxide deposited on a porous support is disclosed. 4 figures.

  15. Vanadium recycling for fusion reactors

    SciTech Connect (OSTI)

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

    1994-04-01

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

  16. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect (OSTI)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  17. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21

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

  18. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11

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

  19. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16

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

  20. Arsenic removal from water

    DOE Patents [OSTI]

    Moore, Robert C.; Anderson, D. Richard

    2007-07-24

    Methods for removing arsenic from water by addition of inexpensive and commonly available magnesium oxide, magnesium hydroxide, calcium oxide, or calcium hydroxide to the water. The hydroxide has a strong chemical affinity for arsenic and rapidly adsorbs arsenic, even in the presence of carbonate in the water. Simple and commercially available mechanical methods for removal of magnesium hydroxide particles with adsorbed arsenic from drinking water can be used, including filtration, dissolved air flotation, vortex separation, or centrifugal separation. A method for continuous removal of arsenic from water is provided. Also provided is a method for concentrating arsenic in a water sample to facilitate quantification of arsenic, by means of magnesium or calcium hydroxide adsorption.

  1. Drum lid removal tool

    SciTech Connect (OSTI)

    Pella, Bernard M.; Smith, Philip D.

    2010-08-24

    A tool for removing the lid of a metal drum wherein the lid is clamped over the drum rim without protruding edges, the tool having an elongated handle with a blade carried by an angularly positioned holder affixed to the midsection of the handle, the blade being of selected width to slice between lid lip and the drum rim and, when the blade is so positioned, upward motion of the blade handle will cause the blade to pry the lip from the rim and allow the lid to be removed.

  2. Removable feedwater sparger assembly

    DOE Patents [OSTI]

    Challberg, Roy C. (Livermore, CA)

    1994-01-01

    A removable feedwater sparger assembly includes a sparger having an inlet pipe disposed in flow communication with the outlet end of a supply pipe. A tubular coupling includes an annular band fixedly joined to the sparger inlet pipe and a plurality of fingers extending from the band which are removably joined to a retention flange extending from the supply pipe for maintaining the sparger inlet pipe in flow communication with the supply pipe. The fingers are elastically deflectable for allowing engagement of the sparger inlet pipe with the supply pipe and for disengagement therewith.

  3. Removable feedwater sparger assembly

    DOE Patents [OSTI]

    Challberg, R.C.

    1994-10-04

    A removable feedwater sparger assembly includes a sparger having an inlet pipe disposed in flow communication with the outlet end of a supply pipe. A tubular coupling includes an annular band fixedly joined to the sparger inlet pipe and a plurality of fingers extending from the band which are removably joined to a retention flange extending from the supply pipe for maintaining the sparger inlet pipe in flow communication with the supply pipe. The fingers are elastically deflectable for allowing engagement of the sparger inlet pipe with the supply pipe and for disengagement therewith. 8 figs.

  4. Condensate removal device

    DOE Patents [OSTI]

    Maddox, James W.; Berger, David D.

    1984-01-01

    A condensate removal device is disclosed which incorporates a strainer in unit with an orifice. The strainer is cylindrical with its longitudinal axis transverse to that of the vapor conduit in which it is mounted. The orifice is positioned inside the strainer proximate the end which is remoter from the vapor conduit.

  5. Methods and apparatus for carbon dioxide removal from a fluid stream

    DOE Patents [OSTI]

    Wei, Wei (Mission Viejo, CA); Ruud, James Anthony (Delmar, NY); Ku, Anthony Yu-Chung (Rexford, NY); Ramaswamy, Vidya (Niskayuna, NY); Liu, Ke (Rancho Santa Margrita, CA)

    2010-01-19

    An apparatus for producing hydrogen gas wherein the apparatus includes a reactor. In one embodiment, the reactor includes at least two conversion-removal portions. Each conversion-removal portion comprises a catalyst section configured to convert CO in the stream to CO.sub.2 and a membrane section located downstream of and in flow communication with the catalyst section. The membrane section is configured to selectively remove the CO.sub.2 from the stream and to be in flow communication with a sweep gas.

  6. Risk Removal | Department of Energy

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

    Risk Removal Risk Removal Workers safely remove old mercury tanks from the Y-12 National Security Complex. Workers safely remove old mercury tanks from the Y-12 National Security Complex. Risk removal is the most crucial and pivotal action for EM to achieve its missions locally. The organization works to protect the environment and residents' and employees' health, provide clean land for future generations, and bolster DOE missions, modernization, and economic development in Oak Ridge. All of

  7. Method and apparatus for enhancing reactor air-cooling system performance

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA)

    1996-01-01

    An enhanced decay heat removal system for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer.

  8. Method and apparatus for enhancing reactor air-cooling system performance

    DOE Patents [OSTI]

    Hunsbedt, A.

    1996-03-12

    An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

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

    DOE Patents [OSTI]

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

    1997-01-01

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

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

    DOE Patents [OSTI]

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

    1997-12-16

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

  11. Nuclear reactor fuel rod attachment system

    DOE Patents [OSTI]

    Not Available

    1980-09-17

    A reusable system is described for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member. The locking cap has two opposing fingers shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed.

  12. Alternative Passive Decay-Heat Systems for the Advanced High-Temperature Reactor

    SciTech Connect (OSTI)

    Forsberg, Charles W.

    2006-07-01

    The Advanced High-Temperature Reactor (AHTR) is a low-pressure, liquid-salt-cooled high-temperature reactor for the production of electricity and hydrogen. The high-temperature (950 deg C) variant is defined as the liquid-salt-cooled very high-temperature reactor (LS-VHTR). The AHTR has the same safety goals and uses the same graphite-matrix coated particle fuel as do modular high-temperature gas-cooled reactors. However, the large AHTR power output [2400 to 4000 MW(t)] implies the need for a different type of passive decay-heat removal system. Because the AHTR is a low-pressure, liquid-cooled reactor like sodium-cooled reactors, similar types of decay-heat-removal systems can be used. Three classes of passive decay heat removal systems have been identified: the reactor vessel auxiliary cooling system which is similar to that proposed for the General Electric S-PRISM sodium-cooled fast reactor; the direct reactor auxiliary cooling system, which is similar to that used in the Experimental Breeder Reactor-II; and a new pool reactor auxiliary cooling system. These options are described and compared. (author)

  13. Design options for a bunsen reactor.

    SciTech Connect (OSTI)

    Moore, Robert Charles

    2013-10-01

    This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

  14. Pneumatic soil removal tool

    DOE Patents [OSTI]

    Neuhaus, John E. (Newport News, VA)

    1992-01-01

    A soil removal tool is provided for removing radioactive soil, rock and other debris from the bottom of an excavation, while permitting the operator to be located outside of a containment for that excavation. The tool includes a fixed jaw, secured to one end of an elongate pipe, which cooperates with a movable jaw pivotably mounted on the pipe. Movement of the movable jaw is controlled by a pneumatic cylinder mounted on the pipe. The actuator rod of the pneumatic cylinder is connected to a collar which is slidably mounted on the pipe and forms part of the pivotable mounting assembly for the movable jaw. Air is supplied to the pneumatic cylinder through a handle connected to the pipe, under the control of an actuator valve mounted on the handle, to provide movement of the movable jaw.

  15. Pneumatic soil removal tool

    DOE Patents [OSTI]

    Neuhaus, J.E.

    1992-10-13

    A soil removal tool is provided for removing radioactive soil, rock and other debris from the bottom of an excavation, while permitting the operator to be located outside of a containment for that excavation. The tool includes a fixed jaw, secured to one end of an elongate pipe, which cooperates with a movable jaw pivotably mounted on the pipe. Movement of the movable jaw is controlled by a pneumatic cylinder mounted on the pipe. The actuator rod of the pneumatic cylinder is connected to a collar which is slidably mounted on the pipe and forms part of the pivotable mounting assembly for the movable jaw. Air is supplied to the pneumatic cylinder through a handle connected to the pipe, under the control of an actuator valve mounted on the handle, to provide movement of the movable jaw. 3 figs.

  16. Coal hydrogenation and deashing in ebullated bed catalytic reactor

    DOE Patents [OSTI]

    Huibers, Derk T. A. (Pennington, NJ); Johanson, Edwin S. (Princeton, NJ)

    1983-01-01

    An improved process for hydrogenation of coal containing ash with agglomeration and removal of ash from an ebullated bed catalytic reactor to produce deashed hydrocarbon liquid and gas products. In the process, a flowable coal-oil slurry is reacted with hydrogen in an ebullated catalyst bed reaction zone at elevated temperature and pressure conditions. The upward velocity and viscosity of the reactor liquid are controlled so that a substantial portion of the ash released from the coal is agglomerated to form larger particles in the upper portion of the reactor above the catalyst bed, from which the agglomerated ash is separately withdrawn along with adhering reaction zone liquid. The resulting hydrogenated hydrocarbon effluent material product is phase separated to remove vapor fractions, after which any ash remaining in the liquid fraction can be removed to produce substantially ash-free coal-derived liquid products.

  17. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-01-01

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

  18. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-09-28

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

  19. B Reactor - Hanford Site

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

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

  20. Results from shallow research drilling at Inyo Domes, Long Valley Caldera, California and Salton Sea geothermal field, Salton Trough, California

    SciTech Connect (OSTI)

    Younker, L.W.; Eichelberger, J.C.; Kasameyer, P.W.; Newmark, R.L.; Vogel, T.A.

    1987-09-01

    This report reviews the results from two shallow drilling programs recently completed as part of the United States Department of Energy Continental Scientific Drilling Program. The purpose is to provide a broad overview of the objectives and results of the projects, and to analyze these results in the context of the promise and potential of research drilling in crustal thermal regimes. The Inyo Domes drilling project has involved drilling 4 shallow research holes into the 600-year-old Inyo Domes chain, the youngest rhyolitic event in the coterminous United States and the youngest volcanic event in Long Valley Caldera, California. The purpose of the drilling at Inyo was to understand the thermal, chemical and mechanical behavior of silicic magma as it intrudes the upper crust. This behavior, which involves the response of magma to decompression and cooling, is closely related to both eruptive phenomena and the establishment of hydrothermal circulation. The Salton Sea shallow research drilling project involved drilling 19 shallow research holes into the Salton Sea geothermal field, California. The purpose of this drilling was to bound the thermal anomaly, constrain hydrothermal flow pathways, and assess the thermal budget of the field. Constraints on the thermal budget links the local hydrothermal system to the general processes of crustal rifting in the Salton Trough.

  1. Period meter for reactors

    DOE Patents [OSTI]

    Rusch, Gordon K.

    1976-01-06

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

  2. Metal Cutting for Large Component Removal

    SciTech Connect (OSTI)

    Hulick, Robert M.

    2008-01-15

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

  3. Dismantlement of the TSF-SNAP Reactor Assembly

    SciTech Connect (OSTI)

    Peretz, Fred J

    2009-01-01

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

  4. Process for treating effluent from a supercritical water oxidation reactor

    DOE Patents [OSTI]

    Barnes, Charles M. (Idaho Falls, ID); Shapiro, Carolyn (Idaho Falls, ID)

    1997-01-01

    A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor.

  5. Process for treating effluent from a supercritical water oxidation reactor

    DOE Patents [OSTI]

    Barnes, C.M.; Shapiro, C.

    1997-11-25

    A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor. 6 figs.

  6. Improved vortex reactor system

    DOE Patents [OSTI]

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

    1995-01-01

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

  7. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28

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

  8. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-03-02

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

  9. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-01-01

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

  10. Electrochemically assisted paint removal

    SciTech Connect (OSTI)

    Keller, R.; Hydock, D.M.; Burleigh, T.D.

    1995-12-31

    A method to remove paint coatings from metal and other electronically conductive substrates is being studied. In particular, the remediation of objects coated with lead based paints is the focus of research. The approach also works very well with automotive coatings and may be competitive with sandblasting. To achieve debonding of the coating, the deteriorated or artifically damaged surface of the object is cathodically polarized. The object can be immersed in a benign aqueous electrolyte for treatment, or the electrolyte can be retained in an absorbent pad covering the surface to be treated.

  11. Reactor vessel support system

    DOE Patents [OSTI]

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

    1982-01-01

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

  12. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOE Patents [OSTI]

    Hunsbedt, A.; Boardman, C.E.

    1995-04-11

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

  13. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOE Patents [OSTI]

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

    1995-01-01

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

  14. Hanford Site Inspections Give Four Cocooned Reactors Clean Bill of Health

    Broader source: Energy.gov [DOE]

    RICHLAND, Wash. – Workers at the Hanford Site took a step back in time as they removed welds from the doors and entered four of the site’s cocooned reactors for required inspections.

  15. Spinning fluids reactor

    DOE Patents [OSTI]

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

    2012-11-20

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

  16. Reactor water cleanup system

    DOE Patents [OSTI]

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

    1994-12-20

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

  17. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

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

  18. Fuel rod retention device for a nuclear reactor

    DOE Patents [OSTI]

    Hylton, Charles L. (Madison Heights, VA)

    1984-01-01

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

  19. Sodium removal process development for LMFBR fuel subassemblies

    SciTech Connect (OSTI)

    Simmons, C.R.; Taylor, G.R.

    1981-10-01

    Two 37-pin scale models of Clinch River Breeder Reactor Plant fuel subassemblies were designed, fabricated and used at Westinghouse Advanced Reactors Division in the development and proof-testing of a rapid water-based sodium removal process for the ORNL Hot Experimental Facility, Liquid Metal Fast Breeder Reactor Fuel Reprocessing Cycle. Through a series of development tests on one of the models, including five (5) sodium wettings and three (3) high temperature sodium removal operations, optimum process parameters for a rapid water vapor-argon-water rinse process were identified and successfully proof-tested on a second model containing argon-pressurized, sodium-corroded model fuel pins simulating the gas plenum and cladding conditions expected for spent fuel pins in full scale subassemblies. Based on extrapolations of model proof test data, preliminary process parameters for a water vapor-nitrogen-water rinse process were calculated and recommended for use in processing full scale fuel subassemblies in the Sodium Removal Facility of the Fuel Receiving Cell, ORNL HEF.

  20. Shielded fluid stream injector for particle bed reactor

    DOE Patents [OSTI]

    Notestein, John E. (Morgantown, WV)

    1993-01-01

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

  1. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    SciTech Connect (OSTI)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phnix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  2. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    SciTech Connect (OSTI)

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma; Al Rashdan, Ahmad; Tsvetkov, Pavel Valeryevich; Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  3. Recovery of hydrogen from impurities using a palladium membrane reactor

    SciTech Connect (OSTI)

    Willms, R.S. [Los Alamos National Lab., NM (United States); Okuno, K. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)

    1993-12-01

    One of the important steps in processing the exhaust from a fusion reactor is recovering tritium which is incorporated into molecules such as water and methane. One device which may prove to be very effective for this purpose is a palladium membrane reactor. This is a reactor which incorporates a Pd/Ag membrane in the reactor geometry. Reactions such as water gas shift, steam reforming and methane cracking can be carried out over the reactor catalyst, and the product hydrogen can be simultaneously removed from the reacting mixture. Because product is removed, greater than usual conversions can be obtained. In addition ultrapure hydrogen is produced, eliminating the need for an additional processing step. A palladium membrane reactor has been built and tested with three different catalysts. Initial results with a Ni-based catalyst show that it is very effective at promoting all three reactions listed above. Under the proper conditions, hydrogen recoveries approaching 100% have been observed. This study serves to experimentally validate the palladium membrane reactor as potentially important tool for fusion fuel processing.

  4. The behavior of fission products during nuclear rocket reactor tests

    SciTech Connect (OSTI)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    The experience base regarding fission product behavior developed during the Rover program, the nuclear rocket development program of 1955--1972, will be useful in planning a renewed nuclear rocket program. During the Rover program, 20 reactors were tested at the Nuclear Rocket Development Station in Nevada. Nineteen of these discharged effluent directly into the atmosphere; the last reactor tested, a non-flight-prototypic, fuel-element-testing reactor called the Nuclear Furnace (NF-1) was connected to an effluent cleanup system that removed fission products before the hydrogen coolant (propellant) was discharged to the atmosphere. In general, we are able to increase both test duration and fuel temperature during the test series. Therefore fission product data from the later part of the program are more interesting and more applicable to future reactors. We have collected fission product retention (and release) data reported in both formal and informal publications for six of the later reactor tests; five of these were Los Alamos reactors that were firsts of a kind in configuration or operating conditions. We have also, with the cooperation of Westinghouse, included fission product data from the NRX-A6 reactor, the final member of series of developmental reactors with the same basic geometry, but with significant design and fabrication improvements as the series continued. Table 1 lists the six selected reactors and the test parameters for each.

  5. IMPROVED PROCESSES TO REMOVE NAPHTHENIC ACIDS

    SciTech Connect (OSTI)

    Aihua Zhang; Qisheng Ma; William A. Goddard; Yongchun Tang

    2004-04-28

    In the first year of this project, we have established our experimental and theoretical methodologies for studies of the catalytic decarboxylation process. We have developed both glass and stainless steel micro batch type reactors for the fast screening of various catalysts with reaction substrates of model carboxylic acid compounds and crude oil samples. We also developed novel product analysis methods such as GC analyses for organic acids and gaseous products; and TAN measurements for crude oil. Our research revealed the effectiveness of several solid catalysts such as NA-Cat-1 and NA-Cat-2 for the catalytic decarboxylation of model compounds; and NA-Cat-5{approx}NA-Cat-9 for the acid removal from crude oil. Our theoretical calculations propose a three-step concerted oxidative decarboxylation mechanism for the NA-Cat-1 catalyst.

  6. Improved Fischer-Tropsch Slurry Reactors

    SciTech Connect (OSTI)

    Andrew Lucero

    2009-03-20

    The conversion of synthesis gas to hydrocarbons or alcohols involves highly exothermic reactions. Temperature control is a critical issue in these reactors for a number of reasons. Runaway reactions can be a serious safety issue, even raising the possibility of an explosion. Catalyst deactivation rates tend to increase with temperature, particularly of there are hot spots in the reactor. For alcohol synthesis, temperature control is essential because it has a large effect on the selectivity of the catalysts toward desired products. For example, for molybdenum disulfide catalysts unwanted side products such as methane, ethane, and propane are produced in much greater quantities if the temperature increases outside an ideal range. Slurry reactors are widely regarded as an efficient design for these reactions. In a slurry reactor a solid catalyst is suspended in an inert hydrocarbon liquid, synthesis gas is sparged into the bottom of the reactor, un-reacted synthesis gas and light boiling range products are removed as a gas stream, and heavy boiling range products are removed as a liquid stream. This configuration has several positive effects for synthesis gas reactions including: essentially isothermal operation, small catalyst particles to reduce heat and mass transfer effects, capability to remove heat rapidly through liquid vaporization, and improved flexibility on catalyst design through physical mixtures in addition to use of compositions that cannot be pelletized. Disadvantages include additional mass transfer resistance, potential for significant back-mixing on both the liquid and gas phases, and bubble coalescence. In 2001 a multiyear project was proposed to develop improved FT slurry reactors. The planned focus of the work was to improve the reactors by improving mass transfer while considering heat transfer issues. During the first year of the project the work was started and several concepts were developed to prepare for bench-scale testing. PowerEnerCat was unable to raise their cash contribution for the project, and the work was stopped. This report summarizes some of the progress of the project and the concepts that were intended for experimental tests.

  7. Sodium Reactor Experiment decommissioning. Final report

    SciTech Connect (OSTI)

    Carroll, J.W.; Conners, C.C.; Harris, J.M.; Marzec, J.M.; Ureda, B.F.

    1983-08-15

    The Sodium Reactor Experiment (SRE) located at the Rockwell International Field Laboratories northwest of Los Angeles was developed to demonstrate a sodium-cooled, graphite-moderated reactor for civilian use. The reactor reached full power in May 1958 and provided 37 GWh to the Southern California Edison Company grid before it was shut down in 1967. Decommissioning of the SRE began in 1974 with the objective of removing all significant radioactivity from the site and releasing the facility for unrestricted use. Planning documentation was prepared to describe in detail the equipment and techniques development and the decommissioning work scope. A plasma-arc manipulator was developed for remotely dissecting the highly radioactive reactor vessels. Other important developments included techniques for using explosives to cut reactor vessel internal piping, clamps, and brackets; decontaminating porous concrete surfaces; and disposing of massive equipment and structures. The documentation defined the decommissioning in an SRE dismantling plan, in activity requirements for elements of the decommissioning work scope, and in detailed procedures for each major task.

  8. MEANS FOR SHIELDING AND COOLING REACTORS

    DOE Patents [OSTI]

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

    1959-02-10

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

  9. Control Rod Malfunction at the NRAD Reactor

    SciTech Connect (OSTI)

    Thomas L. Maddock

    2010-05-01

    The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

  10. Rubber stopper remover

    DOE Patents [OSTI]

    Stitt, Robert R. (Arvada, CO)

    1994-01-01

    A device for removing a rubber stopper from a test tube is mountable to an upright wall, has a generally horizontal splash guard, and a lower plate spaced parallel to and below the splash guard. A slot in the lower plate has spaced-apart opposing edges that converge towards each other from the plate outer edge to a narrowed portion, the opposing edges shaped to make engagement between the bottom of the stopper flange and the top edge of the test tube to wedge therebetween and to grasp the stopper in the slot narrowed portion to hold the stopper as the test tube is manipulated downwardly and pulled from the stopper. The opposing edges extend inwardly to adjoin an opening having a diameter significantly larger than that of the stopper flange.

  11. Gas-cooled nuclear reactor

    DOE Patents [OSTI]

    Peinado, Charles O. (La Jolla, CA); Koutz, Stanley L. (San Diego, CA)

    1985-01-01

    A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

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

    DOE Patents [OSTI]

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

    1993-01-01

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

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

    SciTech Connect (OSTI)

    Cooke, Conrad; Spann, Holger

    2013-07-01

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

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

    SciTech Connect (OSTI)

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

    2011-07-01

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

  15. Improved vortex reactor system

    DOE Patents [OSTI]

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

    1995-05-09

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

  16. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani (Karhula, FI)

    1996-01-01

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

  17. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19

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

  18. H Reactor - Hanford Site

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

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

  19. Tokamak reactor first wall

    DOE Patents [OSTI]

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

    1984-11-20

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

  20. Protection #1: Remove the Source

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

    Remove the Source Protection #1: Remove the Source The 3 Protections = Defense in Depth August 1, 2013 Waste being removed from MDA-B inside a metal building Excavation of waste from MDA-B thumbnail of Removing the source means excavating contaminants, sorting these by waste type, and transporting to a disposal area in which contaminants are contained. RELATED IMAGES http://farm8.staticflickr.com/7388/9571274521_679fe1e34a_t.jpg Enlarge http://farm4.staticflickr.com/3726/9571272211_6873a571

  1. Removal to Maximum Extent Practical

    Broader source: Energy.gov [DOE]

    Summary Notes from 1 November 2007 Generic Technical Issue Discussion on Removal of Highly Radioactive Radionuclides/Key Radionuclides to the Maximum Extent Practical

  2. Removing Arsenic from Drinking Water

    SciTech Connect (OSTI)

    2011-01-01

    See how INL scientists are using nanotechnology to remove arsenic from drinking water. For more INL research, visit http://www.facebook.com/idahonationallaboratory

  3. Removing Arsenic from Drinking Water

    ScienceCinema (OSTI)

    None

    2013-05-28

    See how INL scientists are using nanotechnology to remove arsenic from drinking water. For more INL research, visit http://www.facebook.com/idahonationallaboratory

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

    Energy Savers [EERE]

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

  5. REFLECTOR FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Fraas, A.P.

    1963-08-01

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

  6. Influence factors on the flue gas desulfurization in the circulating fluidized bed reactor

    SciTech Connect (OSTI)

    Gao, J.; Tang, D.; Liu, H.; Suzuki, Yoshizo; Kito, Nobo

    1997-12-31

    This paper describes a dry SO{sub 2} removal method -- the absorbent (Ca(OH){sub 2}) was injected into the Circulating Fluidized Bed (CFB) reactor at the coolside of the duct to abate SO{sub 2} in the flue gas -- with the potential to significantly enhance desulfurization performance over that of existing dry/semi-dry Flue Gas Desulfurization (FGD) technology such as Spray Drying. A patent for coolside Flue Gas Desulfurization in the Circulating Fluidized Bed reactor (CFB-FGD) was approved by the China Patent Bureau in September of 1995 and the additional laboratory experiment was carried out in an electrically heated bench scale quartz circulating fluidized bed reactor of 2350mm in height and 23mm in diameter in January, 1996. The influences of steam, ratio of calcium and sulfur, reactor temperature, and absorbent utilization efficiency were invested. The results show that: (1) Water steam plays a key role in the reaction of Ca(OH){sub 2} and SO{sub 2} in the CFB reactor; (2) There is a positive effect of Ca/S on SO{sub 2} removal efficiency; (3) The temperature is an another key factor for SO{sub 2} removal efficiency for the CFB-FGD process; (4) The absorbent can be enhanced in the CFB reactor; (5) The CFB reactor is better than the dry/semi-dry FDG technology. SO{sub 2} removal efficiency can be as high as 84.8%.

  7. Pulsations and period changes of the non-Blazhko RR lyrae variable Y oct observed from Dome A, Antarctica

    SciTech Connect (OSTI)

    Zhihua, Huang; Jianning, Fu; Weikai, Zong; Lingzhi, Wang; Zonghong, Zhu; M, Macri Lucas; Lifan, Wang; Ashley, Michael C. B.; S, Lawrence Jon; Daniel, Luong-Van; Xiangqun, Cui; Long-Long, Feng; Xuefei, Gong; Qiang, Liu; Huigen, Yang; Xiangyan, Yuan; Xu, Zhou; Zhenxi, Zhu; R, Pennypacker Carl; G, York Donald

    2015-01-01

    During the operation of the Chinese Small Telescope Array (CSTAR) in Dome A of Antarctica in the years 2008, 2009, and 2010, large amounts of photometric data have been obtained for variable stars in the CSTAR field. We present here the study of one of six RR Lyrae variables, Y Oct, observed with CSTAR in Dome A, Antarctica. Photometric data in the i band were obtained in 2008 and 2010, with a duty cycle (defined as the fraction of time representing scientifically available data to CSTAR observation time) of about 44% and 52%, respectively. In 2009, photometric data in the g and r bands were gathered for this star, with a duty cycle of 65% and 60%, respectively. Fourier analysis of the data in the three bands only shows the fundamental frequency and its harmonics, which is characteristic of the non-Blazhko RR Lyrae variables. Values of the fundamental frequency and the amplitudes, as well as the total pulsation amplitude, are obtained from the data in the three bands separately. The amplitude of the fundamental frequency and the total pulsation amplitude in the g band are the largest, and those in the i band the smallest. Two-hundred fifty-one times of maximum are obtained from the three seasons of data, which are analyzed together with 38 maximum times provided in the GEOS RR Lyrae database. A period change rate of −0.96 ± 0.07 days Myr{sup −1} is then obtained, which is a surprisingly large negative value. Based on relations available in the literature, the following physical parameters are derived: [Fe/H] = −1.41 ± 0.14, M{sub V} = 0.696 ± 0.014 mag, V−K = 1.182 ± 0.028 mag, logT{sub eff} = 3.802 ± 0.003 K, logg = 2.705 ± 0.004, logL/L{sub ⊙} = 1.625 ± 0.013, and logM/M{sub ⊙} = −0.240 ± 0.019.

  8. Method for automatically scramming a nuclear reactor

    DOE Patents [OSTI]

    Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.

    2005-12-27

    An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

  9. Plasma generators, reactor systems and related methods

    DOE Patents [OSTI]

    Kong, Peter C. (Idaho Falls, ID); Pink, Robert J. (Pocatello, ID); Lee, James E. (Idaho Falls, ID)

    2007-06-19

    A plasma generator, reactor 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 coupled together to form a chamber. Associated with each section is an electrode set including three electrodes with each electrode being coupled to a single phase of a three-phase alternating current (AC) power supply. The electrodes are disposed about a longitudinal centerline of the chamber and are arranged to provide and extended arc and generate an extended body of plasma. The electrodes are displaceable relative to the longitudinal centerline of the chamber. A control system may be utilized so as to automatically displace the electrodes and define an electrode gap responsive to measure voltage or current levels of the associated power supply.

  10. Advanced Reactor Technologies | Department of Energy

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

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

  11. Treatment of VOCs in high strength wastes using an anaerobic expanded-bed GAC reactor

    SciTech Connect (OSTI)

    Narayanian, B.; Suidan, M.T.; Gelderloos, A.B.; Brenner, R.C.

    1993-01-01

    The potential of the expanded-bed granular activated carbon (GAC) anaerobic reactor in treating a high strength waste containing RCRA volatile organic compounds (VOCs) was studied. A total of six VOCs, methylene chloride, chlorobenzene, carbon tetrachloride, chloroform, toluene and tetrachloroethylene, were fed to the reactor in a high strength matrix of background solvents. Performance was evaluated. The reactor was found to effect excellent removal of all VOCs (97%). Chloroform, while itself removed at levels in excess of 97%, was found to inhibit the degradation of acetate and acetone, two of the background solvents. Without any source of chloroform in the feed, excellent COD removals were obtained in addition to near-complete removal of all the VOCs.

  12. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01

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

  13. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M.

    1982-01-01

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

  14. NUCLEAR REACTOR FUEL SYSTEMS

    DOE Patents [OSTI]

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

    1959-09-15

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

  15. CONTROL FOR NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1959-03-31

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

  16. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03

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

  17. Microfluidic electrochemical reactors

    DOE Patents [OSTI]

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

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

  18. Fast Breeder Reactor studies

    SciTech Connect (OSTI)

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

    1980-07-01

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

  19. Integral Fast Reactor fuel pin processor

    SciTech Connect (OSTI)

    Levinskas, D.

    1993-03-01

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

  20. Integral Fast Reactor fuel pin processor

    SciTech Connect (OSTI)

    Levinskas, D.

    1993-01-01

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

  1. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    SciTech Connect (OSTI)

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions between the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very effective in removing decay heat. By removing the limit on the decay heat removal capability due to the limited available surface area as in a RVACS, the reactor power and power density can be significantly increased, without losing the passive heat removal feature. This paper will introduce the concept of using DRACS to enhance VHTR passive safety and economics. Three design options will be discussed, depending on the cooling pipe locations. Analysis results from a lumped volume based model and CFD simulations will be presented.

  2. Nuclear reactor fuel rod attachment system

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA)

    1982-01-01

    A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

  3. ``Sleeping reactor`` irradiations: Shutdown reactor determination of short-lived activation products

    SciTech Connect (OSTI)

    Jerde, E.A.; Glasgow, D.C.

    1998-09-01

    At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux ({phi}) of {approximately} 4 {times} 10{sup 14} n/cm{sup 2} {center_dot} s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of {approximately} 6 s, but the requirement of immediate counting leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about {+-} 0.5 s) make irradiations of < 6 s less reliable. Therefore, the determination of these ultra-short-lived species in mixed matrices has not generally been made at HFIR. The authors have found that very short lived activation products can be produced easily during the period after reactor shutdown (SCRAM), but prior to the removal of spent fuel elements. During this 24- to 36-h period (dubbed the ``sleeping reactor``), neutrons are produced in the beryllium reflector by the reaction {sup 9}Be({gamma},n){sup 8}Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to {approximately} 1 {times} 10{sup 10} n/cm{sup 2} {center_dot} s within 1 h. By the time the fuel elements are removed, the flux has dropped to {approximately} 6 {times} 10{sup 8}. Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant.

  4. Method and system for the removal of oxides of nitrogen and sulfur from combustion processes

    DOE Patents [OSTI]

    Walsh, John V. (Glendora, CA)

    1987-12-15

    A process for removing oxide contaminants from combustion gas, and employing a solid electrolyte reactor, includes: (a) flowing the combustion gas into a zone containing a solid electrolyte and applying a voltage and at elevated temperature to thereby separate oxygen via the solid electrolyte, (b) removing oxygen from that zone in a first stream and removing hot effluent gas from that zone in a second stream, the effluent gas containing contaminant, (c) and pre-heating the combustion gas flowing to that zone by passing it in heat exchange relation with the hot effluent gas.

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

    Energy Savers [EERE]

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

  6. EIS-0075: Strategic Petroleum Reserve Phase III Development, Texoma and Seaway Group Salt Domes (West Hackberry and Bryan Mound Expansion, Big Hill Development) Cameron Parish, Louisiana, and Brazoria and Jefferson Counties, Texas

    Broader source: Energy.gov [DOE]

    Also see EIS-0021 and EIS-0029. The Strategic Petroleum Reserve (SPR) Office developed this EIS to assess the environmental impacts of expanding the existing SPR storage capacity from 538 million to 750 million barrels of storage and increasing the drawdown capability from 3.5 million to 4.5 million barrels per day. This EIS incorperates two previously issued EISs: DOE/EIS-0021, Seaway Group of Salt Domes, and DOE/EIS-0029, Texoma Group of Salt Domes.

  7. ADVANCES IN HEXAVALENT CHROMIUM REMOVAL AT HANFORD

    SciTech Connect (OSTI)

    NESHEM DO; RIDDELLE J

    2012-01-30

    At the Hanford Site, chromium was used as a corrosion inhibitor in the reactor cooling water and was introduced into the groundwater as a result of planned and unplanned discharges from reactors during plutonium production since 1944. Beginning in 1995, groundwater treatment methods were evaluated leading to the use of pump and treat facilities with ion exchange using Dowex 21 K, a regenerable strong base anion exchange resin. This required regeneration of the resin, which is currently performed offsite. Resin was installed in a 4 vessel train, with resin removal required from the lead vessel approximately once a month. In 2007, there were 8 trains (32 vessels) in operation. In 2008, DOE recognized that regulatory agreements would require significant expansion in the groundwater chromium treatment capacity. Previous experience from one of the DOE project managers led to identification of a possible alternative resin, and the contractor was requested to evaluate alternative resins for both cost and programmatic risk reductions. Testing was performed onsite in 2009 and 2010, using a variety of potential resins in two separate facilities with groundwater from specific remediation sites to demonstrate resin performance in the specific groundwater chemistry at each site. The testing demonstrated that a weak base anion single-use resin, ResinTech SIR-700, was effective at removing chromium, had a significantly higher capacity, could be disposed of efficiently on site, and would eliminate the complexities and programmatic risks from sampling, packaging, transportation and return of resin for regeneration. This resin was installed in Hanford's newest groundwater treatment facility, called 100-DX, which began operations in November, 2010, and used in a sister facility, 100-HX, which started up in September of 2011. This increased chromium treatment capacity to 25 trains (100 vessels). The resin is also being tested in existing facilities that utilize Dowex 21 K for conversion to the new resin. This paper will describe the results of the testing, performance in the facilities, continued optimization in the pump and treat facilities, and the estimated savings and non-tangible benefits of the conversion.

  8. Method of and apparatus for removing silicon from a high temperature sodium coolant

    DOE Patents [OSTI]

    Yunker, Wayne H.; Christiansen, David W.

    1987-01-01

    A method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

  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 Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A...

  10. C Reactor - Hanford Site

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

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

  11. F Reactor - Hanford Site

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

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

  12. Section 46: Removal of Waste

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

    in and around the WIPP site, the EPA did not identify any significant changes in the planning and execution of the DOE's strategy for removal of waste since the 1998...

  13. B Reactor | Department of Energy

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

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

  14. Compact power reactor

    DOE Patents [OSTI]

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

    1978-01-01

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

  15. Molten metal reactors

    DOE Patents [OSTI]

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

    2013-11-05

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

  16. Ion Removal - Energy Innovation Portal

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

    Advanced Materials Advanced Materials Find More Like This Return to Search Ion Removal Idaho National Laboratory Contact INL About This Technology Technology Marketing Summary INL's ion removal technology leverages the ability of phosphazene polymers discriminate between water and metal ions, which allows water to pass through the membrane while retaining the ions. Description The inherent chemical and thermal stability of the phosphazene polymers are an added strengths for separating and

  17. Ethanol Dehydration to Ethylene in a Stratified Autothermal Millisecond Reactor

    SciTech Connect (OSTI)

    Skinner, MJ; Michor, EL; Fan, W; Tsapatsis, M; Bhan, A; Schmidt, LD

    2011-08-10

    The concurrent decomposition and deoxygenation of ethanol was accomplished in a stratified reactor with 50-80 ms contact times. The stratified reactor comprised an upstream oxidation zone that contained Pt-coated Al(2)O(3) beads and a downstream dehydration zone consisting of H-ZSM-5 zeolite films deposited on Al(2)O(3) monoliths. Ethanol conversion, product selectivity, and reactor temperature profiles were measured for a range of fuel:oxygen ratios for two autothermal reactor configurations using two different sacrificial fuel mixtures: a parallel hydrogen-ethanol feed system and a series methane-ethanol feed system. Increasing the amount of oxygen relative to the fuel resulted in a monotonic increase in ethanol conversion in both reaction zones. The majority of the converted carbon was in the form of ethylene, where the ethanol carbon-carbon bonds stayed intact while the oxygen was removed. Over 90% yield of ethylene was achieved by using methane as a sacrificial fuel. These results demonstrate that noble metals can be successfully paired with zeolites to create a stratified autothermal reactor capable of removing oxygen from biomass model compounds in a compact, continuous flow system that can be configured to have multiple feed inputs, depending on process restrictions.

  18. Process for removing copper in a recoverable form from solid scrap metal

    DOE Patents [OSTI]

    Hartman, Alan D. (Albany, OR); Oden, Laurance L. (Albany, OR); White, Jack C. (Albany, OR)

    1995-01-01

    A process for removing copper in a recoverable form from a copper/solid ferrous scrap metal mix is disclosed. The process begins by placing a copper/solid ferrous scrap metal mix into a reactor vessel. The atmosphere within the reactor vessel is purged with an inert gas or oxidizing while the reactor vessel is heated in the area of the copper/solid ferrous scrap metal mix to raise the temperature within the reactor vessel to a selected elevated temperature. Air is introduced into the reactor vessel and thereafter hydrogen chloride is introduced into the reactor vessel to obtain a desired air-hydrogen chloride mix. The air-hydrogen chloride mix is operable to form an oxidizing and chloridizing atmosphere which provides a protective oxide coating on the surface of the solid ferrous scrap metal in the mix and simultaneously oxidizes/chloridizes the copper in the mix to convert the copper to a copper monochloride gas for transport away from the solid ferrous scrap metal. After the copper is completely removed from the copper/solid ferrous scrap metal mix, the flows of air and hydrogen chloride are stopped and the copper monochloride gas is collected for conversion to a recoverable copper species.

  19. Axi-symmetrical flow reactor for .sup.196 Hg photochemical enrichment

    DOE Patents [OSTI]

    Grossman, Mark W.

    1991-01-01

    The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, .sup.196 Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired .sup.196 Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith.

  20. Axi-symmetrical flow reactor for [sup 196]Hg photochemical enrichment

    DOE Patents [OSTI]

    Grossman, M.W.

    1991-04-30

    The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, [sup 196]Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired [sup 196]Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith. 10 figures.

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

    SciTech Connect (OSTI)

    P. Delmolino

    2005-05-06

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

  2. High removal rate laser-based coating removal system

    DOE Patents [OSTI]

    Matthews, Dennis L. (Moss Beach, CA); Celliers, Peter M. (Berkeley, CA); Hackel, Lloyd (Livermore, CA); Da Silva, Luiz B. (Danville, CA); Dane, C. Brent (Livermore, CA); Mrowka, Stanley (Richmond, CA)

    1999-11-16

    A compact laser system that removes surface coatings (such as paint, dirt, etc.) at a removal rate as high as 1000 ft.sup.2 /hr or more without damaging the surface. A high repetition rate laser with multiple amplification passes propagating through at least one optical amplifier is used, along with a delivery system consisting of a telescoping and articulating tube which also contains an evacuation system for simultaneously sweeping up the debris produced in the process. The amplified beam can be converted to an output beam by passively switching the polarization of at least one amplified beam. The system also has a personal safety system which protects against accidental exposures.

  3. Removal - An alternative to clearance

    SciTech Connect (OSTI)

    Feinhals, J.; Kelch, A.; Kunze, V.

    2007-07-01

    This presentation shows the differences between the application of clearance and removal, both being procedures for materials leaving radiation protection areas permanently. The differentiation will be done on the basis of the German legislation but may be also applicable for other national legislation. For clearance in Germany two basic requirements must be given, i.e. that the materials are activated or contaminated and that they result from the licensed use or can be assigned to the scope of the license. Clearance needs not to be applied to objects in Germany which are to be removed only temporarily from controlled areas with the purpose of repair or reuse in other controlled areas. In these cases only the requirements of contamination control apply. In the case of removal it must either be proved by measurements that the relevant materials are neither activated nor contaminated or that the materials result from areas where activation or contamination is impossible due to the operational history considering operational procedures and events. If the material is considered neither activated nor contaminated there is no need for a clearance procedure. Therefore, these materials can be removed from radiation protection areas and the removal is in the responsibility of the licensee. Nevertheless, the removal procedure and the measuring techniques to be applied for the different types of materials need an agreement from the competent authority. In Germany a maximum value of 10% of the clearance values has been established in different licenses as a criterion for the application of removal. As approximately 2/3 of the total mass of a nuclear power plant is not expected to be contaminated or activated there is a need for such a procedure of removal for this non contaminated material without any regulatory control especially in the case of decommissioning. A remarkable example is NPP Stade where in the last three years more than 8600 Mg were disposed of by removal and only 315 Mg were released by clearance, even before the decommissioning licensing procedure was finished. (authors)

  4. Protection #2: Trap and Remove Sediment

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

    Trap and Remove Sediment Protection 2: Trap and Remove Sediment The 3 Protections Defense in Depth August 1, 2013 Sediment behind LA Canyon weir is sampled and excavated...

  5. Zero Power Reactor simulation | Argonne National Laboratory

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

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

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

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

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

  7. Methanation assembly using multiple reactors

    DOE Patents [OSTI]

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

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

  8. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    1986-01-01

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

  9. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-08-01

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

  10. Reactor for exothermic reactions

    DOE Patents [OSTI]

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

    1993-01-01

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

  11. Dynamic bed reactor

    DOE Patents [OSTI]

    Stormo, Keith E. (Moscow, ID)

    1996-07-02

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

  12. Reactor for exothermic reactions

    DOE Patents [OSTI]

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

    1993-03-02

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

  13. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

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

  14. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

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

    1987-01-01

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

  15. Heat exchanger for reactor core and the like

    DOE Patents [OSTI]

    Kaufman, Jay S. (Del Mar, CA); Kissinger, John A. (Del Mar, CA)

    1986-01-01

    A compact bayonet tube type heat exchanger which finds particular application as an auxiliary heat exchanger for transfer of heat from a reactor gas coolant to a secondary fluid medium. The heat exchanger is supported within a vertical cavity in a reactor vessel intersected by a reactor coolant passage at its upper end and having a reactor coolant return duct spaced below the inlet passage. The heat exchanger includes a plurality of relatively short length bayonet type heat exchange tube assemblies adapted to pass a secondary fluid medium therethrough and supported by primary and secondary tube sheets which are releasibly supported in a manner to facilitate removal and inspection of the bayonet tube assemblies from an access area below the heat exchanger. Inner and outer shrouds extend circumferentially of the tube assemblies and cause the reactor coolant to flow downwardly internally of the shrouds over the tube bundle and exit through the lower end of the inner shroud for passage to the return duct in the reactor vessel.

  16. Design of megawatt power level heat pipe reactors

    SciTech Connect (OSTI)

    Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao; Reid, Robert Stowers

    2015-11-12

    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 strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

  17. Fast quench reactor method

    DOE Patents [OSTI]

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

    1999-01-01

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

  18. MEANS FOR SHIELDING REACTORS

    DOE Patents [OSTI]

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

    1959-03-10

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

  19. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, Daniel L. (Princeton, NJ)

    1988-01-01

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

  20. Plug Flow Reactor Simulator

    Energy Science and Technology Software Center (OSTI)

    1996-07-30

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

  1. Nuclear reactor apparatus

    DOE Patents [OSTI]

    Wade, Elman E.

    1978-01-01

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

  2. Fast quench reactor method

    DOE Patents [OSTI]

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

    1999-08-10

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

  3. Perspectives on reactor safety

    SciTech Connect (OSTI)

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

    1994-03-01

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

  4. 2011sr01.doc

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

    Tuesday, February 1, 2011 james-r.giusti@srs.gov Paivi Nettamo, SRNS, (803) 646-6075 paivi.nettamo@srs.gov Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project 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 Cold War-era test reactor. This $25-million reactor decommissioning and deactivation project is funded By the American Recovery and

  5. Pressurized water nuclear reactor system with hot leg vortex mitigator

    DOE Patents [OSTI]

    Lau, Louis K. S. (Monroeville, PA)

    1990-01-01

    A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

  6. An Innovative Reactor Technology to Improve Indoor Air Quality

    SciTech Connect (OSTI)

    Rempel, Jane

    2013-03-30

    As residential buildings achieve tighter envelopes in order to minimize energy used for space heating and cooling, accumulation of indoor air pollutants such as volatile organic compounds (VOCs), becomes a major concern causing poor air quality and increased health risks. Current VOC removal methods include sorbents, ultraviolet photocatalytic oxidation (UVPCO), and increased ventilation, but these methods do not capture or destroy all VOCs or are prohibitively expensive to implement. TIAX's objective in this program was to develop a new VOC removal technology for residential buildings. This novel air purification technology is based on an innovative reactor and light source design along with UVPCO properties of the chosen catalyst to purify indoor air and enhance indoor air quality (IAQ). During the program we designed, fabricated and tested a prototype air purifier to demonstrate its feasibility and effectiveness. We also measured kinetics of VOC destruction on photocatalysts, providing deep insight into reactor design.

  7. removal

    National Nuclear Security Administration (NNSA)

    80 pounds) of highly enriched uranium (HEU) spent fuel from the Institute of Nuclear Physics (INP) in Almaty, Kazakhstan. The HEU was transported via two air shipments to a...

  8. Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

    SciTech Connect (OSTI)

    Koenig, D.R.; Gido, R.G.; Brandon, D.I.

    1985-01-01

    The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations.

  9. B Reactor | Department of Energy

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

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

  10. Piqua, Ohio, Decommissioned Reactor Site

    Office of Legacy Management (LM)

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

  11. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

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

  12. Reactor Materials | Department of Energy

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

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

  13. Hallam, Nebraska, Decommissioned Reactor Site

    Office of Legacy Management (LM)

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

  14. Thermal Reactor Safety

    SciTech Connect (OSTI)

    Not Available

    1980-06-01

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

  15. Cermet fuel reactors

    SciTech Connect (OSTI)

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

    1987-09-01

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

  16. Fossil fuel furnace reactor

    DOE Patents [OSTI]

    Parkinson, William J.

    1987-01-01

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

  17. JACKETED REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

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

    1958-12-01

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

  18. NETL - Chemical Looping Reactor

    SciTech Connect (OSTI)

    2013-07-24

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

  19. Nuclear reactor building

    DOE Patents [OSTI]

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

    1994-01-01

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

  20. NETL - Chemical Looping Reactor

    ScienceCinema (OSTI)

    None

    2014-06-26

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

  1. Nuclear Reactors and Technology

    SciTech Connect (OSTI)

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

    1992-01-01

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

  2. Nuclear reactor building

    DOE Patents [OSTI]

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

    1994-04-05

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

  3. Support arrangements for core modules of nuclear reactors. [PWR

    DOE Patents [OSTI]

    Bollinger, L.R.

    1983-11-03

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

  4. Support arrangement for core modules of nuclear reactors

    DOE Patents [OSTI]

    Bollinger, Lawrence R. (Schenectady, NY)

    1987-01-01

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

  5. Method and apparatus for a combination moving bed thermal treatment reactor and moving bed filter

    DOE Patents [OSTI]

    Badger, Phillip C.; Dunn, Jr., Kenneth J.

    2015-09-01

    A moving bed gasification/thermal treatment reactor includes a geometry in which moving bed reactor particles serve as both a moving bed filter and a heat carrier to provide thermal energy for thermal treatment reactions, such that the moving bed filter and the heat carrier are one and the same to remove solid particulates or droplets generated by thermal treatment processes or injected into the moving bed filter from other sources.

  6. METHOD OF MEASURING THE INTEGRATED ENERGY OUTPUT OF A NEUTRONIC CHAIN REACTOR

    DOE Patents [OSTI]

    Sturm, W.J.

    1958-12-01

    A method is presented for measuring the integrated energy output of a reactor conslsting of the steps of successively irradiating calibrated thin foils of an element, such as gold, which is rendered radioactive by exposure to neutron flux for periods of time not greater than one-fifth the mean life of the induced radioactlvity and producing an indication of the radioactivity induced in each foil, each foil belng introduced into the reactor immediately upon removal of its predecessor.

  7. Passive cooling system for top entry liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

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

    1992-01-01

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

  8. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    SciTech Connect (OSTI)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  9. REACTOR FUEL ELEMENTS TESTING CONTAINER

    DOE Patents [OSTI]

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

    1963-01-15

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

  10. EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

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

    1963-12-24

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

  11. Part 3: Removal Action | Department of Energy

    Office of Environmental Management (EM)

    3: Removal Action Part 3: Removal Action Question: When may removal actions be initiated? Answer: Removal actions may be initiated when DOE determines that the action will prevent, minimize, stabilize, or eliminate a risk to health or the environment. The NCP specifies that the determination that a risk to health or the environment is appropriate for removal action should be based on: actual or potential exposure of humans, animals, or the food chain the presence of contained hazardous

  12. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    SciTech Connect (OSTI)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  13. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

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

    1980-05-09

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

  14. Industrial lead paint removal specifications

    SciTech Connect (OSTI)

    Stone, R.C.

    1997-06-01

    The purpose of this paper is to inform the reader as to some of the pertinent rules and regulations promulgated by the Environmental Protection Agency (EPA) and the Occupational Safety and Health Administration (OSHA) that may effect an industrial lead paint removal project. The paper discusses a recommended schedule of procedures and preparations to be followed by the lead paint removal specification writer when analyzing the possible impact of the project on the environment, the public and workers. Implications of the Clean Air Act, the Clean Water Act and the Resource Conservation and Recovery Act (RCRA) along with hazardous waste handling, manifesting, transporting and disposal procedures are discussed with special emphasis placed as to their impact on the writer and the facility owner. As the rules and regulations are highly complex, the writer has attempted to explain the methodology currently being used in state-of-the-art industrial lead abatement specifications.

  15. removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    removal | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets Newsletters Press Releases Photo Gallery Jobs Apply for Our Jobs Our Jobs Working at NNSA Blog Home

  16. Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

    SciTech Connect (OSTI)

    Scheele, Randall D.; Casella, Andrew M.

    2010-09-28

    This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

  17. Nuclear reactor construction with bottom supported reactor vessel

    DOE Patents [OSTI]

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

    1987-01-01

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

  18. HOMOGENEOUS NUCLEAR REACTOR

    DOE Patents [OSTI]

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

    1959-02-17

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

  19. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, E.

    1983-08-15

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

  20. ENGINEERING TEST REACTOR

    DOE Patents [OSTI]

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

    1958-10-21

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

  1. Neutronic reactor construction

    DOE Patents [OSTI]

    Huston, Norman E.

    1976-07-06

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

  2. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, D.L.

    1987-09-04

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

  3. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M.

    1989-01-01

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

  4. Nuclear reactor shutdown system

    DOE Patents [OSTI]

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

    1981-01-01

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

  5. Interim Safe Storage of Plutonium Production Reactors at the US DOE Hanford Site - 13438

    SciTech Connect (OSTI)

    Schilperoort, Daryl L.; Faulk, Darrin

    2013-07-01

    Nine plutonium production reactors located on DOE's Hanford Site are being placed into an Interim Safe Storage (ISS) period that extends to 2068. The Environmental Impact Statement (EIS) for ISS [1] was completed in 1993 and proposed a 75-year storage period that began when the EIS was finalized. Remote electronic monitoring of the temperature and water level alarms inside the safe storage enclosure (SSE) with visual inspection inside the SSE every 5 years are the only planned operational activities during this ISS period. At the end of the ISS period, the reactor cores will be removed intact and buried in a landfill on the Hanford Site. The ISS period allows for radioactive decay of isotopes, primarily Co-60 and Cs-137, to reduce the dose exposure during disposal of the reactor cores. Six of the nine reactors have been placed into ISS by having an SSE constructed around the reactor core. (authors)

  6. LOADING MACHINE FOR REACTORS

    DOE Patents [OSTI]

    Simon, S.L.

    1959-07-01

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

  7. In situ reactor

    DOE Patents [OSTI]

    Radtke, Corey William; Blackwelder, David Bradley

    2004-01-27

    An in situ reactor for use in a geological strata, is described and which includes a liner defining a centrally disposed passageway and which is placed in a borehole formed in the geological strata; and a sampling conduit is received within the passageway defined by the liner and which receives a geological specimen which is derived from the geological strata, and wherein the sampling conduit is in fluid communication with the passageway defined by the liner.

  8. NUCLEAR REACTOR CORE DESIGN

    DOE Patents [OSTI]

    Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

    1960-03-22

    An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

  9. Nuclear reactor sealing system

    DOE Patents [OSTI]

    McEdwards, James A. (Calabasas, CA)

    1983-01-01

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

  10. DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear

    Energy Savers [EERE]

    Weapons Stockpile | Department of Energy to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile November 7, 2005 - 12:38pm Addthis Will Be Redirected to Naval Reactors, Down-blended or Used for Space Programs WASHINGTON, DC - Secretary of Energy Samuel W. Bodman today announced that the Department of Energy's (DOE) National Nuclear Security Administration (NNSA) will

  11. EM's Paducah Site Completes Building Removals | Department of Energy

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

    EM's Paducah Site Completes Building Removals EM's Paducah Site Completes Building Removals Addthis

  12. Fluid sampling system for a nuclear reactor

    DOE Patents [OSTI]

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

    1994-11-22

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

  13. Fluid sampling system for a nuclear reactor

    DOE Patents [OSTI]

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

    1994-01-01

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

  14. Dismantling of Loop-Type Channel Equipment of MR Reactor in NRC 'Kurchatov Institute' - 13040

    SciTech Connect (OSTI)

    Volkov, Victor; Danilovich, Alexey; Zverkov, Yuri; Ivanov, Oleg; Kolyadin, Vyacheslav; Lemus, Alexey; Pavlenko, Vitaly; Semenov, Sergey; Fadin, Sergey; Shisha, Anatoly; Chesnokov, Alexander

    2013-07-01

    In 2009 the project of decommissioning of MR and RTF reactors was developed and approved by the Expert Authority of the Russian Federation (Gosexpertiza). The main objective of the decommissioning works identified in this project: - complete dismantling of reactor equipment and systems; - decontamination of reactor premises and site in accordance with the established sanitary and hygienic standards. At the preparatory stage (2008-2010) of the project the following works were executed: loop-type channels' dismantling in the storage pool; experimental fuel assemblies' removal from spent fuel repositories in the central hall; spent fuel assembly removal from the liquid-metal-cooled loop-type channel of the reactor core and its placement into the SNF repository; and reconstruction of engineering support systems to the extent necessary for reactor decommissioning. The project assumes three main phases of dismantling and decontamination: - dismantling of equipment/pipelines of cooling circuits and loop-type channels, and auxiliary reactor equipment (2011-2012); - dismantling of equipment in underground reactor premises and of both MR and RTF in-vessel devices (2013-2014); - decontamination of reactor premises; rehabilitation of the reactor site; final radiation survey of reactor premises, loop-type channels and site; and issuance of the regulatory authorities' de-registration statement (2015). In 2011 the decommissioning license for the two reactors was received and direct MR decommissioning activities started. MR primary pipelines and loop-type facilities situated in the underground reactor hall were dismantled. Works were also launched to dismantle the loop-type channels' equipment in underground reactor premises; reactor buildings were reconstructed to allow removal of dismantled equipment; and the MR/RTF decommissioning sequence was identified. In autumn 2011 - spring 2012 results of dismantling activities performed are: - equipment from underground rooms (No. 66, 66A, 66B, 72, 64, 63) - as well as from water and gas loop corridors - was dismantled, with the total radwaste weight of 53 tons and the total removed activity of 5,0 x 10{sup 10} Bq; - loop-type channel equipment from underground reactor hall premises was dismantled; - 93 loop-type channels were characterized, chopped and removed, with radwaste of 2.6 x 10{sup 13} Bq ({sup 60}Co) and 1.5 x 10{sup 13} Bq ({sup 137}Cs) total activity removed from the reactor pool, fragmented and packaged. Some of this waste was placed into the high-level waste (HLW) repository of the Center. Dismantling works were executed with application of remotely operated mechanisms, which promoted decrease of radiation impact on the personnel. The average individual dose for the personnel was 1.9 mSv/year in 2011, and the collective dose is estimated as 0.0605 man x Sv/year. (authors)

  15. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

  16. Water Distribution and Removal Model

    SciTech Connect (OSTI)

    Y. Deng; N. Chipman; E.L. Hardin

    2005-08-26

    The design of the Yucca Mountain high level radioactive waste repository depends on the performance of the engineered barrier system (EBS). To support the total system performance assessment (TSPA), the Engineered Barrier System Degradation, Flow, and Transport Process Model Report (EBS PMR) is developed to describe the thermal, mechanical, chemical, hydrological, biological, and radionuclide transport processes within the emplacement drifts, which includes the following major analysis/model reports (AMRs): (1) EBS Water Distribution and Removal (WD&R) Model; (2) EBS Physical and Chemical Environment (P&CE) Model; (3) EBS Radionuclide Transport (EBS RNT) Model; and (4) EBS Multiscale Thermohydrologic (TH) Model. Technical information, including data, analyses, models, software, and supporting documents will be provided to defend the applicability of these models for their intended purpose of evaluating the postclosure performance of the Yucca Mountain repository system. The WD&R model ARM is important to the site recommendation. Water distribution and removal represents one component of the overall EBS. Under some conditions, liquid water will seep into emplacement drifts through fractures in the host rock and move generally downward, potentially contacting waste packages. After waste packages are breached by corrosion, some of this seepage water will contact the waste, dissolve or suspend radionuclides, and ultimately carry radionuclides through the EBS to the near-field host rock. Lateral diversion of liquid water within the drift will occur at the inner drift surface, and more significantly from the operation of engineered structures such as drip shields and the outer surface of waste packages. If most of the seepage flux can be diverted laterally and removed from the drifts before contacting the wastes, the release of radionuclides from the EBS can be controlled, resulting in a proportional reduction in dose release at the accessible environment. The purposes of this WD&R model (CRWMS M&O 2000b) are to quantify and evaluate the distribution and drainage of seepage water within emplacement drifts during the period of compliance for post-closure performance. The model bounds the fraction of water entering the drift that will be prevented from contacting the waste by the combined effects of engineered controls on water distribution and on water removal. For example, water can be removed during pre-closure operation by ventilation and after closure by natural drainage into the fractured rock. Engineered drains could be used, if demonstrated to be necessary and effective, to ensure that adequate drainage capacity is provided. This report provides the screening arguments for certain Features, Events, and Processes (FEPs) that are related to water distribution and removal in the EBS. Applicable acceptance criteria from the Issue Resolution Status Reports (IRSRs) developed by the U.S. Nuclear Regulatory Commission (NRC 1999a; 1999b; 1999c; and 1999d) are also addressed in this document.

  17. Advanced Reactor Technology Documents | Department of Energy

    Energy Savers [EERE]

    Nuclear Reactor Technologies » Advanced Reactor Technologies » Advanced Reactor Technology Documents Advanced Reactor Technology Documents January 30, 2013 Advanced Reactor Concepts Technical Review Panel Report This report documents the establishment of a technical review process and the findings of the Advanced Reactor Concepts (ARC) Technical Review Panel (TRP).1 The intent of the process is to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D

  18. Advanced Nuclear Reactors | Department of Energy

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

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

  19. Removal of oxides of nitrogen from gases in multi-stage coal combustion

    DOE Patents [OSTI]

    Mollot, Darren J. (Morgantown, WV); Bonk, Donald L. (Louisville, OH); Dowdy, Thomas E. (Orlando, FL)

    1998-01-01

    Polluting NO.sub.x gas values are removed from off-gas of a multi-stage coal combustion process which includes an initial carbonizing reaction, firing of char from this reaction in a fluidized bed reactor, and burning of gases from the carbonizing and fluidized bed reactions in a topping combustor having a first, fuel-rich zone and a second, fuel-lean zone. The improvement by means of which NO.sub.x gases are removed is directed to introducing NO.sub.x -free oxidizing gas such as compressor air into the second, fuel-lean zone and completing combustion with this source of oxidizing gas. Excess air fed to the fluidized bed reactor is also controlled to obtain desired stoichiometry in the first, fuel-rich zone of the topping combustor.

  20. Removal of oxides of nitrogen from gases in multi-stage coal combustion

    DOE Patents [OSTI]

    Mollot, D.J.; Bonk, D.L.; Dowdy, T.E.

    1998-01-13

    Polluting NO{sub x} gas values are removed from off-gas of a multi-stage coal combustion process which includes an initial carbonizing reaction, firing of char from this reaction in a fluidized bed reactor, and burning of gases from the carbonizing and fluidized bed reactions in a topping combustor having a first, fuel-rich zone and a second, fuel-lean zone. The improvement by means of which NO{sub x} gases are removed is directed to introducing NO{sub x}-free oxidizing gas such as compressor air into the second, fuel-lean zone and completing combustion with this source of oxidizing gas. Excess air fed to the fluidized bed reactor is also controlled to obtain desired stoichiometry in the first, fuel-rich zone of the topping combustor. 2 figs.

  1. Mexico HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Apply for Our Jobs Our Jobs Working at NNSA Blog Home content Four-Year Plan Mexico HEU Removal Mexico HEU Removal Location Mexico United States 24 24' 35.298" N, 102...

  2. Nuclear & Radiological Material Removal | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    & Radiological Material Removal | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation...

  3. Method of and apparatus for removing silicon from a high temperature sodium coolant

    DOE Patents [OSTI]

    Yunker, W.H.; Christiansen, D.W.

    1983-11-25

    This patent discloses a method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

  4. Nuclear reactor cooling system decontamination reagent regeneration

    DOE Patents [OSTI]

    Anstine, Larry D. (San Jose, CA); James, Dean B. (Saratoga, CA); Melaika, Edward A. (Berkeley, CA); Peterson, Jr., John P. (Livermore, CA)

    1985-01-01

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  5. Fast quench reactor and method

    DOE Patents [OSTI]

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

    1998-05-12

    A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

  6. Class 1 Permit Modification Notification Addition of Structures within Technical Area 54, Area G, Pad 11, Dome 375 Los Alamos National Laboratory Hazardous Waste Facility Permit, July 2012

    SciTech Connect (OSTI)

    Vigil-Holterman, Luciana R.; Lechel, Robert A.

    2012-08-31

    The purpose of this letter is to notify the New Mexico Environment Department-Hazardous Waste Bureau (NMED-HWB) of a Class 1 Permit Modification to the Los Alamos National Laboratory (LANL) Hazardous Waste Facility Permit issued to the Department of Energy (DOE) and Los Alamos National Security, LLC (LANS) in November 2010. The modification adds structures to the container storage unit at Technical Area (TA) 54 Area G, Pad 11. Permit Section 3.1(3) requires that changes to the location of a structure that does not manage hazardous waste shall be changed within the Permit as a Class 1 modification without prior approval in accordance with Code of Federal Regulations, Title 40 (40 CFR), {section}270.42(a)(1). Structures have been added within Dome 375 located at TA-54, Area G, Pad 11 that will be used in support of waste management operations within Dome 375 and the modular panel containment structure located within Dome 375, but will not be used as waste management structures. The Class 1 Permit Modification revises Figure 36 in Attachment N, Figures; and Figure G.12-1 in Attachment G.12, Technical Area 54, Area G, Pad 11 Outdoor Container Storage Unit Closure Plan. Descriptions of the structures have also been added to Section A.4.2.9 in Attachment A, TA - Unit Descriptions; and Section 2.0 in Attachment G.12, Technical Area 54, Area G, Pad 11 Outdoor Container Storage Unit Closure Plan. Full description of the permit modification and the necessary changes are included in Enclosure 1. The modification has been prepared in accordance with 40 CFR {section}270.42(a)(l). This package includes this letter and an enclosure containing a description of the permit modification, text edits of the Permit sections, and the revised figures (collectively LA-UR-12-22808). Accordingly, a signed certification page is also enclosed. Three hard copies and one electronic copy of this submittal will be delivered to the NMED-HWB.

  7. Small Modular Nuclear Reactors: Parametric Modeling of Integrated...

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

    1 Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ... PDF icon Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ...

  8. Neutrino oscillation studies with reactors

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

    Vogel, P.; Wen, L.J.; Zhang, C.

    2015-04-27

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavours are quantum mechanical mixtures. Over the past several decades, reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle θ13. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  9. Daya Bay Reactor Neutrino Experiment

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

    Ao Nuclear Power Plant reactors. The experiment is being built by blasting three kilometers of tunnel through the granite rock under the mountains where the power plants are...

  10. Reactor Materials Newsletter- Issue 1

    Broader source: Energy.gov [DOE]

    The Reactor Materials (RM) newsletter includes information about key nuclear materials programs, results from ongoing projects across the Office of Nuclear Energy, and other relevant information.

  11. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-11-01

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  12. Nuclear reactor fuel element

    DOE Patents [OSTI]

    Johnson, Carl E. (Elk Grove, IL); Crouthamel, Carl E. (Richland, WA)

    1980-01-01

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

  13. Reactor coolant pump flywheel

    DOE Patents [OSTI]

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  14. High flux reactor

    DOE Patents [OSTI]

    Lake, James A.; Heath, Russell L.; Liebenthal, John L.; DeBoisblanc, Deslonde R.; Leyse, Carl F.; Parsons, Kent; Ryskamp, John M.; Wadkins, Robert P.; Harker, Yale D.; Fillmore, Gary N.; Oh, Chang H.

    1988-01-01

    A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

  15. In-Reactor Experiment

    Office of Environmental Management (EM)

    Update on the TMIST-3 In-Reactor Experiment Tritium Release and Speciation from LiAlO 2 and LiAlO 2 /Zr Cermets DJ SENOR 1 , WG LUSCHER 1 , AND KK CLAYTON 2 1 Pacific Northwest National Laboratory, 2 Idaho National Laboratory Tritium Focus Group Meeting, Princeton, NJ 5 May 2015 1 PNNL-SA-109847 Tritium Production Enterprise: Background Tritium is required for US nuclear weapons stockpile Tritium has a 12.3 year half-life and must be replenished 1988: DOE ceased production of tritium at SRS

  16. Process and system for removing impurities from a gas

    DOE Patents [OSTI]

    Henningsen, Gunnar; Knowlton, Teddy Merrill; Findlay, John George; Schlather, Jerry Neal; Turk, Brian S

    2014-04-15

    A fluidized reactor system for removing impurities from a gas and an associated process are provided. The system includes a fluidized absorber for contacting a feed gas with a sorbent stream to reduce the impurity content of the feed gas; a fluidized solids regenerator for contacting an impurity loaded sorbent stream with a regeneration gas to reduce the impurity content of the sorbent stream; a first non-mechanical gas seal forming solids transfer device adapted to receive an impurity loaded sorbent stream from the absorber and transport the impurity loaded sorbent stream to the regenerator at a controllable flow rate in response to an aeration gas; and a second non-mechanical gas seal forming solids transfer device adapted to receive a sorbent stream of reduced impurity content from the regenerator and transfer the sorbent stream of reduced impurity content to the absorber without changing the flow rate of the sorbent stream.

  17. Method of making thermally removable epoxies

    DOE Patents [OSTI]

    Loy, Douglas A. (Albuquerque, NM); Wheeler, David R. (Albuquerque, NM); Russick, Edward M. (Rio Rancho, NM); McElhanon, James R. (Albuquerque, NM); Saunders, Randall S. (late of Albuquerque, NM)

    2002-01-01

    A method of making a thermally-removable epoxy by mixing a bis(maleimide) compound to a monomeric furan compound containing an oxirane group to form a di-epoxy mixture and then adding a curing agent at temperatures from approximately room temperature to less than approximately 90.degree. C. to form a thermally-removable epoxy. The thermally-removable epoxy can be easily removed within approximately an hour by heating to temperatures greater than approximately 90.degree. C. in a polar solvent. The epoxy material can be used in protecting electronic components that may require subsequent removal of the solid material for component repair, modification or quality control.

  18. Stationary Liquid Fuel Fast Reactor

    SciTech Connect (OSTI)

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

    2015-09-30

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

  19. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    Broader source: Energy.gov [DOE]

    Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary November 2014

  20. Foreign Research Reactor/Domestic Research Reactor Receipt Coordinator,

    National Nuclear Security Administration (NNSA)

    Savannah River Nuclear Solutions | National Nuclear Security Administration Foreign Research Reactor/Domestic Research Reactor Receipt Coordinator, Savannah River Nuclear Solutions | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our

  1. Engineering problems of tandem-mirror reactors

    SciTech Connect (OSTI)

    Moir, R.W.; Barr, W.L.; Boghosian, B.M.

    1981-10-22

    We have completed a comparative evaluation of several end plug configurations for tandem mirror fusion reactors with thermal barriers. The axi-cell configuration has been selected for further study and will be the basis for a detailed conceptual design study to be carried out over the next two years. The axi-cell end plug has a simple mirror cell produced by two circular coils followed by a transition coil and a yin-yang pair, which provides for MHD stability. This paper discusses some of the many engineering problems facing the designer. We estimated the direct cost to be 2$/W/sub e/. Assuming total (direct and indirect) costs to be twice this number, we need to reduce total costs by factors between 1.7 and 2.3 to compete with future LWRs levelized cost of electricity. These reductions may be possible by designing magnets producing over 20T made possible by use of combinations of superconducting and normal conducting coils as well as improvements in performance and cost of neutral beam and microwave power systems. Scientific and technological understanding and innovation are needed in the area of thermal barrier pumping - a process by which unwanted particles are removed (pumped) from certain regions of velocity and real space in the end plug. Removal of exhaust fuel ions, fusion ash and impurities by action of a halo plasma and plasma dump in the mirror end region is another challenging engineering problem discussed in this paper.

  2. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect (OSTI)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of inherent safety concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

  3. Melter Glass Removal and Dismantlement

    SciTech Connect (OSTI)

    Richardson, BS

    2000-10-31

    The U.S. Department of Energy (DOE) has been using vitrification processes to convert high-level radioactive waste forms into a stable glass for disposal in waste repositories. Vitrification facilities at the Savannah River Site (SRS) and at the West Valley Demonstration Project (WVDP) are converting liquid high-level waste (HLW) by combining it with a glass-forming media to form a borosilicate glass, which will ensure safe long-term storage. Large, slurry fed melters, which are used for this process, were anticipated to have a finite life (on the order of two to three years) at which time they would have to be replaced using remote methods because of the high radiation fields. In actuality the melters useable life spans have, to date, exceeded original life-span estimates. Initial plans called for the removal of failed melters by placing the melter assembly into a container and storing the assembly in a concrete vault on the vitrification plant site pending size-reduction, segregation, containerization, and shipment to appropriate storage facilities. Separate facilities for the processing of the failed melters currently do not exist. Options for handling these melters include (1) locating a facility to conduct the size-reduction, characterization, and containerization as originally planned; (2) long-term storing or disposing of the complete melter assembly; and (3) attempting to refurbish the melter and to reuse the melter assembly. The focus of this report is to look at methods and issues pertinent to size-reduction and/or melter refurbishment in particular, removing the glass as a part of a refurbishment or to reduce contamination levels (thus allowing for disposal of a greater proportion of the melter as low level waste).

  4. Design and performance of a high-pressure Fischer-Tropsch fluidized bed reactor

    SciTech Connect (OSTI)

    Weimer, A.W.; Quarderer, G.J.; Cochran, G.A.; Conway, M.M. )

    1988-01-01

    A 900 kg/day, CO/H/sub 2/, high-pressure, fluidized bed, pilot reactor was designed from first principles to achieve high reactant conversions and heat removal rates for the Fischer-Tropsch (F-T) synthesis of liquefied petroleum gases (LPG's). Suppressed bubble growth at high pressure allowed high reactant conversions which nearly matched those obtained at identical conditions in a lab scale fixed bed reactor. For GHSV approximately 1400 hr/sup -1/ and T = 658 {Kappa} at P approximately 7000 {kappa}Pa, reactant conversion exceeded 75%. The reactor heat removal capability exceeded twice design performance with the fluidized bed easily operating under thermally stable conditions. The fluidized catalyst was a potassium promoted, molybdenum on carbon (Mo/{Kappa}/C) catalyst which did not produce any detrimental waxy products. Long catalyst lifetimes of 1000 hrs on steam between regenerations allowed the fluidized bed to be operated in a batch mode.

  5. Chemistry control and corrosion mitigation of heat transfer salts for the fluoride salt reactor (FHR)

    SciTech Connect (OSTI)

    Kelleher, B. C.; Sellers, S. R.; Anderson, M. H.; Sridharan, K.; Scheele, R. D.

    2012-07-01

    The Molten Salt Reactor Experiment (MSRE) was a prototype nuclear reactor which operated from 1965 to 1969 at Oak Ridge National Laboratory. The MSRE used liquid fluoride salts as a heat transfer fluid and solvent for fluoride based {sup 235}U and {sup 233}U fuel. Extensive research was performed in order to optimize the removal of oxide and metal impurities from the reactor's heat transfer salt, 2LiF-BeF{sub 2} (FLiBe). This was done by sparging a mixture of anhydrous hydrofluoric acid and hydrogen gas through the FLiBe at elevated temperatures. The hydrofluoric acid reacted with oxides and hydroxides, fluorinating them while simultaneously releasing water vapor. Metal impurities such as iron and chromium were reduced by hydrogen gas and filtered out of the salt. By removing these impurities, the corrosion of reactor components was minimized. The Univ. of Wisconsin - Madison is currently researching a new chemical purification process for fluoride salts that make use of a less dangerous cleaning gas, nitrogen trifluoride. Nitrogen trifluoride has been predicted as a superior fluorinating agent for fluoride salts. These purified salts will subsequently be used for static and loop corrosion tests on a variety of reactor materials to ensure materials compatibility for the new FHR designs. Demonstration of chemistry control methodologies along with potential reduction in corrosion is essential for the use of a fluoride salts in a next generator nuclear reactor system. (authors)

  6. Fast reactors and nuclear nonproliferation

    SciTech Connect (OSTI)

    Avrorin, E.N.; Rachkov, V.I.; Chebeskov, A.N.

    2013-07-01

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  7. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, R.M.

    1983-11-08

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  8. Reactor Engineering Design | netl.doe.gov

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

    Reactor Engineering Design The Reactor Engineering Design Key Technology will focus on control of chemical reactions with unprecedented precision in increasingly modular and efficient reactors, allowing for smaller reactors and streamlined processes that will convert coal into valuable products at low cost and with high energy efficiency. Here, the specific emphasis will be reactors enabling conversion of coal-biomass to liquid fuels, Novel reactors, advanced manufacturing, etc. will be

  9. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, Robert M.

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  10. Remove Condensate with Minimal Air Loss | Department of Energy

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

    Remove Condensate with Minimal Air Loss Remove Condensate with Minimal Air Loss This tip sheet outlines several condensate removal methods as part of maintaining compressed air ...

  11. MOOSE simulating nuclear reactor CRUD buildup

    ScienceCinema (OSTI)

    None

    2014-07-21

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  12. MOOSE simulating nuclear reactor CRUD buildup

    SciTech Connect (OSTI)

    2014-02-06

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  13. Energy Department Announces Small Modular Reactor Technology...

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

    ... of nuclear reactors by providing more than 200 million through a cost-share agreement to support the licensing reviews for Westinghouse's AP1000 reactor design certification. ...

  14. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

    Howard, D.F.; Motta, E.E.

    1961-06-27

    A method for controlling the excess reactivity in a nuclear reactor throughout the core life while maintaining the neutron flux distribution at the desired level is described. The control unit embodies a container having two electrodes of different surface area immersed in an electrolytic solution of a good neutron sbsorbing metal ion such as boron, gadolinium, or cadmium. Initially, the neutron absorber is plated on the larger electrode to control the greater neutron flux of a freshly refueled core. As the fuel burns up, the excess reactivity decreases and the neutron absorber is then plated onto the smaller electrode so that the number of neutrons absorbed also decreases. The excess reactivity in the core may thus be maintained without the introduction of serious perturbations in the neutron flux distributibn.

  15. GAS COOLED NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Rodwell, W.

    1958-06-10

    A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

  16. Nuclear reactor control

    DOE Patents [OSTI]

    Cawley, William E. (Phoenix, AZ); Warnick, Robert F. (Pasco, WA)

    1982-01-01

    1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

  17. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-10-25

    Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

  18. Preliminary safety evaluation of the advanced burner test reactor.

    SciTech Connect (OSTI)

    Dunn, F. E.; Fanning, T. H.; Cahalan, J. E.; Nuclear Engineering Division

    2006-09-15

    Results of a preliminary safety evaluation of the Advanced Burner Test Reactor (ABTR) pre-conceptual design are reported. The ABTR safety design approach is described. Traditional defense-in-depth design features are supplemented with passive safety performance characteristics that include natural circulation emergency decay heat removal and reactor power reduction by inherent reactivity feedbacks in accidents. ABTR safety performance in design-basis and beyond-design-basis accident sequences is estimated based on analyses. Modeling assumptions and input data for safety analyses are presented. Analysis results for simulation of simultaneous loss of coolant pumping power and normal heat rejection are presented and discussed, both for the case with reactor scram and the case without reactor scram. The analysis results indicate that the ABTR pre-conceptual design is capable of undergoing bounding design-basis and beyond-design-basis accidents without fuel cladding failures. The first line of defense for protection of the public against release of radioactivity in accidents remains intact with significant margin. A comparison and evaluation of general safety design criteria for the ABTR conceptual design phase are presented in an appendix. A second appendix presents SASSYS-1 computer code capabilities and modeling enhancements implemented for ABTR analyses.

  19. Plutonium Consumption Program, CANDU Reactor Project final report

    SciTech Connect (OSTI)

    Not Available

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  20. Diamond Wire Cutting of the Tokamak Fusion Test Reactor

    SciTech Connect (OSTI)

    Keith Rule; Erik Perry; Robert Parsells

    2003-01-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D (Decontamination and Decommissioning) activity.

  1. DIAMOND WIRE CUTTING OF THE TOKAMAK FUSION TEST REACTOR

    SciTech Connect (OSTI)

    Rule, Keith; Perry, Erik; Parsells, Robert

    2003-02-27

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the Tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the techno logy was improved and redesigned for the actual cutting of the vacuum vessel. 10 complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D activity.

  2. Removal of metal ions from aqueous solution

    DOE Patents [OSTI]

    Jackson, Paul J. (both Los Alamos, NM); Delhaize, Emmanuel (both Los Alamos, NM); Robinson, Nigel J. (Durham, GB2); Unkefer, Clifford J. (Los Alamos, NM); Furlong, Clement (Seattle, WA)

    1990-11-13

    A method of removing heavy metals from aqueous solution, a composition of matter used in effecting said removal, and apparatus used in effecting said removal. One or more of the polypeptides, poly (.gamma.-glutamylcysteinyl)glycines, is immobilized on an inert material in particulate form. Upon contact with an aqueous solution containing heavy metals, the polypeptides sequester the metals, removing them from the solution. There is selectivity of poly (.gamma.-glutamylcysteinyl)glycines having a particular number of monomer repeat unit for particular metals. The polypeptides are easily regenerated by contact with a small amount of an organic acid, so that they can be used again to remove heayv metals from solution. This also results in the removal of the metals from the column in a concentrated form.

  3. Removal of metal ions from aqueous solution

    DOE Patents [OSTI]

    Jackson, Paul J. (Los Alamos, NM); Delhaize, Emmanuel (Los Alamos, NM); Robinson, Nigel J. (Durham, GB2); Unkefer, Clifford J. (Los Alamos, NM); Furlong, Clement (Seattle, WA)

    1990-01-01

    A method of removing heavy metals from aqueous solution, a composition of matter used in effecting said removal, and apparatus used in effecting said removal. One or more of the polypeptides, poly (.gamma.-glutamylcysteinyl)glycines, is immobilized on an inert material in particulate form. Upon contact with an aqueous solution containing heavy metals, the polypeptides sequester the metals, removing them from the solution. There is selectivity of poly (.gamma.-glutamylcysteinyl)glycines having a particular number of monomer repeat units for particular metals. The polypeptides are easily regenerated by contact with a small amount of an organic acid, so that they can be used again to remove heavy metals from solution. This also results in the removal of the metals from the column in a concentrated form.

  4. Cooling molten salt reactors using gas-lift

    SciTech Connect (OSTI)

    Zitek, Pavel E-mail: klimko@kke.zcu.cz; Valenta, Vaclav E-mail: klimko@kke.zcu.cz; Klimko, Marek E-mail: klimko@kke.zcu.cz

    2014-08-06

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a Two-phase flow demonstrator (TFD) used for experimental study of the gas-lift system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for gas-lift (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

  5. Hydrothermal Processing of Macroalgal Feedstocks in Continuous-Flow Reactors

    SciTech Connect (OSTI)

    Elliott, Douglas C.; Hart, Todd R.; Neuenschwander, Gary G.; Rotness, Leslie J.; Roesijadi, Guritno; Zacher, Alan H.; Magnuson, Jon K.

    2014-02-18

    Wet macroalgal slurries can be converted into a biocrude by hydrothermal liquefaction (HTL). High levels of carbon conversion to gravity-separable oil product were accomplished at relatively low temperature (350 ?C) in a pressurized (sub-critical liquid water) environment (20 MPa). As opposed to earlier work in batch reactors reported by others, direct oil recovery was achieved without the use of a solvent and biomass trace mineral components were removed by processing steps so that they did not cause processing difficulties. In addition, catalytic hydrothermal gasification was effectively applied for HTL byproduct water cleanup and fuel gas production from water soluble organics. As a result, high conversion of macroalgae to liquid and gas fuel products was found with low levels of organic contamination in byproduct water. Both process steps were accomplished in continuous-flow reactor systems such that design data for process scale-up was generated.

  6. Retrievable fuel pin end member for a nuclear reactor

    DOE Patents [OSTI]

    Rosa, Jerry M. (Los Gatos, CA)

    1982-01-01

    A bottom end member (17b) on a retrievable fuel pin (13b) secures the pin (13b) within a nuclear reactor (12) by engaging on a transverse attachment rail (18) with a spring clip type of action. Removal and reinstallation if facilitated as only axial movement of the fuel pin (13b) is required for either operation. A pair of resilient axially extending blades (31) are spaced apart to define a slot (24) having a seat region (34) which receives the rail (18) and having a land region (37), closer to the tips (39) of the blades (31) which is normally of less width than the rail (18). Thus an axially directed force sufficient to wedge the resilient blades (31) apart is required to emplace or release the fuel pin (13b) such force being greater than the axial forces on the fuel pins (13b) which occur during operation of the reactor (12).

  7. Protection #2: Trap and Remove Sediment

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

    Trap and Remove Sediment Protection #2: Trap and Remove Sediment The 3 Protections = Defense in Depth August 1, 2013 Sediment behind LA Canyon weir is sampled and excavated regularly. As of 2012, no sediment required disposal as hazardous or radioactive waste. Sediment behind LA Canyon weir is sampled and excavated regularly. As of 2012, no sediment required disposal as hazardous or radioactive waste. The 3 Protections Protection #1: Remove the source of contamination Protection #2: Stabilize,

  8. Method for removing contaminants from plastic resin

    DOE Patents [OSTI]

    Bohnert, George W. (Harrisonville, MO); Hand, Thomas E. (Lee's Summit, MO); DeLaurentiis, Gary M. (Jamestown, CA)

    2008-12-09

    A resin recycling method that produces essentially contaminant-free synthetic resin material in an environmentally safe and economical manner. The method includes receiving the resin in container form. The containers are then ground into resin particles. The particles are exposed to a solvent, the solvent contacting the resin particles and substantially removing contaminants on the resin particles. After separating the particles and the resin, a solvent removing agent is used to remove any residual solvent remaining on the resin particles after separation.

  9. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOE Patents [OSTI]

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  10. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOE Patents [OSTI]

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  11. System for removing contaminants from plastic resin

    DOE Patents [OSTI]

    Bohnert, George W. (Harrisonville, MO); Hand, Thomas E. (Lee's Summit, MO); DeLaurentiis, Gary M. (Jamestown, CA)

    2010-11-23

    A resin recycling system that produces essentially contaminant-free synthetic resin material in an environmentally safe and economical manner. The system includes receiving the resin in container form. A grinder grinds the containers into resin particles. The particles are exposed to a solvent in one or more solvent wash vessels, the solvent contacting the resin particles and substantially removing contaminants on the resin particles. A separator is used to separate the resin particles and the solvent. The resin particles are then placed in solvent removing element where they are exposed to a solvent removing agent which removes any residual solvent remaining on the resin particles after separation.

  12. Slag capture and removal during laser cutting

    DOE Patents [OSTI]

    Brown, Clyde O. (Newington, CT)

    1984-05-08

    Molten metal removed from a workpiece in a laser cutting operation is blown away from the cutting point by a gas jet and collected on an electromagnet.

  13. Turkey HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Turkey HEU Removal | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the...

  14. Removal of radioisotopes from waste solutions

    DOE Patents [OSTI]

    Kirby, H.W.

    1973-10-01

    The invention comprises removing radioisotopes from waste liquids or solutions by passing these through filters and through a column containing a suitable salt of phosphoric acid. (Official Gazette)

  15. General Counsel Legal Interpretation Regarding Medical Removal...

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

    Counsel Legal Interpretation Regarding Medical Removal Protection Benefits Pursuant to 10 CFR Part 850, Chronic Beryllium Disease Prevention Program General Counsel Legal ...

  16. Install Removable Insulation on Valves and Fittings

    Broader source: Energy.gov [DOE]

    This tip sheet on installing removable insulation on valves and fittings provides how-to advice for improving steam systems using low-cost, proven practices and technologies.

  17. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    SciTech Connect (OSTI)

    Douglas Morrell

    2011-03-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  18. Fueling of tandem mirror reactors

    SciTech Connect (OSTI)

    Gorker, G.E.; Logan, B.G.

    1985-01-01

    This paper summarizes the fueling requirements for experimental and demonstration tandem mirror reactors (TMRs), reviews the status of conventional pellet injectors, and identifies some candidate accelerators that may be needed for fueling tandem mirror reactors. Characteristics and limitations of three types of accelerators are described; neutral beam injectors, electromagnetic rail guns, and laser beam drivers. Based on these characteristics and limitations, a computer module was developed for the Tandem Mirror Reactor Systems Code (TMRSC) to select the pellet injector/accelerator combination which most nearly satisfies the fueling requirements for a given machine design.

  19. Nuclear reactor downcomer flow deflector

    DOE Patents [OSTI]

    Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

    2011-02-15

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  20. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    SciTech Connect (OSTI)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2 discharge reuse. The EM2 waste disposal profile is effectively only fission products, which reduces the mass (about 3% vs LWR), average half life, heat and long term radio-toxicity of the disposal. Widespread implementation of EM2 fuel cycle is highly significant as it would increase world energy reserves; the remaining energy in U.S. LWR SNF alone exceeds that in the U.S. natural gas reserves. Unlike many LWR SNF disposition concepts, the EM2 fuel cycle conversion of SNF produces energy and associated revenue such that the overall project is cost effective. By providing conversion of SNF to fission products the fuel cycle is closed and a non-repository LWR SNF disposition path is created and overall repository requirements are significantly reduced. (authors)

  1. METHOD OF CONTROLLING CORROSION IN A NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Kidder, C.P.; Sloan, C.K.

    1959-10-01

    A method is described for reducing or removing corrosion and iron deposits on aluminum surfaces from coolant water comprising adding to the coolant alkali metal dichromate in a concentration of between 1.8 and 2.2 ppm, adjusting the pH to between 7.3 and 7.8 by adding CaCO/sub 3/ or other similar material, and adding a silicious material such as diatomaceous earth of a particle size of 5 to 15 microns to effect a suspension of between 2 and 300 ppm and circulating it through the reactor.

  2. METHOD OF OPERATING A HEAVY WATER MODERATED REACTOR

    DOE Patents [OSTI]

    Vernon, H.C.

    1962-08-14

    A method of removing fission products from the heavy water used in a slurry type nuclear reactor is described. According to the process the slurry is steam distilled with carbon tetrachloride so that at least a part of the heavy water and carbon tetrachloride are vaporized; the heavy water and carbon tetrachloride are separated; the carbon tetrachloride is returned to the steam distillation column at different points in the column to aid in depositing the slurry particles at the bottom of the column; and the heavy water portion of the condensate is purified. (AEC)

  3. Thermal-Hydraulic Analyses of Transients in an Actinide-Burner Reactor Cooled by Forced Convection of Lead Bismuth

    SciTech Connect (OSTI)

    Davis, Cliff Bybee

    2003-09-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or leadbismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed.

  4. Packed-Bed Reactor Study of NETL Sample 196c for the Removal...

    Office of Scientific and Technical Information (OSTI)

    Authors: Hoffman, James S. ; Hammache, Sonia ; Gray, McMahan L. ; Fauth Daniel J. ; Pennline, Henry W. Publication Date: 2012-04-24 OSTI Identifier: 1044172 Report Number(s): ...

  5. Removal of 1,082-Ton Reactor Among Richland Operations Office’s 2014 Accomplishments

    Broader source: Energy.gov [DOE]

    RICHLAND, Wash. – Workers with EM’s Richland Operations Office and its contractors made progress this year in several areas of Hanford site cleanup that helped protect employees, the public, environment, and Columbia River.

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

    SciTech Connect (OSTI)

    Prechtl, E.; Suessdorf, W.

    2003-02-27

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

  7. 2012 Annual Report Research Reactor Infrastructure Program

    SciTech Connect (OSTI)

    Douglas Morrell

    2012-11-01

    The content of this report is the 2012 Annual Report for the Research Reactor Infrastructure Program.

  8. METHOD OF OPERATING NUCLEAR REACTORS

    DOE Patents [OSTI]

    Untermyer, S.

    1958-10-14

    A method is presented for obtaining enhanced utilization of natural uranium in heavy water moderated nuclear reactors by charging the reactor with an equal number of fuel elements formed of natural uranium and of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction. The reactor is operated until the rate of burnup of plutonium equals its rate of production, the fuel elements are processed to recover plutonium, the depleted uranium is discarded, and the remaining uranium is formed into fuel elements. These fuel elements are charged into a reactor along with an equal number of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction, and reuse of the uranium is continued as aforesaid until it wlll no longer support a chain reaction when combined with an equal quantity of natural uranium.

  9. FUEL ELEMENTS FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

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

    1963-05-01

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

  10. Combustion synthesis continuous flow reactor

    DOE Patents [OSTI]

    Maupin, Gary D. (Richland, WA); Chick, Lawrence A. (West Richland, WA); Kurosky, Randal P. (Maple Valley, WA)

    1998-01-01

    The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor.

  11. Small modular reactors (SMRs) such...

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

    Small modular reactors (SMRs) such as the one illustrated in Figure 1 are being considered by the commercial nuclear power industry as an option for more distributed generation and...

  12. Nuclear Reactors and Technology; (USA)

    SciTech Connect (OSTI)

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

    1991-01-01

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

  13. Combustion synthesis continuous flow reactor

    DOE Patents [OSTI]

    Maupin, G.D.; Chick, L.A.; Kurosky, R.P.

    1998-01-06

    The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor. 10 figs.

  14. Method of removing contaminants from plastic resins

    DOE Patents [OSTI]

    Bohnert, George W. (Harrisonville, MO); Hand, Thomas E. (Lee's Summit, MO); DeLaurentiis, Gary M. (Jamestown, CA)

    2008-11-18

    A method for removing contaminants from synthetic resin material containers using a first organic solvent system and a second carbon dioxide system. The organic solvent is utilized for removing the contaminants from the synthetic resin material and the carbon dioxide is used to separate any residual organic solvent from the synthetic resin material.

  15. Method of removing contaminants from plastic resins

    DOE Patents [OSTI]

    Bohnert,George W. (Harrisonville, MO); Hand,Thomas E. (Lee's Summit, MO); Delaurentiis,Gary M. (Jamestown, CA)

    2007-08-07

    A method for removing contaminants from synthetic resin material containers using a first organic solvent system and a second carbon dioxide system. The organic solvent is utilized for removing the contaminants from the synthetic resin material and the carbon dioxide is used to separate any residual organic solvent from the synthetic resin material.

  16. Method for removing contaminants from plastic resin

    DOE Patents [OSTI]

    Bohnert, George W. (Harrisonville, MO); Hand, Thomas E. (Lee's Summit, MO); DeLaurentiis, Gary M. (Jamestown, CA)

    2008-12-30

    A method for removing contaminants from synthetic resin material containers using a first organic solvent system and a second carbon dioxide system. The organic solvent is utilized for removing the contaminants from the synthetic resin material and the carbon dioxide is used to separate any residual organic solvent from the synthetic resin material.

  17. Reactor shroud joint

    DOE Patents [OSTI]

    Ballas, Gary J. (San Jose, CA); Fife, Alex Blair (San Jose, CA); Ganz, Israel (San Jose, CA)

    1998-01-01

    A shroud for a nuclear reactor is described. In one embodiment, the shroud includes first and second shroud sections, and each shroud section includes a substantially cylindrical main body having a first end and a second end. With respect to each shroud section, a flange is located at the main body first end, and the flange has a plurality of bolt openings therein and a plurality of scalloped regions. The first shroud section is welded to the second shroud section, and at least some of the bolt openings in the first shroud section flange align with respective bolt openings in the second shroud section flange. In the event that the onset of inter-granular stress corrosion cracking is ever detected in the weld between the shroud section, bolts are inserted through bolt openings in the first shroud section flange and through aligned bolt openings the second shroud section flange. Each bolt, in one embodiment, has a shank section and first and second threaded end sections. Nuts are threadedly engaged to the threaded end sections and tightened against the respective flanges.

  18. Reactor shroud joint

    DOE Patents [OSTI]

    Ballas, G.J.; Fife, A.B.; Ganz, I.

    1998-04-07

    A shroud for a nuclear reactor is described. In one embodiment, the shroud includes first and second shroud sections, and each shroud section includes a substantially cylindrical main body having a first end and a second end. With respect to each shroud section, a flange is located at the main body first end, and the flange has a plurality of bolt openings therein and a plurality of scalloped regions. The first shroud section is welded to the second shroud section, and at least some of the bolt openings in the first shroud section flange align with respective bolt openings in the second shroud section flange. In the event that the onset of inter-granular stress corrosion cracking is ever detected in the weld between the shroud section, bolts are inserted through bolt openings in the first shroud section flange and through aligned bolt openings the second shroud section flange. Each bolt, in one embodiment, has a shank section and first and second threaded end sections. Nuts are threadedly engaged to the threaded end sections and tightened against the respective flanges. 4 figs.

  19. Reactor pressure vessel nozzle

    DOE Patents [OSTI]

    Challberg, Roy C. (Livermore, CA); Upton, Hubert A. (Morgan Hill, CA)

    1994-01-01

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough.

  20. Solar solids reactor

    DOE Patents [OSTI]

    Yudow, B.D.

    1986-02-24

    A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

  1. Solar solids reactor

    DOE Patents [OSTI]

    Yudow, Bernard D. (Chicago, IL)

    1987-01-01

    A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

  2. Reactor pressure vessel nozzle

    DOE Patents [OSTI]

    Challberg, R.C.; Upton, H.A.

    1994-10-04

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. 2 figs.

  3. 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.; Massaro, Lawrence M.; Jensen, Philip J.

    2014-10-01

    This report presents a preliminary evaluation of removing used nuclear fuel (UNF) from 12 shutdown nuclear power plant sites. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. The evaluation was divided into four components: characterization of the UNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory; a description of the on-site infrastructure and conditions relevant to transportation of UNF and GTCC waste; an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing UNF and GTCC waste, including identification of gaps in information; and, an evaluation of the actions necessary to prepare for and remove UNF and GTCC waste. The primary sources for the inventory of UNF and GTCC waste are the U.S. Department of Energy (DOE) RW-859 used nuclear fuel inventory database, industry sources such as StoreFUEL and SpentFUEL, and government sources such as the U.S. Nuclear Regulatory Commission. The primary sources for information on the conditions of site and near-site transportation infrastructure and experience included observations and information collected during visits to the Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion sites; information provided by managers at the shutdown sites; Facility Interface Data Sheets compiled for DOE in 2005; Services Planning Documents prepared for DOE in 1993 and 1994; industry publications such as Radwaste Solutions; and Google Earth. State and Regional Group representatives, a Tribal representative, and a Federal Railroad Administration representative participated in six of the shutdown site visits. Every site was found to have at least one off-site transportation mode option for removing its UNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 12 shutdown sites use designs from 4 different suppliers involving 9 different (horizontal and vertical) dry storage systems that would require the use of 8 different transportation cask designs to remove the UNF and GTCC waste from the shutdown sites; Although there are common aspects, each site has some unique features and/or conditions; Although some regulatory actions will be required, all UNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in licensed systems that can be transported, including a small amount of high-burnup fuel; Each site indicated that 2-3 years of advance time would be required for its preparations before shipments could begin; Most sites have more than one transportation option, e.g., rail, barge, or heavy haul truck, as well as constraints and preferences. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

  4. Small Modular Reactors (SMRs) | Department of Energy

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

    Reactor Technologies » Small Modular Reactors (SMRs) Small Modular Reactors (SMRs) NuScale Power Reactors. ©NuScale Power, LLC, All Rights Reserved NuScale Power Reactors. ©NuScale Power, LLC, All Rights Reserved Small Modular Reactors (SMRs) are nuclear power plants that are smaller in size (300 MWe or less) than current generation base load plants (1,000 MWe or higher). These smaller, compact designs are factory-fabricated reactors that can be transported by truck or rail to a nuclear

  5. Fast quench reactor and method

    DOE Patents [OSTI]

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

    1998-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

  6. Fast quench reactor and method

    DOE Patents [OSTI]

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

    2002-09-24

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

  7. Fast quench reactor and method

    DOE Patents [OSTI]

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

    2002-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

  8. Alternate-fuel reactor studies

    SciTech Connect (OSTI)

    Evans, K. Jr.; Ehst, D.A.; Gohar, Y.; Jung, J.; Mattas, R.F.; Turner, L.R.

    1983-02-01

    A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a /sup 3/He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding.

  9. Automatic safety rod for reactors

    DOE Patents [OSTI]

    Germer, John H. (San Jose, CA)

    1988-01-01

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  10. Biological sludge stabilization reactor evaluations

    SciTech Connect (OSTI)

    Corbitt, R.A.; Bowen, P.T.; Smith, P.E.

    1998-07-01

    Anaerobic digestion was chosen as the means to stabilize primary and thickened waste activated sludge for a 0.88 m{sup 3}/s (20 mgd) advanced wastewater reclamation facility. Two stage digestion was proposed to produce Class B sludge. Reactor shape was an important variable in design of the first stage digestion. Evaluation of conventional and egg shaped anaerobic digesters was performed. Based on the economic and non-economic criteria analysis, egg shaped reactors were selected.

  11. Method And Apparatus For Arbitrarily Large Capacity Removable Media

    DOE Patents [OSTI]

    Milligan, Charles A. (Golden, CO); Hughes, James P. (Lino Lakes, MN); Debiez; Jacques (Cugnaux, FR)

    2003-04-08

    A method and apparatus to handle multiple sets of removable media within a storage system. A first set of removable media are mounted on a set of drives. Data is accepted until the first set of removable media is filled. A second set of removable media is mounted on the drives, while the first set of removable media is removed. When the change in removable media is complete, writing of data proceeds on the second set of removable media. Data may be buffered while the change in removable media occurs. Alternatively, two sets of removable media may be mounted at the same time. When the first set of removable media is filled to a selected amount, the second set of removable media may then be used to write the data. A third set of removable media is set up or mounted for use, while the first set of removable media is removed.

  12. Catalyst regeneration process including metal contaminants removal

    DOE Patents [OSTI]

    Ganguli, Partha S. (Lawrenceville, NJ)

    1984-01-01

    Spent catalysts removed from a catalytic hydrogenation process for hydrocarbon feedstocks, and containing undesired metals contaminants deposits, are regenerated. Following solvent washing to remove process oils, the catalyst is treated either with chemicals which form sulfate or oxysulfate compounds with the metals contaminants, or with acids which remove the metal contaminants, such as 5-50 W % sulfuric acid in aqueous solution and 0-10 W % ammonium ion solutions to substantially remove the metals deposits. The acid treating occurs within the temperature range of 60.degree.-250.degree. F. for 5-120 minutes at substantially atmospheric pressure. Carbon deposits are removed from the treated catalyst by carbon burnoff at 800.degree.-900.degree. F. temperature, using 1-6 V % oxygen in an inert gas mixture, after which the regenerated catalyst can be effectively reused in the catalytic process.

  13. Test Plan for the overburden removal demonstration

    SciTech Connect (OSTI)

    Rice, P.; Thompson, D.; Winberg, M.; Skaggs, J.

    1993-06-01

    The removal of soil overburdens from contaminated pits and trenches involves using equipment that will remove a small layer of soil from 3 to 6 in. at any time. As a layer of soil is removed, overburden characterization techniques perform surveys to a depth that exceeds each overburden removal layer to ensure that the removed soil will be free of contamination. It is generally expected that no contamination will be found in the soil overburden, which was brought in after the waste was put in place. It is anticipated that some containers in the waste zone have lost their integrity, and the waste leakage from those containers has migrated by gravity downward into the waste zone. To maintain a safe work environment, this method of overburden removal should allow safe preparation of a pit or trench for final remediation. To demonstrate the soil overburden techniques, the Buried Waste Integrated Demonstration Program has contracted vendor services to provide equipment and techniques demonstrating soil overburden removal technology. The demonstration will include tests that will evaluate equipment performance and techniques for removal of overburden soil, control of contamination spread, and dust control. To evaluate the performance of these techniques, air particulate samples, physical measurements of the excavation soil cuts, maneuverability measurements, and time versus volume (rate) of soil removal data will be collected during removal operations. To provide a medium for sample evaluation, the overburden will be spiked at specific locations and depths with rare earth tracers. This test plan will be describe the objectives of the demonstration, data quality objectives, methods to be used to operate the equipment and use the techniques in the test area, and methods to be used in collecting data during the demonstration.

  14. A Gas-Cooled Reactor Surface Power System

    SciTech Connect (OSTI)

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  15. Propellant actuated nuclear reactor steam depressurization valve

    DOE Patents [OSTI]

    Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

    1992-01-01

    A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

  16. When Do Commercial Reactors Permanently Shut Down?

    Reports and Publications (EIA)

    2011-01-01

    For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

  17. Integrated process of distillation with side reactors for synthesis of organic acid esters

    DOE Patents [OSTI]

    Panchal, Chandrakant B; Prindle, John C; Kolah, Aspri; Miller, Dennis J; Lira, Carl T

    2015-11-04

    An integrated process and system for synthesis of organic-acid esters is provided. The method of synthesizing combines reaction and distillation where an organic acid and alcohol composition are passed through a distillation chamber having a plurality of zones. Side reactors are used for drawing off portions of the composition and then recycling them to the distillation column for further purification. Water is removed from a pre-reactor prior to insertion into the distillation column. An integrated heat integration system is contained within the distillation column for further purification and optimizing efficiency in the obtaining of the final product.

  18. Decomposition of chlorinated ethylenes and ethanes in an electron beam generated plasma reactor

    SciTech Connect (OSTI)

    Vitale, S.A.

    1996-02-01

    An electron beam generated plasma reactor (EBGPR) is used to determine the plasma chemistry kinetics, energetics and decomposition pathways of six chlorinated ethylenes and ethanes: 1,1,1-trichloroethane, 1,1-dichloroethane, ethyl chloride, trichloroethylene, 1,1-dichloroethylene, and vinyl chloride. A traditional chemical kinetic and chemical engineering analysis of the data from the EBGPR is performed, and the following hypothesis was verified: The specific energy required for chlorinated VOC decomposition in the electron beam generated plasma reactor is determined by the electron attachment coefficient of the VOC and the susceptibility of the molecule to radical attack. The technology was demonstrated at the Hanford Reservation to remove VOCs from soils.

  19. In situ removal of contamination from soil

    DOE Patents [OSTI]

    Lindgren, Eric R. (Albuquerque, NM); Brady, Patrick V. (Albuquerque, NM)

    1997-01-01

    A process of remediation of cationic heavy metal contamination from soil utilizes gas phase manipulation to inhibit biodegradation of a chelating agent that is used in an electrokinesis process to remove the contamination, and further gas phase manipulation to stimulate biodegradation of the chelating agent after the contamination has been removed. The process ensures that the chelating agent is not attacked by bioorganisms in the soil prior to removal of the contamination, and that the chelating agent does not remain as a new contaminant after the process is completed.

  20. In situ removal of contamination from soil

    DOE Patents [OSTI]

    Lindgren, E.R.; Brady, P.V.

    1997-10-14

    A process of remediation of cationic heavy metal contamination from soil utilizes gas phase manipulation to inhibit biodegradation of a chelating agent that is used in an electrokinesis process to remove the contamination. The process also uses further gas phase manipulation to stimulate biodegradation of the chelating agent after the contamination has been removed. The process ensures that the chelating agent is not attacked by bioorganisms in the soil prior to removal of the contamination, and that the chelating agent does not remain as a new contaminant after the process is completed. 5 figs.

  1. Method for changing removable bearing for a wind turbine generator

    DOE Patents [OSTI]

    Bagepalli, Bharat Sampathkumaran (Niskayuna, NY); Jansen, Patrick Lee (Scotia, NY); Gadre, Aniruddha Dattatraya (Rexford, NY)

    2008-04-22

    A wind generator having removable change-out bearings includes a rotor and a stator, locking bolts configured to lock the rotor and stator, a removable bearing sub-assembly having at least one shrunk-on bearing installed, and removable mounting bolts configured to engage the bearing sub-assembly and to allow the removable bearing sub-assembly to be removed when the removable mounting bolts are removed.

  2. Removable bearing arrangement for a wind turbine generator

    DOE Patents [OSTI]

    Bagepalli, Bharat Sampathkumaran; Jansen, Patrick Lee; Gadre, Aniruddha Dattatraya

    2010-06-15

    A wind generator having removable change-out bearings includes a rotor and a stator, locking bolts configured to lock the rotor and stator, a removable bearing sub-assembly having at least one shrunk-on bearing installed, and removable mounting bolts configured to engage the bearing sub-assembly and to allow the removable bearing sub-assembly to be removed when the removable mounting bolts are removed.

  3. Process and system for removing sulfur from sulfur-containing gaseous streams

    DOE Patents [OSTI]

    Basu, Arunabha; Meyer, Howard S.; Lynn, Scott; Leppin, Dennis; Wangerow, James R.

    2012-08-14

    A multi-stage UCSRP process and system for removal of sulfur from a gaseous stream in which the gaseous stream, which contains a first amount of H.sub.2S, is provided to a first stage UCSRP reactor vessel operating in an excess SO.sub.2 mode at a first amount of SO.sub.2, producing an effluent gas having a reduced amount of SO.sub.2, and in which the effluent gas is provided to a second stage UCSRP reactor vessel operating in an excess H.sub.2S mode, producing a product gas having an amount of H.sub.2S less than said first amount of H.sub.2S.

  4. International Research Reactor Decommissioning Project

    SciTech Connect (OSTI)

    Leopando, Leonardo; Warnecke, Ernst

    2008-01-15

    Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement to the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.

  5. Rapid starting methanol reactor system

    DOE Patents [OSTI]

    Chludzinski, Paul J. (38 Berkshire St., Swampscott, MA 01907); Dantowitz, Philip (39 Nancy Ave., Peabody, MA 01960); McElroy, James F. (12 Old Cart Rd., Hamilton, MA 01936)

    1984-01-01

    The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

  6. Method of removing polychlorinated biphenyl from oil

    DOE Patents [OSTI]

    Cook, G.T.; Holshouser, S.K.; Coleman, R.M.; Harless, C.E.; Whinnery, W.N. III

    1982-03-17

    Polychlorinated biphenyls are removed from oil by extracting the biphenyls into methanol. The mixture of methanol and extracted biphenyls is distilled to separate methanol therefrom, and the methanol is recycled for further use in extraction of biphenyls from oil.

  7. Method of removing polychlorinated biphenyl from oil

    DOE Patents [OSTI]

    Cook, Gus T. (Paducah, KY); Holshouser, Stephen K. (Boaz, KY); Coleman, Richard M. (Paducah, KY); Harless, Charles E. (Smithland, KY); Whinnery, III, Walter N. (Paducah, KY)

    1983-01-01

    Polychlorinated biphenyls are removed from oil by extracting the biphenyls into methanol. The mixture of methanol and extracted biphenyls is distilled to separate methanol therefrom, and the methanol is recycled for further use in extraction of biphenyls from oil.

  8. PIA - Advanced Test Reactor National Scientific User Facility...

    Office of Environmental Management (EM)

    Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor ...

  9. Small Modular Nuclear Reactors: Parametric Modeling of Integrated...

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

    2 Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ... This study presents a detailed analysis of the economics of Small Modular Reactors (SMRs), ...

  10. Advanced Nuclear Technology: Advanced Light Water Reactors Utility...

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

    Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary Advanced Nuclear Technology: Advanced Light Water Reactors ...

  11. GNEP Element:Develop Advanced Burner Reactors | Department of...

    Office of Environmental Management (EM)

    Develop Advanced Burner Reactors GNEP Element:Develop Advanced Burner Reactors An article describing burner reactors and the role in GNEP. PDF icon GNEP Element:Develop Advanced...

  12. Method for removing RFI from SAR images

    DOE Patents [OSTI]

    Doerry, Armin W.

    2003-08-19

    A method of removing RFI from a SAR by comparing two SAR images on a pixel by pixel basis and selecting the pixel with the lower magnitude to form a composite image. One SAR image is the conventional image produced by the SAR. The other image is created from phase-history data which has been filtered to have the frequency bands containing the RFI removed.

  13. Laser removal of sludge from steam generators

    DOE Patents [OSTI]

    Nachbar, Henry D. (Ballston Lake, NY)

    1990-01-01

    A method of removing unwanted chemical deposits known as sludge from the metal surfaces of steam generators with laser energy is provided. Laser energy of a certain power density, of a critical wavelength and frequency, is intermittently focused on the sludge deposits to vaporize them so that the surfaces are cleaned without affecting the metal surface (sludge substrate). Fiberoptic tubes are utilized for laser beam transmission and beam direction. Fiberoptics are also utilized to monitor laser operation and sludge removal.

  14. Accident Performance of Light Water Reactor Cladding Materials

    SciTech Connect (OSTI)

    Nelson, Andrew T.

    2012-07-24

    During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200 C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.

  15. Gas tagging and cover gas combination for nuclear reactor

    DOE Patents [OSTI]

    Gross, Kenny C. (Lemont, IL); Laug, Matthew T. (Idaho Falls, ID)

    1985-01-01

    The invention discloses the use of stable isotopes of neon and argon, that are grouped in preselected different ratios one to the other and are then sealed as tags in different cladded nuclear fuel elements to be used in a liquid metal fast breeder reactor. Failure of the cladding of any fuel element allows fission gases generated in the reaction and these tag isotopes to escape and to combine with the cover gas held in the reactor over the fuel elements. The isotopes specifically are Ne.sup.20, Ne.sup.21 and Ne.sup.22 of neon and Ar.sup.36, Ar.sup.38 and Ar.sup.40 of argon, and the cover gas is helium. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between approximately 0.degree. and -25.degree. C. operable to remove the fission gases from the cover gas and tags and the second or tag recovery system bed is held between approximately -170.degree. and -185.degree. C. operable to isolate the tags from the cover gas. Spectrometric analysis further is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be specifically determined.

  16. Imaging Fukushima Daiichi reactors with muons

    SciTech Connect (OSTI)

    Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Milner, Edward C.; Morris, Christopher L.; Lukic, Zarija; Masuda, Koji; Perry, John O.

    2013-05-15

    A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

  17. Business Opportunities for Small Reactors

    SciTech Connect (OSTI)

    Minato, Akio; Nishimura, Satoshi; Brown, Neil W.

    2007-07-01

    This report assesses the market potential and identifies a number of potential paths for developing the small nuclear reactor business. There are several potential opportunities identified and evaluated. Selecting a specific approach for the business development requires additional information related to a specific market and sources of capital to support the investment. If and how a market for small nuclear plants may develop is difficult to predict because of the complexity of the economic and institutional factors that will influence such development. Key factors are; economics, safety, proliferation resistance and investment risk. The economic and political interest of any of the identified markets is also dependent on successful demonstration of the safety and reliability of small nuclear reactor. Obtaining a US-NRC Standard Design approval would be an important development step toward establishing a market for small reactors. (authors)

  18. Reactor control rod timing system

    DOE Patents [OSTI]

    Wu, Peter T. K. (Clifton Park, NY)

    1982-01-01

    A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  19. Reactor control rod timing system

    SciTech Connect (OSTI)

    Wu, P.T.

    1982-02-09

    A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (Above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  20. Packed-Bed Reactor Study of NETL Sample 196c for the

    Office of Scientific and Technical Information (OSTI)

    Packed-Bed Reactor Study of NETL Sample 196c for the Removal of Carbon Dioxide from Simulated Flue Gas Mixture James S. Hoffman 1, Sonia Hammache 1, 2, McMahan L. Gray 1, Daniel J. Fauth 1, and Henry W. Pennline 1 1U.S. Department of Energy National Energy Technology Laboratory P.O. Box 10940 Pittsburgh, PA 15236 2URS PO Box 618 Library, PA 15129 Abstract An amine-based solid sorbent process to remove CO2 from flue gas has been investigated. The sorbent consists of polyethylenimine (PEI)

  1. Cyclic process for producing methane from carbon monoxide with heat removal

    DOE Patents [OSTI]

    Frost, Albert C. (Congers, NY); Yang, Chang-lee (Spring Valley, NY)

    1982-01-01

    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.

  2. Current status of experimental breeder reactor-II [EBR-II] shutdown planning

    SciTech Connect (OSTI)

    McDermott, M. D.; Griffin, C. D.; Michelbacher, J. A.; Earle, O. K.

    2000-05-08

    The Experimental Breeder Reactor--II (EBR-II) at Argonne National Laboratory--West (ANL-W) in Idaho, was shutdown in September, 1994 as mandated by the US Department of Energy. This sodium cooled reactor had been in service since 1964, and was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the Sodium Process Facility. The sodium environment and the EBR-II configuration, combined with the radiation and contamination associated with thirty years of reactor operation, posed problems specific to liquid metal reactor deactivation. The methods being developed and implemented at EBR-II can be applied to other similar situations in the US and abroad.

  3. Toward reactor monitoring with antineutrinos

    SciTech Connect (OSTI)

    Guillon, Benoit; Cormon, S.; Fallot, M.; Giot, L.; Martino, J.; Cribier, M.; Lasserre, T.

    2007-07-01

    The fundamental knowledge on neutrino properties acquired in recent years as well as the great experimental progress made on neutrino detection open nowadays the possibility of applied neutrino physics. Among it, the International Atomic Energy Agency (IAEA) asked to its member states to study the possibility of nuclear reactor monitoring applications, such as the thermal power measurement or the fuel composition bookkeeping. In this context, we report studies aiming at a better determination of the antineutrino energy spectrum emitted by nuclear power plants, necessary for reactor monitoring applications, but also for experiments studying the ground properties of these particles. (authors)

  4. Actinide Burning in CANDU Reactors

    SciTech Connect (OSTI)

    Hyland, B.; Dyck, G.R.

    2007-07-01

    Actinide burning in CANDU reactors has been studied as a method of reducing the actinide content of spent nuclear fuel from light water reactors, and thereby decreasing the associated long term decay heat load. In this work simulations were performed of actinides mixed with natural uranium to form a mixed oxide (MOX) fuel, and also mixed with silicon carbide to form an inert matrix (IMF) fuel. Both of these fuels were taken to a higher burnup than has previously been studied. The total transuranic element destruction calculated was 40% for the MOX fuel and 71% for the IMF. (authors)

  5. Horizontal baffle for nuclear reactors

    DOE Patents [OSTI]

    Rylatt, John A. (Monroeville, PA)

    1978-01-01

    A horizontal baffle disposed in the annulus defined between the core barrel and the thermal liner of a nuclear reactor thereby physically separating the outlet region of the core from the annular area below the horizontal baffle. The horizontal baffle prevents hot coolant that has passed through the reactor core from thermally damaging apparatus located in the annulus below the horizontal baffle by utilizing the thermally induced bowing of the horizontal baffle to enhance sealing while accommodating lateral motion of the baffle base plate.

  6. Reactor Application for Coaching Newbies

    Energy Science and Technology Software Center (OSTI)

    2015-06-17

    RACCOON is a Moose based reactor physics application designed to engage undergraduate and first-year graduate students. The code contains capabilities to solve the multi group Neutron Diffusion equation in eigenvalue and fixed source form and will soon have a provision to provide simple thermal feedback. These capabilities are sufficient to solve example problems found in Duderstadt & Hamilton (the typical textbook of senior level reactor physics classes). RACCOON does not contain any advanced capabilities asmore » found in YAK.« less

  7. Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

    SciTech Connect (OSTI)

    Lee, D.D.; Collins, J.L.

    2000-02-01

    One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

  8. Radiation dosimetry at the BNL reactor facilities

    SciTech Connect (OSTI)

    Holden, N.E.; Hu, J.P.; Reciniello, R.N.; Greenberg, D.D.; Sengupta, S.; Farrell, K.; Greenwood, L.R.

    1999-07-01

    Neutron and gamma-ray dosimetry measurements have been performed at various facilities in the High Flux Beam Reactor (HFBR) and in the Brookhaven National Laboratory Medical Research Reactor (BMRR). These experimental results are discussed.

  9. Auxiliary reactor for a hydrocarbon reforming system

    DOE Patents [OSTI]

    Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

    2006-01-17

    An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

  10. Reactivity control assembly for nuclear reactor

    DOE Patents [OSTI]

    Bollinger, Lawrence R. (Schenectady, NY)

    1984-01-01

    Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

  11. What can recycling in thermal reactors accomplish?

    SciTech Connect (OSTI)

    Piet, Steven J.; Matthern, Gretchen E.; Jacobson, Jacob J.

    2007-07-01

    Thermal recycle provides several potential benefits when used as stop-gap, mixed, or backup recycling to recycling in fast reactors. These three roles involve a mixture of thermal and fast recycling; fast reactors are required to some degree at some time. Stop-gap uses thermal reactors only until fast reactors are adequately deployed and until any thermal-recycle-only facilities have met their economic lifetime. Mixed uses thermal and fast reactors symbiotically for an extended period of time. Backup uses thermal reactors only if problems later develop in the fast reactor portion of a recycling system. Thermal recycle can also provide benefits when used as pure thermal recycling, with no intention to use fast reactors. However, long term, the pure thermal recycling approach is inadequate to meet several objectives. (authors)

  12. Challenges in the Development of Advanced Reactors

    SciTech Connect (OSTI)

    P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

    2012-08-01

    Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

  13. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

    SciTech Connect (OSTI)

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

  14. Closed Brayton cycle power conversion systems for nuclear reactors :

    SciTech Connect (OSTI)

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.; Sanchez, Travis

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at the manufacturers site (Barber-Nichols Inc.) and installed and operated at Sandia. A sufficiently detailed description of the loop is provided in this report along with the design characteristics of the turbo-alternator-compressor set to allow other researchers to compare their results with those measured in the Sandia test-loop. The third task consisted of a validation effort. In this task the test loop was operated and compared with the modeled results to develop a more complete understanding of this electrically heated closed power generation system and to validate the model. The measured and predicted system temperatures and pressures are in good agreement, indicating that the model is a reasonable representation of the test loop. Typical deviations between the model and the hardware results are less than 10%. Additional tests were performed to assess the capability of the Brayton engine to continue to remove decay heat after the reactor/heater is shutdown, to develop safe and effective control strategies, and to access the effectiveness of gas inventory control as an alternative means to provide load following. In one test the heater power was turned off to simulate a rapid reactor shutdown, and the turbomachinery was driven solely by the sensible heat stored in the heater for over 71 minutes without external power input. This is an important safety feature for CBC systems as it means that the closed Brayton loop will keep cooling the reactor without the need for auxiliary power (other than that needed to circulate the waste heat rejection coolant) provided the heat sink is available.

  15. Experimental Breeder Reactor-II Primary Tank System Wash Water Workshop

    Office of Environmental Management (EM)

    Pre-Developmental INL EBR-II Wash Water Treatment Technologies (PBS # ADSHQTD0100 (0003199)) EBR-II Wash Water Workshop - The majority of the sodium has been removed, remaining material is mostly passivated. Similar closure projects have been successfully completed. Engineering needs to be developed to apply the OBA path. Page 1 of 2 Idaho National Laboratory Idaho Experimental Breeder Reactor-II Primary Tank System Wash Water Workshop Challenge In 1994 Congress ordered the shutdown of the

  16. Nuclear reactor characteristics and operational history

    Gasoline and Diesel Fuel Update (EIA)

    Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: TBD See also: Table 1. Capacity and Generation, Table 2. Ownership Data Table 3. Nuclear Reactor Characteristics and Operational History PDF XLS Plant Name Generator ID Type Reactor Supplier and Model Construction Start Grid Connection Original Expiration Date License Renewal Application License Renewal Issued Extended Expiration Arkansas Nuclear One 1

  17. Research Reactor Conversion | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Reactor Conversion | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the...

  18. Naval Reactors | Y-12 National Security Complex

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

    Naval Reactors Naval Reactors Y-12 processes the feedstock to power the nation's submarines and aircraft carriers. Y-12 processes highly enriched uranium for use by the Naval Reactors Program for Naval Nuclear Propulsion. Our support of the Naval Reactors program began in Fiscal Year 2002 and is currently planned through FY 2050 and beyond. We use dismantled weapons to provide feedstock, moving the material off-site and reducing Y-12's storage footprint and risk. The United States stopped

  19. Digital computer operation of a nuclear reactor

    DOE Patents [OSTI]

    Colley, Robert W. (Richland, WA)

    1984-01-01

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

  20. Digital computer operation of a nuclear reactor

    DOE Patents [OSTI]

    Colley, R.W.

    1982-06-29

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.