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1

Savannah River Site Removes Dome, Opening Reactor for Recovery Act  

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

Savannah River Site Removes Dome, Opening Reactor for Recovery Act Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning 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 rusty, orange, 75-foot-tall dome. With the help of a 660-ton crane and lifting lugs, the workers pulled the 174,000-pound dome off the Heavy Water Components Test Reactor, capping more than 16 months of preparations. Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning More Documents & Publications Recovery Act Changes Hanford Skyline with Explosive Demolitions Recovery Act Workers Add Time Capsule Before Sealing Reactor for Hundreds

2

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

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

Recovery Act Funds Test Reactor Dome Removal in Historic D&D Recovery Act 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 Cold War-era test reactor. This $25-million reactor decommissioning and deactivation project is funded By the American Recovery and Reinvestment Act. Affectionately known by SRS employees as "Hector," the iconic Heavy Water Components Test Reactor (HWCTR) has stood in the Site's B Area since 1959

3

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

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

American Recovery and Reinvestment 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 allows workers to access the 219,000-pound reactor vessel and two steam generators so they can remove and permanently dispose them onsite. Re- maining equipment will be moved to the cavity vacated by the vessel, and below-grade portions of the reactor will be

4

Confirmatory Survey Results for the Reactor Building Dome Upper Surfaces, Rancho Saco Nuclear Generating Station  

SciTech Connect

Results from a confirmatory survey of the upper structural surfaces of the Reactor Building Dome at the Rancho Seco Nuclear Generating Station (RSNGS) performed by the Oak Ridge Institute for Science and Education for the NRC. Also includes results of interlaboratory comparison analyses on several archived soil samples that would be provided by RSNGS personnel. The confirmatory surveys were performed on June 7 and 8, 2006.

Wade C. Adams

2006-10-25T23:59:59.000Z

5

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

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).

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

6

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

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).

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

7

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

SciTech Connect

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

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

1994-08-01T23:59:59.000Z

8

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

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

9

Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1  

SciTech Connect

This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

Owen, M.B.

1997-04-01T23:59:59.000Z

10

Innovative Graphite Removal Technology for Graphite Moderated Reactor Decommissioning  

Science Conference Proceedings (OSTI)

This report defines a trial program to support the development of a new concept for the removal of reactor graphite by remote in-situ size reduction and vacuum transfer, known as nibble-and-vacuum. This new approach to graphite retrieval has significant potential for simplifying the decommissioning process of graphite moderated reactors. It produces graphite gravel, which has potential as feedstock for processes such as gasification/steam reforming. This report includes definition of the trial program, t...

2010-09-28T23:59:59.000Z

11

Emergency heat removal system for a nuclear reactor  

DOE Patents (OSTI)

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

Dunckel, Thomas L. (Potomac, MD)

1976-01-01T23:59:59.000Z

12

Control of reactor coolant flow path during reactor decay heat removal  

DOE Patents (OSTI)

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.

Hunsbedt, Anstein N. (Los Gatos, CA)

1988-01-01T23:59:59.000Z

13

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

SciTech Connect

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

P.C. Weaver

2010-07-09T23:59:59.000Z

14

Process and apparatus for adding and removing particles from pressurized reactors  

DOE Patents (OSTI)

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

Milligan, John D. (Little Silver, NJ)

1983-01-01T23:59:59.000Z

15

System for fuel rod removal from a reactor module  

DOE Patents (OSTI)

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.

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)

1990-01-01T23:59:59.000Z

16

System for fuel rod removal from a reactor module  

DOE Patents (OSTI)

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.

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

1988-07-28T23:59:59.000Z

17

Method for removing cesium from a nuclear reactor coolant  

DOE Patents (OSTI)

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.

Colburn, R.P.

1983-08-10T23:59:59.000Z

18

Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts  

SciTech Connect

Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850ºC at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05).

K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

2005-09-01T23:59:59.000Z

19

The removal and segmentation of the Yankee Rowe reactor vessel internals  

Science Conference Proceedings (OSTI)

A major element of the reactor decommissioning of the Rowe Yankee reactor was the segmentation and packaging of the reactor internals. PCI Energy Services, specializing in remote cutting, machining, and welding, performed this work under contract to Yankee Atomic Electric Company. Removal techniques are described.

Child, C.; McGough, M.; Smith, G. [Power Cutting Inc., Lake Bluff, IL (United States)

1995-12-31T23:59:59.000Z

20

100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT  

Science Conference Proceedings (OSTI)

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.

HARRINGTON RA

2010-01-15T23:59:59.000Z

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


21

Passive decay heat removal system for water-cooled nuclear reactors  

DOE Patents (OSTI)

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.

Forsberg, Charles W. (Oak Ridge, TN)

1991-01-01T23:59:59.000Z

22

Start-up of Autotrophic Nitrogen Removal Reactors via Sequential  

E-Print Network (OSTI)

, filtration, centrifugation, etc.) (10) Analyze the liquid portion for the residual, final, equili- brium is that based on a flow-through fixed- bed type of a reactor/contactor (Fig. 4). The sorption column would in parallel. Rarely, usually because of special process requirements, a mixed contactor could also be used

Smets, Barth F.

23

Nuclear reactor having a polyhedral primary shield and removable vessel insulation  

DOE Patents (OSTI)

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.

Ekeroth, D.E.; Orr, R.

1993-12-07T23:59:59.000Z

24

Nuclear reactor having a polyhedral primary shield and removable vessel insulation  

DOE Patents (OSTI)

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.

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

1993-01-01T23:59:59.000Z

25

Removal of polychlorinated phenols in sequential anaerobic-aerobic biofilm reactors packed with tire chips  

Science Conference Proceedings (OSTI)

Scrap vehicle tire chips were used as packing material for sequential anaerobic-aerobic biofilm reactors to remove persistent chlorinated hydrocarbons. Adsorption capacity of scrap tires was greater under acidic conditions than under basic conditions. However, it was only approximately 0.04 to 0.3% of that of activated carbon. The amount of biomass that attached to the surface of scrap tires was 3.16 and 3.72 mg volatile suspended solids/cm{sup 2} after 14 and 37 days, respectively. Two laboratory-scale, down-flow anaerobic-aerobic biofilm reactors packed with tire chips were operated to remove 2,4-dichlorophenol (DCP) and 4-chlorophenol (CP). More than 98% of DCP was dehalogenated to CP in the anaerobic reactor, 70 to 98% of which was subsequently degraded in the aerobic reactor. Scrap tires did not cause any operational problems when used as biofilter media.

Shin, H.S.; Yoo, K.S.; Park, J.K.

1999-05-01T23:59:59.000Z

26

Nuclear reactor with makeup water assist from residual heat removal system  

DOE Patents (OSTI)

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.

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

1993-12-07T23:59:59.000Z

27

Nuclear reactor with makeup water assist from residual heat removal system  

DOE Patents (OSTI)

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.

Corletti, Michael M. (New Kensington, PA); Schulz, Terry L. (Murrysville, PA)

1993-01-01T23:59:59.000Z

28

Removal of uranium and salt from the Molten Salt Reactor Experiment  

SciTech Connect

In 1994, migration of {sup 233}U was discovered to have occurred at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). This paper describes the actions now underway to remove uranium from the off-gas piping and the charcoal bed, to remove and stabilize the salts, and to convert the uranium to a stable oxide for long-term storage.

Peretz, F.J.; Rushton, J.E.; Faulkner, R.L.; Walker, K.L.; Del Cul, G.D.

1998-06-01T23:59:59.000Z

29

Passive decay heat removal system for water-cooled nuclear reactors  

DOE Patents (OSTI)

This document describes passive decay-heat removal system for a water-cooled nuclear reactor which employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated evaporator 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.

Forseberg, C.W.

1990-01-01T23:59:59.000Z

30

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

Science Conference Proceedings (OSTI)

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)

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

2006-07-01T23:59:59.000Z

31

Passive decay heat removal system for water-cooled nuclear reactors  

DOE Patents (OSTI)

This document describes passive decay-heat removal system for a water-cooled nuclear reactor which employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated evaporator 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.

Forseberg, C.W.

1990-12-31T23:59:59.000Z

32

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

DOE Patents (OSTI)

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.

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)

1978-01-01T23:59:59.000Z

33

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

SciTech Connect

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

Wei Yanjie [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Key Laboratory of Environmental Protection in Water Transport Engineering Ministry of Communications, Tianjin Research Institute of Water Transport Engineering, Tianjin 300456 (China); Ji Min, E-mail: jmtju@yahoo.cn [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Li Ruying [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Qin Feifei [Tianjin Tanggu Sino French Water Supply Co. Ltd., Tianjin 300450 (China)

2012-03-15T23:59:59.000Z

34

Highly Selective Nuclide Removal from the R-Reactor Disassembly Basin at SRS  

Science Conference Proceedings (OSTI)

This paper describes the results of a deployment of highly selective ion-exchange resin technologies for the in-situ removal of Cs-137 and Sr-90 from the Savannah River Site (SRS) R-Reactor Disassembly Basin. The deployment was supported by the DOE Office of Science and Technology's (OST, EM-50) National Engineering Technology Laboratory (NETL), as a part of an Accelerated Site Technology Deployment (ASTD) project. The Facilities Decontamination and Decommissioning (FDD) Program at the SRS conducted this deployment as a part of an overall program to deactivate three of the site's five reactor disassembly basins

Pickett, J.B.

2002-08-13T23:59:59.000Z

35

Highly Selective Nuclide Removal from the R-Reactor Disassembly Basin at the SRS  

Science Conference Proceedings (OSTI)

This paper describes the results of a deployment of highly selective ion-exchange resin technologies for the in-situ removal of Cs-137 and Sr-90 from the Savannah River Site (SRS) R-Reactor Disassembly Basin. The deployment was supported by the DOE Office of Science and Technology's (OST, EM-50) National Engineering Technology Laboratory (NETL), as a part of an Accelerated Site Technology Deployment (ASTD) project. The Facilities Decontamination and Decommissioning (FDD) Program at the SRS conducted this deployment as a part of an overall program to deactivate three of the site's five reactor disassembly basins.

Pickett, J. B.; Austin, W. E.; Dukes, H. H.

2002-02-26T23:59:59.000Z

36

Trojan Nuclear Power Plant Reactor Vessel and Internals Removal: Trojan Nuclear Plant Decommissioning Experience  

Science Conference Proceedings (OSTI)

One goal of the EPRI Decommissioning Technology Program is to capture the growing utility experience in nuclear plant decommissioning activities for the benefit of other utilities facing similar challenges in the future. This report provides historical information on the background, scope, organization, schedule, cost, contracts, and support activities associated with the Trojan Nuclear Plant Reactor Vessel and Internals Removal (RVAIR) Project. Also discussed are problems, successes, and lessons learned...

2000-10-16T23:59:59.000Z

37

Assessment of passive decay heat removal in the General Atomics Modular Helium Reactor  

E-Print Network (OSTI)

The purpose of this report is to present the results of the study and analysis of loss-of-coolant and loss-of-flow simulations performed on the Modular Helium Reactor developed by General Atomics using the thermal-hydraulics code RELAP5-3D/ATHENA. The MHR is a high temperature gas cooled reactor. It is a prismatic core concept for New Generation Nuclear Plant (NGNP). Very few reactors of that kind have been designed in the past. Furthermore, the MHR is supposed to be a highly passively safe concept. So there are high needs for numerical simulations in order to confirm the design. The project is dedicated to the assessment of the passive decay heat capabilities of the reactor under abnormal transient conditions. To comply with the requirements of the NGNP, fuel and structural temperatures must be kept under design safety limits under any circumstances. During the project, the MHR has been investigated: first under steady-state conditions and then under transient settings. The project confirms that satisfying passive decay heat removal by means of natural heat transfer mechanisms (convection, conduction and radiation) occurs.

Cocheme, Francois Guilhem

2004-12-01T23:59:59.000Z

38

Ceramic dome receiver technology developments  

DOE Green Energy (OSTI)

The development and experimental demonstration of a high-temperature seal for the SHARE ceramic dome cavity receiver is reported. The mechanical contact seal which was tested on one-foot-diameter silicon-carbide ceramic-dome hardware at pressure differentials to four atmospheres and dome temperatures to 2200/sup 0/F (1200/sup 0/C) showed negligible leakage at expected receiver operating conditions. Potential solar receiver applications for the technology are illustrated.

Jarvinen, P. O.

1980-01-01T23:59:59.000Z

39

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

DOE Patents (OSTI)

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.

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

1986-01-01T23:59:59.000Z

40

Waste removal in pyrochemical fuel processing for the Integral Fast Reactor  

SciTech Connect

Electrorefining in a molten salt electrolyte is used in the Integral Fast Reactor fuel cycle to recover actinides from spent fuel. Processes that are being developed for removing the waste constituents from the electrorefiner and incorporating them into the waste forms are described in this paper. During processing, halogen, chalcogen, alkali, alkaline earth, and rare earth fission products build up in the molten salt as metal halides and anions, and fuel cladding hulls and noble metal fission products remain as metals of various particle sizes. Essentially all transuranic actinides are collected as metals on cathodes, and are converted to new metal fuel. After processing, fission products and other waste are removed to a metal and a mineral waste form. The metal waste form contains the cladding hulls, noble metal fission products, and (optionally) most rare earths in a copper or stainless steel matrix. The mineral waste form contains fission products that have been removed from the salt into a zeolite or zeolite-derived matrix.

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

1994-01-01T23:59:59.000Z

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


41

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

Science Conference Proceedings (OSTI)

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

P.C. Weaver

2010-11-03T23:59:59.000Z

42

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

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

E.M. Harpenau

2010-12-15T23:59:59.000Z

43

Solution-reactor-produced-{sup 99}Mo using activated carbon to remove {sup 131}I  

SciTech Connect

This research explores the idea of producing {sup 99}Mo in a solution reactor. The Solution High Energy Burst Assembly (SHEBA), located at the Los Alamos Critical Assembly Facility, was used to facilitate this study. The goal of this study was to build on work previously completed and to investigate a possible mode of radioactive contaminant removal prior to a {sup 99}Mo extraction process. Prior experiments, performed using SHEBA and a single-step sorption process, showed a significant amount of {sup 131}I present along with the {sup 99}Mo on the alumina that was used to isolate the {sup 99}Mo. A high concentration of {sup 131}I and/or other contaminants present in a sample prohibits the Food and Drug Administration from approving an extraction of that nature for radiopharmaceutical use. However, if it were possible to remove the {sup 131}I and other contaminants prior to a {sup 99}Mo extraction, a simple column extraction process might be feasible. Activated charcoal was used to try to filter the {sup 131}I contaminant from an irradiated fuel solution. Gamma spectroscopy confirmed that the activated carbon trapped a significant amount of the {sup 131}I, as well as notable amounts of {sup 133}Xe, {sup 105}Rb, and {sup 140}Ba. Most importantly, the carbon traps a diminutive amount of {sup 99}Mo.

Kitten, S.; Cappiello, C. [Los Alamos National Lab., NM (United States)

1998-09-01T23:59:59.000Z

44

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

Science Conference Proceedings (OSTI)

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

Lata

1996-09-26T23:59:59.000Z

45

Dome Tech | Open Energy Information  

Open Energy Info (EERE)

Dome Tech 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 consulting, & project development with the intention of optimising building performance. Coordinates 40.556614°, -82.866255° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":40.556614,"lon":-82.866255,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

46

Design of passive decay heat removal system for the lead cooled flexible conversion ratio fast reactor  

E-Print Network (OSTI)

The lead-cooled flexible conversion ratio fast reactor shows many benefits over other fast-reactor designs; however, the higher power rating and denser primary coolant present difficulties for the design of a passive decay ...

Whitman, Joshua (Joshua J.)

2007-01-01T23:59:59.000Z

47

Simultaneous Removal of NOx and Mercury in Low Temperature Selective Catalytic and Adsorptive Reactor  

SciTech Connect

The results of a 18-month investigation to advance the development of a novel Low Temperature Selective Catalytic and Adsorptive Reactor (LTSCAR), for the simultaneous removal of NO{sub x} and mercury (elemental and oxidized) from flue gases in a single unit operation located downstream of the particulate collectors, are reported. In the proposed LTSCAR, NO{sub x} removal is in a traditional SCR mode but at low temperature, and, uniquely, using carbon monoxide as a reductant. The concomitant capture of mercury in the unit is achieved through the incorporation of a novel chelating adsorbent. As conceptualized, the LTSCAR will be located downstream of the particulate collectors (flue gas temperature 140-160 C) and will be similar in structure to a conventional SCR. That is, it will have 3-4 beds that are loaded with catalyst and adsorbent allowing staged replacement of catalyst and adsorbent as required. Various Mn/TiO{sub 2} SCR catalysts were synthesized and evaluated for their ability to reduce NO at low temperature using CO as the reductant. It has been shown that with a suitably tailored catalyst more than 65% NO conversion with 100% N{sub 2} selectivity can be achieved, even at a high space velocity (SV) of 50,000 h-1 and in the presence of 2 v% H{sub 2}O. Three adsorbents for oxidized mercury were developed in this project with thermal stability in the required range. Based on detailed evaluations of their characteristics, the mercaptopropyltrimethoxysilane (MPTS) adsorbent was found to be most promising for the capture of oxidized mercury. This adsorbent has been shown to be thermally stable to 200 C. Fixed-bed evaluations in the targeted temperature range demonstrated effective removal of oxidized mercury from simulated flue gas at very high capacity ({approx}>58 mg Hg/g adsorbent). Extension of the capability of the adsorbent to elemental mercury capture was pursued with two independent approaches: incorporation of a novel nano-layer on the surface of the chelating mercury adsorbent to achieve in situ oxidation on the adsorbent, and the use of a separate titania-supported manganese oxide catalyst upstream of the oxidized mercury adsorbent. Both approaches met with some success. It was demonstrated that the concept of in situ oxidation on the adsorbent is viable, but the future challenge is to raise the operating capacity beyond the achieved limit of 2.7 mg Hg/g adsorbent. With regard to the manganese dioxide catalyst, elemental mercury was very efficiently oxidized in the absence of sulfur dioxide. Adequate resistance to sulfur dioxide must be incorporated for the approach to be feasible in flue gas. A preliminary benefits analysis of the technology suggests significant potential economic and environmental advantages.

Neville G. Pinto; Panagiotis G. Smirniotis

2006-03-31T23:59:59.000Z

48

Passive residual heat removal systems for current and future light water reactors.  

E-Print Network (OSTI)

??The main problem with nuclear power during a shutdown is that decay heat is still present and needs to be removed to prevent a number… (more)

Treleaven, Ian James

2013-01-01T23:59:59.000Z

49

North Dome decision expected soon  

Science Conference Proceedings (OSTI)

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

Not Available

1981-08-01T23:59:59.000Z

50

Dome-Tech and Merck Teaming Profile  

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

Dome-Tech, Inc. Merck & Co., Inc. Dome-Tech, Inc. Merck & Co., Inc. 510 Thornall Street One Merck Drive Edison, New Jersey 08837 Whitehouse Station, New Jersey 08889 Business: Energy Engineering Consulting Business: Pharmaceutical Manufacturing Ed Liberty Rob Colucci Vice President, Energy Advisors Senior Director, Energy and Sustainability Phone: 732-590-0122 ext. 165 Phone: 908-423-4971 Email: e_liberty@dome-tech.com Email: robert_colucci@merck.com Dome-Tech installs 1.6 MW DC solar power tracking system at Merck to reduce CO 2 emissions Project Scope Dome-Tech installed a 1.6 megawatt solar photovoltaic (PV) system at Merck's corporate headquarters in New Jersey - with no capital investment by Merck. Project Summary Dome-Tech worked with SunPower to install 7,000 solar panels on 7 acres of land at Merck's

51

Resurgent Dome Complex | Open Energy Information  

Open Energy Info (EERE)

source source History View New Pages Recent Changes All Special Pages Semantic Search/Querying Get Involved Help Apps Datasets Community Login | Sign Up Search Page Edit History Facebook icon Twitter icon » Resurgent Dome Complex Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Print PDF Resurgent Dome Complex Dictionary.png Resurgent Dome Complex: Resurgent domes are encountered near the center of many caldera depressions, and form via uplift of the caldera valley floor due to movement in the underlying magma chamber. Resurgent domes typically host numerous deformation structures that act as conduits for hydrothermal fluids in the shallow crust. Other definitions:Wikipedia Reegle Topographic Features List of topographic features commonly encountered in geothermal resource

52

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

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...

53

WA_1994_001_TEXACO_EXPLORATION_AND_PRODUCTION_Waiver_of_Dome...  

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

01TEXACOEXPLORATIONANDPRODUCTIONWaiverofDome.pdf WA1994001TEXACOEXPLORATIONANDPRODUCTIONWaiverofDome.pdf WA1994001TEXACOEXPLORATIONANDPRODUCTIONWaiverofDom...

54

WA_1993_023_TEXACO_EXPLORATION_AND_PRODUCTION_Waiver_of_Dome...  

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

23TEXACOEXPLORATIONANDPRODUCTIONWaiverofDome.pdf WA1993023TEXACOEXPLORATIONANDPRODUCTIONWaiverofDome.pdf WA1993023TEXACOEXPLORATIONANDPRODUCTIONWaiverofDom...

55

Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.  

DOE Green Energy (OSTI)

As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong 3-D effects result in large heat flux, temperature, and heat transfer variations around the tube wall; (b) there is a large difference in the heat transfer coefficient predicted by turbulence models and heat transfer correlations, and this underscores the need of experimental work to validate the thermal performance of the RCCS; and (c) tests at the NSTF would embody all important fluid flow and heat transfer phenomena in the RCCS, in addition to covering the entire parameter ranges that characterize these phenomena. Additional supporting scaling study results are available in Reference 2. The purpose of this work is to develop a high-level engineering plan for mechanical and instrumentation modifications to NSTF in order to meet the following two technical objectives: (1) provide CFD and system-level code development and validation data for the RCCS under prototypic (full-scale) natural convection flow conditions, and (2) support RCCS design validation and optimization. As background for this work, the report begins by providing a summary of the original NSTF design and operational capabilities. Since the facility has not been actively utilized since the early 1990's, the next step is to assess the current facility status. With this background material in place, the data needs and requirements for the facility are then defined on the basis of supporting analysis activities. With the requirements for the facility established, appropriate mechanical and instrumentation modifications to NSTF are then developed in order to meet the overall project objectives. A cost and schedule for modifying the facility to satisfy the RCCS data needs is then provided.

Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

2005-09-01T23:59:59.000Z

56

Effective pine bark composting with the Dome Aeration Technology  

SciTech Connect

In South Africa garden refuse is primarily disposed of in domestic landfills. Due to the large quantities generated, any form of treatment would be beneficial for volume reduction, waste stabilization and resource recovery. Dome Aeration Technology (DAT) is an advanced process for aerobic biological degradation of garden refuse and general waste [Paar, S., Brummack, J., Gemende, B., 1999a. Advantages of dome aeration in mechanical-biological waste treatment. In: Proceedings of the 7th International Waste Management and Landfill Symposium, Cagliari, 4-8 October 1999; Paar, S., Brummack, J., Gemende, B., 1999b. Mechanical-biological waste stabilization by the dome aeration method. Environment Protection Engineering 25 (3/99). Mollekopf, N., Brummack, J., Paar, S., Vorster, K., 2002. Use of the Dome Aeration Technology for biochemical stabilization of waste prior to landfilling. In: Proceedings of the Wastecon 2002, Waste Congress and Exhibition, Durban, South Africa.]. It is a non-reactor open windrow composting process, with the main advantage being that the input material needs no periodic turning. A rotting time of only 3-4 months indicates the high efficiency. Additionally, the low capital/operational costs, low energy inputs and limited plant requirements provide potential for use in aerobic refuse stabilization. The innovation in the DAT process is the passive aeration achieved by thermally driven advection through open windrows caused by temperature differences between the degrading material and the outside environment. This paper investigates the application of Dome Aeration Technology to pine bark composting as part of an integrated waste management strategy. A full-scale field experiment was performed at the Bisasar Road Landfill Site in Durban to assess the influence of climate, waste composition and operational conditions on the process. A test windrow was constructed and measurements of temperature and airflow through the material were taken. The process monitoring revealed that prevailing climatic conditions in a subtropical location do not affect the high efficiency of this technology. However, the composition of the input material can be detrimental for production of high quality compost because of a lack of nitrate.

Trois, Cristina [CRECHE Centre for Research in Environmental, Coastal and Hydrological Engineering, School of Civil Engineering, Survey and Construction, University of KwaZulu-Natal, Durban 4041 (South Africa)]. E-mail: troisc@ukzn.ac.za; Polster, Andreas [Dresden University of Technology, Institute for Process Engineering and Environmental Technology, Helmholtzstrasse 14, 01062 Dresden (Germany)

2007-07-01T23:59:59.000Z

57

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING  

Science Conference Proceedings (OSTI)

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.

Austin, W.; Brinkley, D.

2011-10-13T23:59:59.000Z

58

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

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

Complex Description Complex Description Current HFBR Complex The HFBR complex consists of multiple structures and systems that were necessary to operate and maintain the reactor. The most recognizable features of the complex are the domed reactor confinement building and the distinctive red-and-white stack. Portions of the complex building structures, systems, and components, some of which are underground, were contaminated with radionuclides and chemicals as a result of previous HFBR and Brookhaven Graphite Research Reactor (BGRR) operations. A number of decommissioning and preparation for long-term safe storage actions have been taken including the removal of contaminated structures, hazardous materials, and contaminated equipment and components. The structures and systems, both current and former, are

59

Dome-Tech & Schering-Plough Teaming Profile | ENERGY STAR  

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

Profile cover page of document Dome-Tech's retro-commissioning and recalibration program led to Schering-Plough realizing annual savings of 450,000. Dome-Tech & Schering-Plough...

60

A retrospective look at the UMass DOME mobile testbed  

Science Conference Proceedings (OSTI)

In this paper we describe the evolution of DOME, a diverse outdoor testbed for mobile experimentation. In addition, while highlighting the challenges faced in construction of DOME, we describe a concrete set of scientific results derived from this experience ...

Hamed Soroush; Nilanjan Banerjee; Mark Corner; Brian Levine; Brian Lynn

2012-03-01T23:59:59.000Z

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


61

Study on the Isolation Screening and Characteristic Identification of Denitrifying Phosphorus Removing Bacteria in Sequencing Batch Biofilm Reactor  

Science Conference Proceedings (OSTI)

The processes of denitrification and dephosphorization which were independent originally were integrated into Biological Denitrifying Phosphorus and removing process. It was widely considered a bright technology of biological phosphorus removal. According ... Keywords: denitrifying phosphorus removing bacteria, enrichment and screening, characteristic identification, flat plate scribing method, metachromatic granules dyeing, nitrate reduction test

Yafeng Li; Hongtao Liu; Jing Ren

2012-05-01T23:59:59.000Z

62

SUMMARY AND RESULTS LETTER REPORT – INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PROJECT, PHASE 3: TRENCHES 2, 3, AND 4 BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK  

SciTech Connect

5098-LR-02-0 SUMMARY AND RESULTS LETTER REPORT – INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PROJECT, PHASE 3 TRENCHES 2, 3, AND 4 BROOKHAVEN NATIONAL LABORATORY

E.M. Harpenau

2010-11-15T23:59:59.000Z

63

Reactor pressure vessel vented head  

DOE Patents (OSTI)

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.

Sawabe, James K. (San Jose, CA)

1994-01-11T23:59:59.000Z

64

REACTOR  

DOE Patents (OSTI)

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

Roman, W.G.

1961-06-27T23:59:59.000Z

65

Decommissioning of the high flux beam reactor at Brookhaven Lab  

Science Conference Proceedings (OSTI)

The high-flux beam reactor (HFBR) at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on Oct. 31, 1965. It operated at a power level of 40 megawatts. An equipment upgrade in 1982 allowed operations at 60 megawatts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 megawatts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of groundwater from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost three years for safety and environmental reviews. In November 1999 the United States Dept. of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel, is presently under 24/7 surveillance for safety. Detailed dosimetry performed for the HFBR decommissioning during 1996-2009 is described in the paper. (authors)

Hu, J.P. [National Synchrotron Light Source, Brookhaven Laboratory, Upton, NY 11973 (United States); Reciniello, R.N. [Radiological Control Div., Brookhaven Laboratory, Upton, NY 11973 (United States); Holden, N.E. [National Nuclear Data Center, Brookhaven Laboratory, Upton, NY 11973 (United States)

2011-07-01T23:59:59.000Z

66

Gas-cooled reactor for space power systems  

Science Conference Proceedings (OSTI)

Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors.

Walter, C.E.; Pearson, J.S.

1987-05-01T23:59:59.000Z

67

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

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

Reactor Decommissioning Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility...

68

Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor using RELAP5 and TEMPEST: Part 1, Models and simulation results  

Science Conference Proceedings (OSTI)

A study was conducted to examine decay heat removal requirements in the High Flux Isotope Reactor (HFIR) following shutdown from 85 MW. The objective of the study was to determine when forced flow through the core could be terminated without causing the fuel to melt. This question is particularly relevant when a station blackout caused by an external event is considered. Analysis of natural circulation in the core, vessel upper plenum, and reactor pool indicates that 12 h of forced flow will permit a safe shutdown with some margin. However, uncertainties in the analysis preclude conclusive proof that 12 h is sufficient. As a result of the study, two seismically qualified diesel generators were installed in HFIR. 9 refs., 4 figs.

Morris, D.G.; Wendel, M.W.; Chen, N.C.J.; Ruggles, A.E.; Cook, D.H.

1989-01-01T23:59:59.000Z

69

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

Daniels, F.

1959-10-27T23:59:59.000Z

70

Big Sky Regional Carbon Sequestration Partnership--Kevin Dome...  

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

Carbon Services Vecta Oil & Gas, Ltd. Washington State University Big Sky Regional Carbon Sequestration Partnership-Kevin Dome Development Phase Project Background The U.S....

71

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

Open Energy Info (EERE)

dome. Although this system apparently died off as a result of mineral deposition and cooling (andor deepening) of magmatic heat sources, flow testing and tidal analyses of...

72

Prospective Type Ia supernova surveys from Dome A  

E-Print Network (OSTI)

Prospective Type Ia Supernova Surveys From Dome A A. Kim a ,are conducive toward Type Ia supernova surveys forheterogeneities within the Type Ia supernova class, reducing

Kim, A.

2010-01-01T23:59:59.000Z

73

Distribution of quaternary rhyolite dome of the Coso Range, California...  

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Distribution of quaternary rhyolite dome of the Coso Range, California: Implications for extent...

74

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

SciTech Connect

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.

Neville Pinto; Panagiotis Smirniotis; Stephen Thiel

2010-08-31T23:59:59.000Z

75

Simulation of the Tropical Thermal Domes in the Atlantic: A Seasonal Cycle  

Science Conference Proceedings (OSTI)

A general ocean circulation model is used to analyze seasonal variations in the Guinea Dome and Angola Dome regions. The cold Guinea Dome develops from June through September due to active divergence of heat transport. The cooling is related to ...

Toshio Yamagata; Satoshi Iizuka

1995-09-01T23:59:59.000Z

76

SOLUTION MINING IN SALT DOMES OF THE GULF COAST EMBAYMENT  

Science Conference Proceedings (OSTI)

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.

Griswold, G. B.

1981-02-01T23:59:59.000Z

77

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

Science Conference Proceedings (OSTI)

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.

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

2012-04-24T23:59:59.000Z

78

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

SciTech Connect

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.

CHENG,L.Y.; LUDEWIG, H.

2007-06-01T23:59:59.000Z

79

Inflatable Evergreen Polar Zone Dome (EPZD) Settlements  

E-Print Network (OSTI)

Sustaining human life at the Earth antipodal Polar Regions is very difficult especially during Winter when water-freezing air temperature, blizzards and whiteouts make normal human existence dangerous. To counter these environmental stresses, we offer the innovative artificial Evergreen Polar Zone Dome (EPZD), an inflated half-hemisphere with interiors continuously providing a Mediterranean Sea-like climate. The Evergreen EPZD structural theory is developed, substantiated by key computations that show it is possible for current building technology to construct and heat large enclosed volumes inexpensively. Specifically, a satisfactory result is reached by using sunlight reflectors and a special double thin film, which concentrates all available solar energy inside the EPZD while, at the same time markedly decreasing the heat loss to exterior Polar Region air. Someday a similar, but remarkably more technological, EPZD design may be employed at proposed Moon and Mars settlements. Key words: artificial hemisphere, inflatable film building, Polar Region homes, solar energy concentrator.

Alexander Bolonkin; Richard Cathcart

2007-01-09T23:59:59.000Z

80

Thermohydraulic investigations by decay heat removal systems; By natural convection for liquid-metal fast breeder reactors  

SciTech Connect

This paper reports, to examine the function of the safety-related SNR-2 decay heat removal concept, natural convection experiments performed in two-and three-dimensional water models, scaled 1:20, under conditions of symmetric and nonsymmetric loads of the immersed coolers installed in the upper plenum at 180-deg positions with respect to each other. The temperature and velocity distributions were measured and the flow patterns recorded for different configurations of the instrumented plug. For symmetric load conditions, symmetric temperature and flow distributions were measured in two- and three-dimensional models. Nonsymmetric load conditions produce remarkable temperature differences between the two separated plenums of the two-dimensional model if fluid circulation is suppressed by a closed plug. An open plug allows fluid to pass through and shows lower temperature differences. In contrast, in the three-dimensional experiment, azimuthal fluid flow inside the plenum prevails even with the plug closed, and identical temperature distributions are measured. The calculations using the COMMIX-1B code are generally in good agreement with the measurements.

Hoffmann, H.; Weinberg, D. (Kernforschungszentrum Karlsruhe GmbH Institut fur Reaktorbaulemente, Postfach 3640, 7500 Karsruhe 1 (DE)); Ieday (Power Reactor and Nuclear Fuel Development Corporation Oarai Engineering Center, Oarai (JP)); Martex, K.; Tschoke, H.H.; Dres, K. (Kernforschungszentrum Karlsruhe GmbH, Institut fur Reaktorbauelemente, Postfach 3640, 7500 Karlsruhe 1 (DE))

1989-10-01T23:59:59.000Z

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


81

Distribution of quaternary rhyolite dome of the Coso Range, California:  

Open Energy Info (EERE)

of quaternary rhyolite dome of the Coso Range, California: of quaternary rhyolite dome of the Coso Range, California: Implications for extent of the geothermal anomaly Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Distribution of quaternary rhyolite dome of the Coso Range, California: Implications for extent of the geothermal anomaly Details Activities (1) Areas (1) Regions (0) Abstract: Thirty-eight separate domes and flows of phenocryst-poor, high-silica rhyolite of similar major element chemical composition were erupted over the past 1 m.y. from vents arranged in a crudely S-shaped array atop a granitic horst in the Coso Range, California. Most of the extrusions are probably less than about 0.3 m.y. old. The area is one of Quaternary basaltic volcanism and crustal extension. The central part of

82

Development of construction techniques in the Mamluk domes of Cairo  

E-Print Network (OSTI)

This dissertation reconstructs the building features, the construction methods and the esthetic and structural changes of the Mamluk Mausolea in Cairo (1250-1517 A.D.); a special attention is dedicated to the domes that ...

Cipriani, Barbara

2005-01-01T23:59:59.000Z

83

Reactor pressure vessel vented head  

DOE Patents (OSTI)

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.

Sawabe, J.K.

1994-01-11T23:59:59.000Z

84

LANL demolishes first containment dome at disposal area  

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

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

85

FLAMMABLE GAS DIFFUSION THROUGH SINGLE SHELL TANK (SST) DOMES  

DOE Green Energy (OSTI)

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.

MEACHAM, J.E.

2003-11-10T23:59:59.000Z

86

Historical CO2 Records from the Law Dome DE08, DE08-2, and DSS...  

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

Law Dome Ice Cores Historical CO2 Records from the Law Dome DE08, DE08-2, and DSS Ice Cores graphics Graphics data Data Investigators D.M. Etheridge, L.P. Steele, R.L. Langenfelds...

87

A Method to Correct the Thermal Dome Effect of Pyranometers in Selected Historical Solar Irradiance Measurements  

Science Conference Proceedings (OSTI)

In using pyranometers to measure solar irradiance, it is important to know the magnitudes and the consequences of the thermal effect, which is introduced by the glass domes of the instruments. Historically, the thermal dome effect was not ...

Qiang Ji

2007-03-01T23:59:59.000Z

88

Effects of Solar Radiation on the Performance of Pyrgeometers with Silicon Domes  

Science Conference Proceedings (OSTI)

The performance of the Eppley pyrgeometer with a silicon dome presents several anomalies during daytime measurements. These problems are related mainly to the solar heating of the dome, which causes nearly instantaneous fluctuations, about ±1%–2% ...

L. Alados-Arboledas; J. Vida; J. I. Jiménez

1988-10-01T23:59:59.000Z

89

Performance evaluation of 24 ion exchange materials for removing cesium and strontium from actual and simulated N-Reactor storage basin water  

Science Conference Proceedings (OSTI)

This report describes the evaluation of 24 organic and inorganic ion exchange materials for removing cesium and strontium from actual and simulated waters from the 100 Area 105 N-Reactor fuel storage basin. The data described in this report can be applied for developing and evaluating ion exchange pre-treatment process flowsheets. Cesium and strontium batch distribution ratios (K{sub d}`s), decontamination factors (DF), and material loadings (mmol g{sup -1}) are compared as a function of ion exchange material and initial cesium concentration. The actual and simulated N-Basin waters contain relatively low levels of aluminum, barium, calcium, potassium, and magnesium (ranging from 8.33E-04 to 6.40E-05 M), with slightly higher levels of boron (6.63E-03 M) and sodium (1.62E-03 M). The {sup 137}Cs level is 1.74E-06 Ci L-{sup 1} which corresponds to approximately 4.87E-10 M Cs. The initial Na/Cs ratio was 3.33E+06. The concentration of total strontium is 4.45E-06 M, while the {sup 90}Sr radioactive component was measured to be 6.13E-06 Ci L{sup -1}. Simulant tests were conducted by contacting 0.067 g or each ion exchange material with approximately 100 mL of either the actual or simulated N-Basin water. The simulants contained variable initial cesium concentrations ranging from 1.00E-04 to 2.57E- 10 M Cs while all other components were held constant. For all materials, the average cesium K{sub d} was independent of cesium concentration below approximately 1.0E-06 M. Above this level, the average cesium K{sub d} values decreased significantly. Cesium K{sub d} values exceeding 1.0E+07 mL g{sup -1} were measured in the simulated N-Basin water. However, when measured in the actual N-Basin water the values were several orders of magnitude lower, with a maximum of 1.24E+05 mL g{sup -1} observed.

Brown, G.N.; Carson, K.J.; DesChane, J.R.; Elovich, R.J.

1997-09-01T23:59:59.000Z

90

NUCLEAR REACTOR CONTROL SYSTEM  

DOE Patents (OSTI)

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.

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

1959-11-01T23:59:59.000Z

91

NEUTRONIC REACTOR POWER PLANT  

DOE Patents (OSTI)

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

Metcalf, H.E.

1962-12-25T23:59:59.000Z

92

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

Science Conference Proceedings (OSTI)

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

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

1980-05-30T23:59:59.000Z

93

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

Open Energy Info (EERE)

In The Resurgent Dome Of Long Valley Caldera- Implications From Thermal Data And Deep Electrical Sounding Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal...

94

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

Science Conference Proceedings (OSTI)

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

Not Available

1984-12-01T23:59:59.000Z

95

NEUTRONIC REACTORS  

DOE Patents (OSTI)

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

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

1958-10-14T23:59:59.000Z

96

Three-dimensional Thermal and Airflow (3D-TAF) Model of a Dome...  

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

Three-dimensional Thermal and Airflow (3D-TAF) Model of a Dome-covered House in Canada Speaker(s): Yaolin Lin Date: October 6, 2009 - 12:00pm Location: 90-3122 A dome-covered house...

97

Nuclear reactor  

DOE Patents (OSTI)

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

Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)

1977-01-01T23:59:59.000Z

98

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

Open Energy Info (EERE)

Fluid Flow In The Resurgent Dome Of Long Valley Caldera- Implications From Fluid Flow In The Resurgent Dome Of Long Valley Caldera- Implications From Thermal Data And Deep Electrical Sounding Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Fluid Flow In The Resurgent Dome Of Long Valley Caldera- Implications From Thermal Data And Deep Electrical Sounding Details Activities (5) Areas (1) Regions (0) Abstract: Temperatures of 100°C are measured at 3 km depth in a well located on the resurgent dome in the center of Long Valley Caldera, California, despite an assumed >800°C magma chamber at 6-8 km depth. Local downflow of cold meteoric water as a process for cooling the resurgent dome is ruled out by a Peclet-number analysis of temperature logs. These analyses reveal zones with fluid circulation at the upper and lower

99

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

Science Conference Proceedings (OSTI)

A sensitivity study was performed utilizing a three dimensional finite element model to assess allowable cavern field sizes in 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 a 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.

Ehgartner, Brian L.; Park, Byoung Yoon; Herrick, Courtney Grant

2010-06-01T23:59:59.000Z

100

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

Science Conference Proceedings (OSTI)

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.

Ehgartner, Brian L. (Sandia National Laboratories, Albuquerque, NM); Park, Byoung Yoon

2009-03-01T23:59:59.000Z

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


101

Strategic Petroleum Reserve (SPR) geological site characterization report, Big Hill Salt Dome  

Science Conference Proceedings (OSTI)

Geological and geophysical analyses of the Big Hill Salt Dome were performed to determine the suitability of this site for use in the Strategic Petroleum Reserve (SPR). Development of 140 million barrels (MMB) of storage capacity in the Big Hill Salt Dome is planned as part of the SPR expansion to achieve 750 MMB of storage capacity. Objectives of the study were to: (1) Acquire, evaluate, and interpret existing data pertinent to geological characterization of the Big Hill Dome; (2) Characterize the surface and near-surface geology and hydrology; (3) Characterize the geology and hydrology of the overlying cap rock; (4) Define the geometry and geology of the dome; (5) Determine the feasibility of locating and constructing 14 10-MMB storage caverns in the south portion of the dome; and (6) Assess the effects of natural hazards on the SPR site. Recommendations are included. (DMC)

Hart, R.J.; Ortiz, T.S.; Magorian, T.R.

1981-09-01T23:59:59.000Z

102

Vlf Electromagnetic Investigations Of The Crater And Central Dome Of Mount  

Open Energy Info (EERE)

Vlf Electromagnetic Investigations Of The Crater And Central Dome Of Mount Vlf Electromagnetic Investigations Of The Crater And Central Dome Of Mount St Helens, Washington Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Vlf Electromagnetic Investigations Of The Crater And Central Dome Of Mount St Helens, Washington Details Activities (1) Areas (1) Regions (0) Abstract: A very low frequency (VLF) electromagnetic induction survey in the crater of Mount St. Helens has identified several electrically conductive structures that appear to be associated with thermal anomalies and ground water within the crater. The most interesting of these conductive structures lies beneath the central dome. It is probably a partial melt of dacite similar to that comprising the June 1981 lobe of the central dome. Author(s): James N. Towle

103

Metals removal from spent salts  

DOE Patents (OSTI)

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.

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)

2002-01-01T23:59:59.000Z

104

Actinide removal from spent salts  

DOE Patents (OSTI)

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.

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)

2002-01-01T23:59:59.000Z

105

Ion Removal  

INL’s ion removal technology leverages the ability of phosphazene polymers discriminate between water and metal ions, which allows water to pass ...

106

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

Science Conference Proceedings (OSTI)

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)

Souza Dos Santos, R. [Instituto de Engenharia Nuclear CNEN/IEN, Cidade Universitaria, Rua Helio de Almeida, 75 - Ilha do Fundiao, 21945-970 Rio de Janeiro (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores / CNPq (Brazil)

2012-07-01T23:59:59.000Z

107

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Treshow, M.

1958-08-19T23:59:59.000Z

108

22.312 Engineering of Nuclear Reactors, Fall 2004  

E-Print Network (OSTI)

Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

Buongiorno, Jacopo, 1971-

109

22.312 Engineering of Nuclear Reactors, Fall 2002  

E-Print Network (OSTI)

Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

Todreas, Neil E.

110

Survey of Optimization of Reactor Coolant Cleanup Systems: For Boiling Water Reactors and Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Optimization of the reactor coolant cleanup systems in the boiling water reactor (BWR) and pressurized water reactor (PWR) environment is important for controlling the transport of corrosion products (metals and activated metals), fission products, and coolant impurities (soluble and insoluble) throughout the reactor coolant loop, and this optimization contributes to reducing primary system radiation fields. The removal of radionuclides and corrosion products is just one of many functions (both ...

2013-09-27T23:59:59.000Z

111

CRBRP decay heat removal systems  

SciTech Connect

The Decay Heat Removal Systems for the Clinch River Breeder Reactor Plant (CRBRP) are designed to adequately remove sensible and decay heat from the reactor following normal shutdown, operational occurrences, and postulated accidents on both a short term and a long term basis. The Decay Heat Removal Systems are composed of the Main Heat Transport System, the Main Condenser and Feedwater System, the Steam Generator Auxiliary Heat Removal System (SGAHRS), and the Direct Heat Removal Service (DHRS). The overall design of the CRBRP Decay Heat Removal Systems and the operation under normal and off-normal conditions is examined. The redundancies of the system design, such as the four decay heat removal paths, the emergency diesel power supplies, and the auxiliary feedwater pumps, and the diversities of the design such as forced circulation/natural circulation and AC Power/DC Power are presented. In addition to overall design and system capabilities, the detailed designs for the Protected Air Cooled Condensers (PACC) and the Air Blast Heat Exchangers (ABHX) are presented.

Hottel, R.E.; Louison, R.; Boardman, C.E.; Kiley, M.J.

1977-01-01T23:59:59.000Z

112

The Thermal Regime In The Resurgent Dome Of Long Valley Caldera,  

Open Energy Info (EERE)

Thermal Regime In The Resurgent Dome Of Long Valley Caldera, Thermal Regime In The Resurgent Dome Of Long Valley Caldera, California- Inferences From Precision Temperature Logs In Deep Wells Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: The Thermal Regime In The Resurgent Dome Of Long Valley Caldera, California- Inferences From Precision Temperature Logs In Deep Wells Details Activities (1) Areas (1) Regions (0) Abstract: Long Valley Caldera in eastern California formed 0.76 Ma ago in a cataclysmic eruption that resulted in the deposition of 600 km3 of Bishop Tuff. The total current heat flow from the caldera floor is estimated to be ~ 290 MW, and a geothermal power plant in Casa Diablo on the flanks of the resurgent dome (RD) generates ~40 MWe. The RD in the center of the caldera was uplifted by ~ 80 cm between 1980 and 1999 and was explained by most

113

Dome-Tech Merck Teaming Profile Pres 3-17-05  

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

800 Presented by: TM , CEM, NEBB Vincent Gates, Energy Manager, Merck Keith A. Rinaldi, LEED 2 10-Mar-05 Dome-Tech Field Services We specialize in "hands on" testing, evaluation...

114

740,000-year Deuterium Record in an Ice Core from Dome C, Antarctica  

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

Trace Gases 740,000-year Deuterium Record 740,000-year Deuterium Record in an Ice Core from Dome C, Antarctica graphics Graphics data Data Authors Jean Jouzel Institut...

115

Structural constraints on the exhumation of the Tso Morari Dome, NW Himalaya  

E-Print Network (OSTI)

The Tso Morari culmination in the Ladakh region of northwest India is a large (>3,000 km²) structural dome cored by coesite-bearing rocks of Indian continental crustal affinity. As one of only two localities in the Himalaya ...

Clark, Ryan J

2005-01-01T23:59:59.000Z

116

HORIZONTAL BOILING REACTOR SYSTEM  

DOE Patents (OSTI)

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

Treshow, M.

1958-11-18T23:59:59.000Z

117

Large Component Removal/Disposal  

Science Conference Proceedings (OSTI)

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.

Wheeler, D. M.

2002-02-27T23:59:59.000Z

118

NEUTRONIC REACTOR CHARGING AND DISCHARGING  

DOE Patents (OSTI)

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

Zinn, W.H.

1959-07-14T23:59:59.000Z

119

Features of Bayou Choctaw SPR caverns and internal structure of the salt dome.  

Science Conference Proceedings (OSTI)

The intent of this study is to examine the internal structure of the Bayou Choctaw salt dome utilizing the information obtained from graphical representations of sonar survey data of the internal cavern surfaces. Many of the Bayou Choctaw caverns have been abandoned. Some existing caverns were purchased by the Strategic Petroleum Reserve (SPR) program and have rather convoluted histories and complex cavern geometries. In fact, these caverns are typically poorly documented and are not particularly constructive to this study. Only two Bayou Choctaw caverns, 101 and 102, which were constructed using well-controlled solutioning methods, are well documented. One of these was constructed by the SPR for their use while the other was constructed and traded for another existing cavern. Consequently, compared to the SPR caverns of the West Hackberry and Big Hill domes, it is more difficult to obtain a general impression of the stratigraphy of the dome. Indeed, caverns of Bayou Choctaw show features significantly different than those encountered in the other two SPR facilities. In the number of abandoned caverns, and some of those existing caverns purchased by the SPR, extremely irregular solutioning has occurred. The two SPR constructed caverns suggest that some sections of the caverns may have undergone very regular solutioning to form uniform cylindrical shapes. Although it is not usually productive to speculate, some suggestions that point to the behavior of the Bayou Choctaw dome are examined. Also the primary differences in the Bayou Choctaw dome and the other SPR domes are noted.

Munson, Darrell E.

2007-07-01T23:59:59.000Z

120

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

Science Conference Proceedings (OSTI)

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.``

Not Available

1989-04-01T23:59:59.000Z

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


121

Hanford Deep Dig Removes Contaminated Soil | Department of Energy  

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

Hanford Deep Dig Removes Contaminated Soil Hanford 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 contaminated with sodium dichromate to prevent the chemical from reaching the groundwater and eventually the Columbia River. Workers remove soil contaminated with sodium dichromate to prevent the chemical from reaching the groundwater and eventually the Columbia River.

122

Hanford Deep Dig Removes Contaminated Soil | Department of Energy  

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

Deep Dig Removes Contaminated Soil 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 contaminated with sodium dichromate to prevent the chemical from reaching the groundwater and eventually the Columbia River. Workers remove soil contaminated with sodium dichromate to prevent the chemical from reaching the groundwater and eventually the Columbia River.

123

Nuclear reactor reflector  

DOE Patents (OSTI)

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.

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

1994-01-01T23:59:59.000Z

124

Nuclear reactor reflector  

DOE Patents (OSTI)

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.

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

1994-06-07T23:59:59.000Z

125

Selective purge for hydrogenation reactor recycle loop  

SciTech Connect

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.

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

2001-01-01T23:59:59.000Z

126

Flattening Scientific CCD Imaging Data with a Dome Flat Field System  

E-Print Network (OSTI)

We describe the flattening of scientific CCD imaging data using a dome flat field system. The system uses light emitting diodes (LEDs) to illuminate a carefully constructed dome flat field screen. LEDs have several advantages over more traditional illumination sources: they are available in a wide range of output wavelengths, are inexpensive, have a very long source lifetime, and are straightforward to control digitally. The circular dome screen is made of a material with Lambertian scattering properties that efficiently reflects light of a wide range of wavelengths and incident angles. We compare flat fields obtained using this new system with two types of traditionally-constructed flat fields: twilight sky flats and nighttime sky flats. Using photometric standard stars as illumination sources, we test the quality of each flat field by applying it to a set of standard star observations. We find that the dome flat field system produces flat fields that are superior to twilight or nighttime sky flats, particularly for photometric calibration. We note that a ratio of the twilight sky flat to the nighttime sky flat is flat to within the expected uncertainty; but since both of these flat fields are inferior to the dome flat, this common test is not an appropriate metric for testing a flat field. Rather, the only feasible and correct method for determining the appropriateness of a flat field is to use standard stars to measure the reproducibility of known magnitudes across the detector.

J. L. Marshall; D. L. DePoy

2005-10-07T23:59:59.000Z

127

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

Daniels, F.

1962-12-18T23:59:59.000Z

128

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

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.

Scott, C.D.

1993-12-14T23:59:59.000Z

129

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

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.

Scott, Charles D. (Oak Ridge, TN)

1993-01-01T23:59:59.000Z

130

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

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.

Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

1996-01-01T23:59:59.000Z

131

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

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.

Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

1995-01-01T23:59:59.000Z

132

NEUTRONIC REACTOR FUEL PUMP  

DOE Patents (OSTI)

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

Cobb, W.G.

1959-06-01T23:59:59.000Z

133

Inferences On The Hydrothermal System Beneath The Resurgent Dome In Long  

Open Energy Info (EERE)

Inferences On The Hydrothermal System Beneath The Resurgent Dome In Long 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: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Inferences On The Hydrothermal System Beneath The Resurgent Dome In Long Valley Caldera, East-Central California, Usa, From Recent Pumping Tests And Geochemical Sampling Details Activities (6) Areas (1) Regions (0) Abstract: Quaternary volcanic unrest has provided heat for episodic hydrothermal circulation in the Long Valley caldera, including the present-day hydrothermal system, which has been active over the past 40 kyr. The most recent period of crustal unrest in this region of east-central California began around 1980 and has included periods of

134

Response of the Western Tropical Pacific to the Asian Winter Monsoon: The Generation of the Mindanao Dome  

Science Conference Proceedings (OSTI)

We have investigated the evolution of the Mindanao Dome off the Philippine coast using the GFDL ocean model. It is found that the model's Mindanao Dome evolves in late fall due to local upwelling when a positive curl associated with the northeast ...

Y. Masumoto; T. Yamagata

1991-09-01T23:59:59.000Z

135

RADIATION FACILITY FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

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

1961-12-12T23:59:59.000Z

136

METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS  

DOE Patents (OSTI)

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

Reed, G.A.

1961-01-10T23:59:59.000Z

137

Reactor Thermal-Hydraulics  

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

Thermal-Hydraulics Thermal-Hydraulics Dr. Tanju Sofu, Argonne National Laboratory In a power reactor, the energy produced in fission reaction manifests itself as heat to be removed by a coolant and utilized in a thermodynamic energy conversion cycle to produce electricity. A simplified schematic of a typical nuclear power plant is shown in the diagram below. Primary coolant loop Steam Reactor Heat exchanger Primary pump Secondary pump Condenser Turbine Water Although this process is essentially the same as in any other steam plant configuration, the power density in a nuclear reactor core is typically four orders of magnitude higher than a fossil fueled plant and therefore it poses significant heat transfer challenges. Maximum power that can be obtained from a nuclear reactor is often limited by the

138

Nuclear Reactors  

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

Reactors Nuclear reactors created not only large amounts of plutonium needed for the weapons programs, but a variety of other interesting and useful radioisotopes. They produced...

139

Liquid metal cooled nuclear reactor plant system  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

140

Three-dimensional Thermal and Airflow (3D-TAF) Model of a Dome-covered  

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

Three-dimensional Thermal and Airflow (3D-TAF) Model of a Dome-covered Three-dimensional Thermal and Airflow (3D-TAF) Model of a Dome-covered House in Canada Speaker(s): Yaolin Lin Date: October 6, 2009 - 12:00pm Location: 90-3122 A dome-covered house is an example of sustainable design that draws from biological forms in nature. A three-dimensional thermal and air flow (3D-TAF) model was developed to estimate the energy needs of a dome-covered house. This model has two components: a thermal model to calculate the temperature; and an air flow model to find the velocities, which are needed to estimate the surface convection. The two models are solved iteratively at every time step until they converge. I will present the numerical methods for solving the mathematical models, and compared the results with other simulated and experimental results from similar structures. I will

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


141

Almost a Century of “Imaging” Clouds Over the Whole-Sky Dome  

Science Conference Proceedings (OSTI)

Meteorological whole-sky photography can be traced back to just after the turn of the century. Capturing an objective and well-determined view of the cloud cover over the whole-sky dome has been one of the principal goals of subsequent ...

K. McGuffe; A. Henderson-Sellers

1989-10-01T23:59:59.000Z

142

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

DOE Patents (OSTI)

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

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

1991-01-01T23:59:59.000Z

143

Slurry reactor design studies  

SciTech Connect

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.

Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

1990-06-01T23:59:59.000Z

144

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab...  

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

Brookhaven Lab American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once...

145

Nitrogen Removal in Aerobic Granular Sludge SBR: Real?time Control Strategies  

Science Conference Proceedings (OSTI)

A sequencing batch reactor (SBR) with aerobic granules was operated to determine the effect of different DO concentration on biological nitrogen removal for synthetic sewage treatment

Xiangjuan Yuan; Dawen Gao

2010-01-01T23:59:59.000Z

146

Natural Convection Shutdown Heat Removal Test Facility (NSTF)  

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

Natural Convection Natural Convection Shutdown Heat Removal Test Facility Scaling Basis Full Scale Half Scale NSTF Argonne National Laboratory's Natural Convection Shutdown Heat Removal Test Facility (NSTF) - one of the world's largest facilities for ex-vessel passive decay heat removal testing-confirms the performance of reactor cavity cooling systems (RCCS) and similar passive confinement or containment decay heat removal systems in modern Small Modular Reactors. Originally built to aid in the development of General Electric's Power Reactor Innovative Small Module (PRISM) Reactor Vessel Auxiliary Cooling System (RVACS), the NSTF has a long history of providing confirmatory data for the airside of the RVACS. Argonne National Laboratory's NSTF is a state-of-the-art, large-scale facility for evaluating performance

147

Effective Boron Removal by Calcium Silicate Slags Combined with ...  

Science Conference Proceedings (OSTI)

A New Centrifuge CVD Reactor that will Challenge the Siemens Process ... Boron Removal from Silicon Melts by H2O/H2 Gas Blowing – Gas-phase Mass ...

148

Removal of Phosphor in Metallurgical Silicon by Rare Earth Elements  

Science Conference Proceedings (OSTI)

A New Centrifuge CVD Reactor that will Challenge the Siemens Process ... Boron Removal from Silicon Melts by H2O/H2 Gas Blowing – Gas-phase Mass ...

149

Research on the Forecast Model of the Boron Removal from ...  

Science Conference Proceedings (OSTI)

A New Centrifuge CVD Reactor that will Challenge the Siemens Process ... Boron Removal from Silicon Melts by H2O/H2 Gas Blowing – Gas-phase Mass ...

150

Program on Technology Innovation: Feasibility Assessment of a Core Vacuum for Foreign Material and Activity Removal  

Science Conference Proceedings (OSTI)

The need for increased fuel reliability and radioactive source term reduction motivated EPRI to investigate methods for removing small foreign material and activated corrosion products from reactor vessels. Several methods exist to remove these materials from above the core plate of the reactor vessel, but there has been limited research and development of techniques to remove them from underneath the core plate. This report investigates the development of a core vacuum to remove debris and corrosion pro...

2008-09-30T23:59:59.000Z

151

Preliminary geological and geophysical evaluation of the Castle Dome HDR geothermal prospect, Southwestern Arizona  

DOE Green Energy (OSTI)

The Castle Dome HDR geothermal prospect is located in Yuma County, Arizona, in a region centered about 80 km north of Yuma along US Rte. 95. The area of interest is broadly defined by a negative residual Bouguer gravity anomaly which is about 45 km across, steep-sided in many places, and as much as 30 mgals in magnitude. The geology of this Basin and Range area is poorly known, but the few published reports and current Los Alamos Scientific Laboratory (LASL) field studies indicate that the Castle Dome Mountains and adjacent ranges are chiefly a thick pile of welded ash-flow tuffs of probable mid-Tertiary age. The tuffs rest unconformably on Mesozoic metasedimentary rocks exposed only outside steep edges of the gravity low. This gravity anomaly may reflect the presence of a large caldera. A regional magnetotelluric study now in progress will define the depths to electrical conductors within the crust and upper mantle and contribute to understanding of crustal structure, the gravity anomaly, and the Hot Dry Rock (HDR) geothermal potential of the Castle Dome area.

Gutmann, J.T.; Aiken, C.L.V.; Ander, M.E.; Laney, R.T.

1980-01-01T23:59:59.000Z

152

Preliminary long-term stability criteria for compressed air energy storage caverns in salt domes  

DOE Green Energy (OSTI)

Air storage caverns, which are an essential and integral component of a CAES plant, should be designed and operated so as to perform satisfactorily over the intended life of the overall facility. It follows that the long-term ''stability'' of air storage caverns must be considered as a primary concern in projecting the satisfactory operation of CAES facilities. As used in the report, ''stability'' of a storage cavern implies the extent to which an acceptable amount of cavern storage volume can be utilized with routine maintenance for a specified time interval, e.g., 35 years. In this context, cavern stability is relative to both planned utilization and time interval of operation. The objective of the study was to review the existing literature and consult knowledgeable workers in the storage industry, and then report state-of-the-art findings relative to long-term stability of compressed air energy storage caverns in salt domes. Further, preliminary cavern stability criteria were to be presented in a form consistent with the amount of information available on cavern performance in salt domes. Another objective of the study was to outline a methodology for determining the long-term stability of site-specific CAES cavern systems in salt domes.

Thoms, R.L.; Martinez, J.D.

1978-08-01T23:59:59.000Z

153

Multi-point interactions with immersive omnidirectional visualizations in a dome  

E-Print Network (OSTI)

video-conferencing; c) astronomical data from the World Wide Telescope application; and d) a multi-player game. This paper describes an interactive immersive experience using mid-air gestures to interact with a large curved display: a projected dome. Our Pinch-the-Sky Dome is an immersive installation where several users can interact simultaneously with omnidirectional data using freehand gestures. The system consists of a single centrally-located omnidirectional projector-camera unit where the projector is able to project an image spanning the entire 360 degrees and a camera is used to track gestures for navigation of the content. We combine speech commands with freehand pinch and clasping gestures and infra-red laser pointers to provide a highly immersive and interactive experience to several users inside the dome, with a very wide field of view for each user. The interactive applications include: 1) the astronomical data exploration, 2) social networking 3D graph visualizations, 3) immersive panoramic images, 4) 360 degree video conferencing, 5) a drawing canvas, and 6) a multi-user interactive game. Finally, we discuss the user reactions and feedback from two demo events where more than 1000 people had the chance to experience our work. ACM Classification: H5.2 [Information interfaces and presentation]: User Interfaces – Input devices and strategies: Graphical user interfaces.

Hrvoje Benko; Andrew D. Wilson

2010-01-01T23:59:59.000Z

154

NUCLEAR REACTOR  

DOE Patents (OSTI)

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

Treshow, M.

1961-09-01T23:59:59.000Z

155

Fuel Reformation: Microchannel Reactor Design  

DOE Green Energy (OSTI)

Fuel processing is used to extract hydrogen from conventional vehicle fuel and allow fuel cell powered vehicles to use the existing petroleum fuel infrastructure. Kilowatt scale micro-channel steam reforming, water-gas shift and preferential oxida-tion reactors have been developed capable of achieving DOE required system performance metrics. Use of a microchannel design effectively supplies heat to the highly endothermic steam reforming reactor to maintain high conversions, controls the temperature profile for the exothermic water gas shift reactor, which optimizes the overall reaction conversion, and removes heat to prevent the unwanted hydrogen oxidation in the prefer-ential oxidation reactor. The reactors combined with micro-channel heat exchangers, when scaled to a full sized 50 kWe automotive system, will be less than 21 L in volume and 52 kg in weight.

Brooks, Kriston P.; Davis, James M.; Fischer, Christopher M.; King, David L.; Pederson, Larry R.; Rawlings, Gregg C.; Stenkamp, Victoria S.; TeGrotenhuis, Ward E.; Wegeng, Robert S.; Whyatt, Greg A.

2005-09-01T23:59:59.000Z

156

CONVECTION REACTOR  

DOE Patents (OSTI)

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.

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

1960-03-22T23:59:59.000Z

157

Fusion reactor design studies. [ARIES Tokamak  

SciTech Connect

This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources. (LSP)

Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

1990-10-12T23:59:59.000Z

158

Liquid metal cooled nuclear reactors with passive cooling system  

SciTech Connect

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

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

1991-01-01T23:59:59.000Z

159

Public comment sought on final end state of Experimental Breeder Reactor-II  

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

Media Contacts: Danielle Miller, 208-526-5709 Media Contacts: Danielle Miller, 208-526-5709 Joseph Campbell, CWI, 208-360-0142 Public comment sought on final end state of Experimental Breeder Reactor-II The U.S. Department of Energy (DOE) is seeking public comment on a range of alternatives for disposition of the landmark Experimental Breeder Reactor-II (EBR-II) building and reactor vessel at the Idaho Site's Materials and Fuels Complex. An Engineering Evaluation/Cost Analysis (EE/CA) document with four proposed alternatives for the final end state of the reactor facility and support structures is currently under evaluation by DOE, the U.S. Environmental Protection Agency, and Idaho's Department of Environmental Quality. Experimental Breeder Reactor-II containment dome The EBR-II was an innovative sodium-cooled reactor with an output of 62

160

Brookhaven Lab Completes Decommissioning of Graphite Research Reactor:  

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

Brookhaven Lab Completes Decommissioning of Graphite Research 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 Brookhaven Graphite Research Reactor’s bioshield, which contains the 700-ton reactor core, is shown prior to decommissioning. The Brookhaven Graphite Research Reactor's bioshield, which contains the 700-ton reactor core, is shown prior to decommissioning. Pictured here is the Brookhaven Graphite Research Reactor, where major decommissioning milestones were recently reached after the remaining radioactive materials from the facility’s bioshield were shipped to a licensed offsite disposal facility.

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


161

Capillary-Pumped Passive Reactor Concept for Space Nuclear Power  

Science Conference Proceedings (OSTI)

To develop the passively-cooled space reactor concept using the capillary-induced lithium flow, since molten lithium possesses a very favorable surface tension characteristic. In space where the gravitational field is minimal, the gravity-assisted natural convection cooling is not effective nor an option for reactor heat removal, the capillary induced cooling becomes an attractive means of providing reactor cooling.

Dr. Thomas F. Lin; Dr. Thomas G. Hughes; Christopher G. Miller

2008-05-30T23:59:59.000Z

162

Passive cooling safety system for liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

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

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

1991-01-01T23:59:59.000Z

163

Indirect passive cooling system for liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

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

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

1990-01-01T23:59:59.000Z

164

Packed fluidized bed blanket for fusion reactor  

DOE Patents (OSTI)

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.

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

1984-01-01T23:59:59.000Z

165

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites  

SciTech Connect

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.

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

2013-09-30T23:59:59.000Z

166

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

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

1961-11-21T23:59:59.000Z

167

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

Science Conference Proceedings (OSTI)

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.

Neal, J.T. [Sandia National Labs., Albuquerque, NM (United States); Magorian, T.R. [Magorian (Thomas R.), Amherst, NY (United States); Byrne, K.O.; Denzler, S. [Acres International Corp., Amherst, NY (United States)

1993-09-01T23:59:59.000Z

168

REACTOR COOLING  

DOE Patents (OSTI)

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

Quackenbush, C.F.

1959-09-29T23:59:59.000Z

169

Turbomachinery debris remover  

DOE Patents (OSTI)

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

Krawiec, Donald F. (Pittsburgh, PA); Kraf, Robert J. (North Huntingdon, PA); Houser, Robert J. (Monroeville, PA)

1988-01-01T23:59:59.000Z

170

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

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

Brookhaven Graphite Research Reactor Documents Brookhaven Graphite Research Reactor Documents Feasibility Study (PDF) Proposed Remedial Action Plan (PDF) Record of Decision (PDF) RD/RA Work Plan for the BGRR Pile (PDF) RD/RA Work Plan for the Bioshield (PDF) RD/RA Work Plan for the BGRR Cap (PDF) Brookhaven Graphite Research Reactor Explanation of Significant Differences (PDF) (4/12) NYSDEC Approval Letter for BGRR ESD (PDF) (5/12) USEPA Approval Letter for BGRR ESD (PDF) (6/12) DOE BGRR ESD Transmittal Letter (PDF) (7/12) Remedial Design Implementation Report (PDF) (12/11) Completion Reports Removal of the Above-Ground Ducts and Preparation of the Instrument House (708) for Removal (PDF) - April 2002 Below-Ground Duct Outlet Air Coolers, Filters and Primary Liner Removal (PDF) - April 2005 Canal and Deep Soil Pockets Excavation and Removal (PDF) - August

171

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

Wigner, E.P.

1958-04-22T23:59:59.000Z

172

HOMOGENEOUS NUCLEAR REACTOR  

SciTech Connect

This homogeneous reactor comprises a core occupied by a solution of a fissile material in a moderator liquid and a breeder region enclosing the core and having a suspension of fertile material in the same moderator liquid. There is communication between the core and breeder to allow mass transfer and pressure equalization between the regions. The zones each have a separate circuit for removing heat by a mixer chamber situated inside the reactor vessel. The effluents coming from the two regions are mixed and led to a common device for separation into a clear solution and suspension, which are each led back to its corresponding circuit. To control the relative concentration of the two regions, an evaporator is provided separating a part of the moderator liquid from the solution occupying the core, the condensed separated moderator liquid being led into the breeder region. (NPO)

1960-07-11T23:59:59.000Z

173

Reactor core isolation cooling system  

DOE Patents (OSTI)

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.

Cooke, F.E.

1992-12-08T23:59:59.000Z

174

Reactor core isolation cooling system  

DOE Patents (OSTI)

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.

Cooke, Franklin E. (San Jose, CA)

1992-01-01T23:59:59.000Z

175

Potential radiological exposure rates resulting from hypothetical dome failure at Tank W-10  

SciTech Connect

The main plant area at Oak Ridge National Laboratory (ORNL) contains 12 buried Gunite tanks that were used for the storage and transfer of liquid radioactive waste. Although the tanks are no longer in use, they are known to contain some residual contaminated sludges and liquids. In the event of an accidental tank dome failure, however unlikely, the liquids, sludges, and radioactive contaminants within the tank walls themselves could create radiation fields and result in above-background exposures to workers nearby. This Technical Memorandum documents a series of calculations to estimate potential radiological exposure rates and total exposures to workers in the event of a hypothetical collapse of a Gunite tank dome. Calculations were performed specifically for tank W-10 because it contains the largest radioactivity inventory (approximately half of the total activity) of all the Gunite tanks. These calculations focus only on external, direct gamma exposures for prescribed, hypothetical exposure scenarios and do not address other possible tank failure modes or routes of exposure. The calculations were performed with established, point-kernel gamma ray modeling codes.

Not Available

1994-07-01T23:59:59.000Z

176

Dome takes a 20% interest in the Arctic pilot project to move LNG  

Science Conference Proceedings (OSTI)

According to B. Richards of Dome Petroleum Ltd., Dome's interest will be shared with its partially owned subsidiary, Trans-Canada Pipe Lines Ltd. According to J. Bell of Petro-Canada, the operator for the Arctic project, negotiations are under way with Tenneco Inc. for gas sales of up to 225 million cu ft/day to begin in 1985-86. At first, two tankers would ship LNG to a delivery terminal at an as yet unselected site on Canada's east coast, but by 1992, nine ships capable of delivering 1.23 billion cu ft/day of LNG, could be in service. The U.S. and European potential LNG markets amounts to 3-4 trillion cu ft/yr and 3.5-4 trillion cu ft/yr, respectively. Petro-Canada also supports the Polar Gas Ltd. project to lay a gas pipeline from the Arctic Islands and Mackenzie Delta to the south; the projects are not considered to be in competition.

Richards, B.; Bell, J.

1980-05-05T23:59:59.000Z

177

Geologic framework and hot dry rock geothermal potential of the Castle Dome area, Yuma County, Arizona  

DOE Green Energy (OSTI)

The Castle Dome Mountains and surrounding ranges constitute a voluminous pile of silicic volcanic rocks within the Basin and Range province of southwestern Arizona. Previously reported as Cretaceous and Quaternary in age, these volcanics all are of late Oligocene to early Miocene age as indicated by five new K-Ar dates. Reconnaissance field studies indicate that the volcanic section locally has undergone large rotations that contrast with the usual structural style of the Basin and Range and resemble the thin-skinned rotational tectonics documented for earlier, mid-Tertiary extensional deformation in ranges to the north and northeast. Significant geothermal potential of the Castle Dome area is suggested by a shallow depth to the Curie isotherm and by the apparent presence of a good electrical conductor at anomalously shallow depth in the crust. Warm wells exist in the area and Shearer (1979) reported a geothermal gradient of about 70/sup 0/C/km in a dry well near the center of the gravity low. Radiogenic heat production in the silicic batholith inferred above constitutes a reasonable candidate for a shallow regional heat source.

Gutmann, J.T.

1981-02-01T23:59:59.000Z

178

Heat dissipating nuclear reactor with metal liner  

DOE Patents (OSTI)

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.

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

1987-01-01T23:59:59.000Z

179

Heat dissipating nuclear reactor with metal liner  

DOE Patents (OSTI)

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.

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

1985-11-21T23:59:59.000Z

180

TA-2 Water Boiler Reactor Decommissioning Project  

Science Conference Proceedings (OSTI)

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

Durbin, M.E. (ed.); Montoya, G.M.

1991-06-01T23:59:59.000Z

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


181

Graphitic packing removal tool  

DOE Patents (OSTI)

Graphitic packing removal tools are described for removal of the seal rings in one piece from valves and pumps. 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.

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

1996-12-31T23:59:59.000Z

182

Reactor refueling containment system  

DOE Patents (OSTI)

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.

Gillett, J.E.; Meuschke, R.E.

1995-05-02T23:59:59.000Z

183

NEUTRONIC REACTOR CONSTRUCTION  

DOE Patents (OSTI)

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

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

1958-09-01T23:59:59.000Z

184

Reactor refueling containment system  

DOE Patents (OSTI)

This report describes 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.

Gillett, J.E.; Meuschke, R.E.

1992-12-31T23:59:59.000Z

185

FAST NEUTRONIC REACTOR  

DOE Patents (OSTI)

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.

Snell, A.H.

1957-12-01T23:59:59.000Z

186

EM Employs Innovative Technology to Remove Radioactive Sludge | Department  

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

Employs Innovative Technology to Remove Radioactive Sludge Employs Innovative Technology to Remove Radioactive Sludge EM Employs Innovative Technology to Remove Radioactive Sludge September 1, 2012 - 12:00pm Addthis Testing and equipment simulations ensure first-of-a-kind technological processes for sludge removal can be conducted safely and efficiently. Testing and equipment simulations ensure first-of-a-kind technological processes for sludge removal can be conducted safely and efficiently. RICHLAND, Wash. - The Richland Operations Office and contractor CH2M HILL Plateau Remediation Company successfully removed a portion of a highly radioactive sludge from underwater storage in a large basin adjacent to the K West reactor at the Hanford site this month. In that milestone, workers removed sludge originating from knock-out pots,

187

NETL: Gasification Systems - Warm Gas Multi-Contaminant Removal System  

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

Warm Gas Multi-Contaminant Removal System Warm Gas Multi-Contaminant Removal System Project Number: DE-SC00008243 TDA Research, Inc. is developing a high-capacity, low-cost sorbent that removes anhydrous ammonia (NH3), mercury (Hg), and trace contaminants from coal- and coal/biomass-derived syngas. The clean-up system will be used after the bulk warm gas sulfur removal step, and remove NH3 and Hg in a regenerable manner while irreversibly capturing all other trace metals (e.g., Arsenic, Selenium) reducing their concentrations to sub parts per million (ppm) levels. Current project plans include identifying optimum chemical composition and structure that provide the best sorbent performance for removing trace contaminants, determining the effect of operating parameters, conducting multiple-cycle experiments to test the life of the sorbent for NH3 and Hg removal, and conducting a preliminary design of the sorbent reactor.

188

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

DOE Patents (OSTI)

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.

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

1982-08-19T23:59:59.000Z

189

Process to remove rare earth from IFR electrolyte  

DOE Patents (OSTI)

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.

Ackerman, John P. (Downers Grove, IL); Johnson, Terry R. (Wheaton, IL)

1994-01-01T23:59:59.000Z

190

Process to remove rare earth from IFR electrolyte  

DOE Patents (OSTI)

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.

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

1994-08-09T23:59:59.000Z

191

Process to remove rare earth from IFR electrolyte  

DOE Patents (OSTI)

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.

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

1992-01-01T23:59:59.000Z

192

NUCLEAR REACTOR  

DOE Patents (OSTI)

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.

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

1958-12-01T23:59:59.000Z

193

Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor (HFIR) using RELAP5 and TEMPEST: Part 2, Interpretation and validation of results  

SciTech Connect

The RELAP5/MOD2 code was used to predict the thermal-hydraulic behavior of the HFIR core during decay heat removal through boiling natural circulation. The low system pressure and low mass flux values associated with boiling natural circulation are far from conditions for which RELAP5 is well exercised. Therefore, some simple hand calculations are used herein to establish the physics of the results. The interpretation and validation effort is divided between the time average flow conditions and the time varying flow conditions. The time average flow conditions are evaluated using a lumped parameter model and heat balance. The Martinelli-Nelson correlations are used to model the two-phase pressure drop and void fraction vs flow quality relationship within the core region. Systems of parallel channels are susceptible to both density wave oscillations and pressure drop oscillations. Periodic variations in the mass flux and exit flow quality of individual core channels are predicted by RELAP5. These oscillations are consistent with those observed experimentally and are of the density wave type. The impact of the time varying flow properties on local wall superheat is bounded herein. The conditions necessary for Ledinegg flow excursions are identified. These conditions do not fall within the envelope of decay heat levels relevant to HFIR in boiling natural circulation. 14 refs., 5 figs., 1 tab.

Ruggles, A.E.; Morris, D.G.

1989-01-01T23:59:59.000Z

194

Reservoir simulation of co2 sequestration and enhanced oil recovery in Tensleep Formation, Teapot Dome field  

E-Print Network (OSTI)

Teapot Dome field is located 35 miles north of Casper, Wyoming in Natrona County. This field has been selected by the U.S. Department of Energy to implement a field-size CO2 storage project. With a projected storage of 2.6 million tons of carbon dioxide a year under fully operational conditions in 2006, the multiple-partner Teapot Dome project could be one of the world's largest CO2 storage sites. CO2 injection has been used for decades to improve oil recovery from depleted hydrocarbon reservoirs. In the CO2 sequestration technique, the aim is to "co-optimize" CO2 storage and oil recovery. In order to achieve the goal of CO2 sequestration, this study uses reservoir simulation to predict the amount of CO2 that can be stored in the Tensleep Formation and the amount of oil that can be produced as a side benefit of CO2 injection. This research discusses the effects of using different reservoir fluid models from EOS regression and fracture permeability in dual porosity models on enhanced oil recovery and CO2 storage in the Tensleep Formation. Oil and gas production behavior obtained from the fluid models were completely different. Fully compositional and pseudo-miscible black oil fluid models were tested in a quarter of a five spot pattern. Compositional fluid model is more convenient for enhanced oil recovery evaluation. Detailed reservoir characterization was performed to represent the complex characteristics of the reservoir. A 3D black oil reservoir simulation model was used to evaluate the effects of fractures in reservoir fluids production. Single porosity simulation model results were compared with those from the dual porosity model. Based on the results obtained from each simulation model, it has been concluded that the pseudo-miscible model can not be used to represent the CO2 injection process in Teapot Dome. Dual porosity models with variable fracture permeability provided a better reproduction of oil and water rates in the highly fractured Tensleep Formation.

Gaviria Garcia, Ricardo

2005-12-01T23:59:59.000Z

195

Reactor Materials  

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

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...

196

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

1962-10-23T23:59:59.000Z

197

NEUTRONIC REACTORS  

DOE Patents (OSTI)

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.

Wigner, E.P.

1960-11-22T23:59:59.000Z

198

REACTOR SHIELD  

DOE Patents (OSTI)

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

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

1959-02-17T23:59:59.000Z

199

NUCLEAR REACTOR  

DOE Patents (OSTI)

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.

Grebe, J.J.

1959-07-14T23:59:59.000Z

200

Solid State Reactor Final Report  

DOE Green Energy (OSTI)

The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas of research were undertaken: (1) establishing the design and safety-related basis via neutronic and reactor control assessments with the graphite foam as heat transfer medium; (2) evaluating the thermal performance of the graphite foam for heat removal, reactor stability, reactor operations, and overall core thermal characteristics; (3) characterizing the physical properties of the graphite foam under normal and irradiated conditions to determine any effects on structure, dimensional stability, thermal conductivity, and thermal expansion; and (4) developing a power conversion system design to match the reactor operating parameters.

Mays, G.T.

2004-03-10T23:59:59.000Z

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


201

Dome-Tech Merck Teaming Profile Pres 3-17-05  

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

, 2005 , 2005 Energy Conservation Through Retro- Commissioning of Building 800 Presented by: TM , CEM, NEBB Vincent Gates, Energy Manager, Merck Keith A. Rinaldi, LEED 2 10-Mar-05 Dome-Tech Field Services We specialize in "hands on" testing, evaluation and Retro-Cx of existing HVAC and utility systems. We identify and solve problems, reduce energy expenses and improve indoor air quality. We save companies significant money without requiring large capital outlays. 3 10-Mar-05 Building 800 - Background * Multi-Science Building * Constructed and occupied in 2000 * 325,000 square feet * Consists of R&D laboratories, support spaces and administrative offices * Sixteen air handling units with VFD's * Sixteen constant volume exhaust fans * Three 200 HP chilled water pumps

202

ON THE NATURE OF RECONNECTION AT A SOLAR CORONAL NULL POINT ABOVE A SEPARATRIX DOME  

SciTech Connect

Three-dimensional magnetic null points are ubiquitous in the solar corona and in any generic mixed-polarity magnetic field. We consider magnetic reconnection at an isolated coronal null point whose fan field lines form a dome structure. Using analytical and computational models, we demonstrate several features of spine-fan reconnection at such a null, including the fact that substantial magnetic flux transfer from one region of field line connectivity to another can occur. The flux transfer occurs across the current sheet that forms around the null point during spine-fan reconnection, and there is no separator present. Also, flipping of magnetic field lines takes place in a manner similar to that observed in the quasi-separatrix layer or slip-running reconnection.

Pontin, D. I. [Division of Mathematics, University of Dundee, Dundee DD1 4HN (United Kingdom); Priest, E. R. [School of Mathematics and Statistics, University of St Andrews, Fife KY16 9SS (United Kingdom); Galsgaard, K., E-mail: dpontin@maths.dundee.ac.uk [Niels Bohr Institute, Copenhagen DK-2100 (Denmark)

2013-09-10T23:59:59.000Z

203

Microsoft Word - RMOTC Partners Honored for Teapot Dome Technology Test.docx  

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

October 30, 2008 The Rocky Mountain Oilfield Testing Center (RMOTC) is providing the following information on local activities: RMOTC: PARTNERS HONORED FOR TEAPOT DOME TECHNOLOGY TEST Casper, Wyoming - Two partners of the Rocky Mountain Oilfield Testing Center (RMOTC) were honored at the 2008 Federal Laboratory Consortium (FLC) Mid-Continent Region meeting in Denver, Colo in September. WhisperGen LLC of New Zealand and BP America shared an Excellence in Technology Transfer award for their combined efforts in testing Stirling Cycle electrical generators for use at remote wellsites and the wide dissemination of those test results to the oil and gas industry. Stirling Cycle engines are external combustion engines which offer advantages over traditional

204

STATEMENT OF CONSID ERATIONS CLASS WAIVER OF THE GOVERNMENT'S DOMES'I'!C: AND FO  

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

CONSID CONSID ERATIONS CLASS WAIVER OF THE GOVERNMENT'S DOMES'I'!C: AND FO REIGN PATENT RIGHTS AND ALLOGATION OF DATA RIGHTS ARlSlNG FROM THE USE OF DOE FACI L!TI ES ANP FACILITY CON1'RACTORS BY OR FOR Tl !IRD-PARTY SPONSORS: DOE W A.IVER NO. W(C)-2011-009. Introduction The Deparlrnenr of Energy (and its predecessor agencies) (collectively. '·DOE'' or "Department") considers each of its DOE Facilities (i.e .. National Laboratori es, single-purpose research facilities, and other Department faci lities hereinafter referred to individually as '·Facility'' or collectively as "Facilities'') a unique and val uable national resource that should be made available to the extent feas ible fo r non- Federa l research and development activities and studies for third-party Sponsors.

205

Teapot Dome: Site Characterization of a CO2- Enhanced Oil Recovery Site in Eastern Wyoming  

Science Conference Proceedings (OSTI)

Naval Petroleum Reserve No. 3 (NPR-3), better known as the Teapot Dome oil field, is the last U.S. federally-owned and -operated oil field. This provides a unique opportunity for experiments to provide scientific and technical insight into CO{sub 2}-enhanced oil recovery (EOR) and other topics involving subsurface fluid behavior. Towards that end, a combination of federal, academic, and industrial support has produced outstanding characterizations of important oil- and brine-bearing reservoirs there. This effort provides an unparalleled opportunity for industry and others to use the site. Data sets include geological, geophysical, geochemical, geomechanical, and operational data over a wide range of geological boundary conditions. Importantly, these data, many in digital form, are available in the public domain due to NPR-3's federal status. Many institutions are already using portions of the Teapot Dome data set as the basis for a variety of geoscience, modeling, and other research efforts. Fifteen units, 9 oil-bearing and 6 brine-bearing, have been studied to varying degrees. Over 1200 wells in the field are active or accessible, and over 400 of these penetrate 11 formations located below the depth that corresponds to the supercritical point for CO{sub 2}. Studies include siliciclastic and carbonate reservoirs; shale, carbonate, and anhydrite cap rocks; fractured and unfractured units; and over-pressured and under-pressured zones. Geophysical data include 3D seismic and vertical seismic profiles. Reservoir data include stratigraphic, sedimentological, petrologic, petrographic, porosity, and permeability data. These have served as the basis for preliminary 3D flow simulations. Geomechanical data include fractures (natural and drilling induced), in-situ stress determination, pressure, and production history. Geochemical data include soil gas, noble gas, organic, and other measures. The conditions of these reservoirs directly or indirectly represent many reservoirs in the U.S., Canada, and overseas.

Friedmann, S J; Stamp, V

2005-11-01T23:59:59.000Z

206

Anomalous zones in Gulf Coast Salt domes with special reference to Big Hill, TX, and Weeks Island, LA  

Science Conference Proceedings (OSTI)

Anomalous features in Gulf Coast Salt domes exhibit deviations from normally pure salt and vary widely in form from one dome to the next, ranging considerably in length and width. They have affected both conventional and solution mining in several ways. Gas outbursts, insolubles, and potash (especially carnallite) have led to the breakage of tubing in a number of caverns, and caused irregular shapes of many caverns through preferential leaching. Such anomalous features essentially have limited the lateral extent of conventional mining at several salt mines, and led to accidents and even the closing of several other mines. Such anomalous features, are often aligned in anomalous zones, and appear to be related to diapiric processes of salt dome development. Evidence indicates that anomalous zones are found between salt spines, where the differential salt intrusion accumulates other materials: Anhydrite bands which are relatively strong, and other, weaker impurities. Shear zones and fault displacement detected at Big Hill and Weeks Island domes have not yet had any known adverse impacts on SPR oil storage, but new caverns at these sites conceivably may encounter some potentially adverse conditions. Seismic reflection profiles at Big Hill dome have shown numerous fractures and faults in the caprock, and verified the earlier recognition of a major shear zone transecting the entire salt stock and forming a graben in the overlying caprock. Casing that is placed in such zones can be at risk. Knowledge of these zones should create awareness of possible effects rather than preclude the future emplacement of caverns. To the extent possible, major anomalous zones and salt stock boundaries should be avoided. Shear zones along overhangs may be particularly hazardous, and otherwise unknown valleys in the top of salt may occur along shear zones. These zones often can be mapped geophysically, especially with high-resolution seismic techniques.

Neal, J.T. [Sandia National Labs., Albuquerque, NM (United States); Magorian, T.R. [Magorian (Thomas R.), Amherst, NY (United States); Thoms, R.L. [AGM, Inc., College Station, TX (United States); Autin, W.J.; McCulloh, R.P. [Louisiana Geological Survey, Baton Rouge, LA (United States); Denzler, S.; Byrne, K.O. [Acres International Corp., Amherst, NY (United States)

1993-07-01T23:59:59.000Z

207

Research reactors - an overview  

SciTech Connect

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

West, C.D.

1997-03-01T23:59:59.000Z

208

Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites |  

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

Evaluation of Removing Used Nuclear Fuel From Shutdown 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 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. Shutdown sites are defined as those commercial nuclear power reactor sites where the

209

Mechanical cutting of irradiated reactor internal components  

Science Conference Proceedings (OSTI)

Mechanical cutting methods to volume reduce and package reactor internal components are now a viable solution for stakeholders challenged with the retirement of first generation nuclear facilities. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods, inclusive of plasma arc and abrasive water-jet cutting, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. Reactor internal components were segmented, packaged, and removed from the reactor building for shipment or storage, allowing the reactor cavity to be drained and follow-on reactor segmentation activities to proceed in the dry state. Area exposure rates at the work positions during the segmentation process were generally 1 mR per hr. Radiological exposure documented for the underwater segmentation processes totaled 13 person rem. The reactor internals weighing 343,000 pounds were segmented into over 200 pieces for maximum shipping package efficiency and produced 5,600 lb of stainless steel chips and shavings which were packaged in void spaces of existing disposal containers, therefore creating no additional disposal volume. Because no secondary waste was driven into suspension in the reactor cavity water, the water was free released after one pass through a charcoal bed and ion exchange filter system. Mechanical cutting techniques are capable of underwater segmentation of highly radioactive components on a large scale. This method minimized radiological exposure and costly water cleanup while creating no secondary waste.

Anderson, Michael G. [MOTA Corporation: 3410 Sunset Boulevard, West Columbia, SC, 29169 (United States)

2008-01-15T23:59:59.000Z

210

Device for removing blackheads  

DOE Patents (OSTI)

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.

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

1995-03-07T23:59:59.000Z

211

Brookhaven Medical Research Reactor  

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

Medical Research Reactor BMRR The last of the Lab's reactors, the Brookhaven Medical Research Reactor (BMRR), was shut down in December 2000. The BMRR was a three megawatt...

212

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

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

1958-09-01T23:59:59.000Z

213

Silica Scaling Removal Process  

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

Silica Scaling Removal Process Silica Scaling Removal Process Silica 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. Available for thumbnail of Feynman Center (505) 665-9090 Email Silica Scaling Removal Process Applications: Cooling tower systems Water treatment systems Water evaporation systems Potential mining applications (produced water) Industry applications for which silica scaling must be prevented Benefits: Reduces scaling in cooling towers by up to 50% Increases the number of cycles of concentration substantially Reduces the amount of antiscaling chemical additives needed Decreases the amount of makeup water and subsequent discharged water (blowdown) Enables considerable cost savings derived from reductions in

214

POWER REACTOR  

DOE Patents (OSTI)

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

Zinn, W.H.

1958-07-01T23:59:59.000Z

215

REACTOR CONTROL  

DOE Patents (OSTI)

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

Fortescue, P.; Nicoll, D.

1962-04-24T23:59:59.000Z

216

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Christy, R.F.

1958-07-15T23:59:59.000Z

217

Catalytic reactor  

DOE Patents (OSTI)

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.

Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

2009-03-10T23:59:59.000Z

218

Decommissioning: Reactor Pressure Vessel Internals Segmentation  

Science Conference Proceedings (OSTI)

Decommissioning a nuclear plant covers a wide variety of challenging projects. One of the most challenging areas is the removal and disposal of the reactor pressure vessel (RPV) and the RPV internals. This report describes commercial reactor pressure vessel segmentation projects that have been completed and discusses several projects that are still in the planning stages. The report also covers lessons learned from each project.

2001-10-11T23:59:59.000Z

219

Mechanical Cutting of Irradiated Reactor Internal Components  

SciTech Connect

This paper discusses the use of mechanical cutting methods to volume reduce and package irradiated reactor internal components. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods used for similar projects, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. (authors)

Anderson, M.G.; Fennema, J.A. [MOTA Corporation, West Columbia, SC (United States)

2007-07-01T23:59:59.000Z

220

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Young, G.

1963-01-01T23:59:59.000Z

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


221

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

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

1958-04-15T23:59:59.000Z

222

Homogeneous fast-flux isotope-production reactor  

DOE Patents (OSTI)

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.

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

1982-08-19T23:59:59.000Z

223

A New Centrifuge CVD Reactor that will Challenge the Siemens ...  

Science Conference Proceedings (OSTI)

In this presentation, results from the new centrifuge CVD reactor along with two other ... Boron Removal from Silicon Melts by H2O/H2 Gas Blowing – Gas-phase

224

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor  

DOE Green Energy (OSTI)

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).

Ehud Greenspan

2008-09-30T23:59:59.000Z

225

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)

Wigner, E.P.

1962-12-25T23:59:59.000Z

226

Neutronic reactor  

DOE Patents (OSTI)

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

Wende, Charles W. J. (West Chester, PA)

1976-08-17T23:59:59.000Z

227

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)

Post, R.G.

1963-05-01T23:59:59.000Z

228

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Starr, C.

1963-01-01T23:59:59.000Z

229

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

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

1959-10-27T23:59:59.000Z

230

NEUTRONIC REACTORS  

DOE Patents (OSTI)

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.

Anderson, H.L.

1958-10-01T23:59:59.000Z

231

105-H Reactor Interim Safe Storage Project Final Report  

SciTech Connect

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.

E.G. Ison

2008-11-08T23:59:59.000Z

232

Continuous sulfur removal process  

DOE Patents (OSTI)

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.

Jalan, V.; Ryu, J.

1994-04-26T23:59:59.000Z

233

The NSTF at Argonne: Passive Safety and Decay Heat Removal for Advanced  

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

Engineering Engineering Experimentation > RSTA > Natural convection Shutdown heat removal Test Facility (NSTF) Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr The NSTF at Argonne: Passive Safety and Decay Heat Removal for Advanced Nuclear Reactor Designs 1 2 3 4 5 6 7 8 9 Argonne's Passive Safety Experiments to Support Modern Reactor Design Argonne National Laboratory's Natural convection Shutdown heat removal Test Facility (NSTF) is a state-of-the-art, large-scale facility for evaluating performance capabilities of decay heat removal systems. NSTF's purpose is to:

234

Geomechanical testing of MRIG-9 core for the potential SPR siting at the Richton salt dome.  

Science Conference Proceedings (OSTI)

A laboratory testing program was developed to examine the mechanical behavior of salt from the Richton salt dome. The resulting information is intended for use in design and evaluation of a proposed Strategic Petroleum Reserve storage facility in that dome. Core obtained from the drill hole MRIG-9 was obtained from the Texas Bureau of Economic Geology. Mechanical properties testing included: (1) acoustic velocity wave measurements; (2) indirect tensile strength tests; (3) unconfined compressive strength tests; (4) ambient temperature quasi-static triaxial compression tests to evaluate dilational stress states at confining pressures of 725, 1450, 2175, and 2900 psi; and (5) confined triaxial creep experiments to evaluate the time-dependent behavior of the salt at axial stress differences of 4000 psi, 3500 psi, 3000 psi, 2175 psi and 2000 psi at 55 C and 4000 psi at 35 C, all at a constant confining pressure of 4000 psi. All comments, inferences, discussions of the Richton characterization and analysis are caveated by the small number of tests. Additional core and testing from a deeper well located at the proposed site is planned. The Richton rock salt is generally inhomogeneous as expressed by the density and velocity measurements with depth. In fact, we treated the salt as two populations, one clean and relatively pure (> 98% halite), the other salt with abundant (at times) anhydrite. The density has been related to the insoluble content. The limited mechanical testing completed has allowed us to conclude that the dilatational criteria are distinct for the halite-rich and other salts, and that the dilation criteria are pressure dependent. The indirect tensile strengths and unconfined compressive strengths determined are consistently lower than other coastal domal salts. The steady-state-only creep model being developed suggests that Richton salt is intermediate in creep resistance when compared to other domal and bedded salts. The results of the study provide only limited information for structural modeling needed to evaluate the integrity and safety of the proposed cavern field. This study should be augmented with more extensive testing. This report documents a series of test methods, philosophies, and empirical relationships, etc., that are used to define and extend our understanding of the mechanical behavior of the Richton salt. This understanding could be used in conjunction with planned further studies or on its own for initial assessments.

Dunn, Dennis P.; Broome, Scott Thomas; Bronowski, David R.; Bauer, Stephen J.; Hofer, John H.

2009-02-01T23:59:59.000Z

235

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab | Department  

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

Workers Clear Reactor Shields from Brookhaven Lab Workers Clear Reactor Shields from Brookhaven Lab Recovery Act Workers Clear Reactor Shields from Brookhaven Lab 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. Recovery Act Workers Clear Reactor Shields from Brookhaven Lab More Documents & Publications Brookhaven Graphite Research Reactor Workshop 2011 ARRA Newsletters Idaho Crews Overcome Challenges to Safely Dispose 1-Million-Pound Hot Cell

236

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

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

Printer Friendly 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 Facility (ICDF). The long history of the ETR began for this water-cooled reactor with its start up in 1957, after taking only 2 years to build. According to "Proving the Principles," by Susan M. Stacy: When the Engineering Test Reactor started up at the Test Reactor Area in

237

Sandia National Laboratories Medical Isotope Reactor concept.  

SciTech Connect

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.

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

2010-04-01T23:59:59.000Z

238

REMOVAL OF EBWR FUEL ELEMENT SCALE BY SLURRY HONING  

SciTech Connect

The scale deposit on the Experimental Boiling Water Reactor fuel plates can be removed by slurry honing the plates with an abrasive-water mixture. Problems inherent in any production operation of this type are discussed. Areas of continued investigation of the method are suggested. (auth)

Charak, I.

1960-09-01T23:59:59.000Z

239

Threat of a sinkhole: A reevaluation of Cavern 4, Bayou Choctaw salt dome, Louisiana  

Science Conference Proceedings (OSTI)

Cavern Lake at Bayou Choctaw salt dome resulted from the failure of Cavern 7 in 1954. Uncontrolled solutioning of this cavern through the thin caprock had set the stage for overburden to collapse into the cavern below. A similar situation developed with nearby Cavern 4, but with less dissolutioning of the caprock. Because pressure loss was already a problem and because another 800 ft diameter lake would have endangered surface operations, solutioning of Cavern 4 was stopped and the cavern abandoned in 1957 in order to protect the already-small site. In 1978 the Strategic Petroleum Reserve (SPR) acquired a number of caverns at Bayou Choctaw, including Cavern 4, and the possible repeat of the Cavern 7 failure and formation of another lake thus became an issue. The cavern dimensions were re-sonared in 1980 for comparison with 1963 and 1977 surveys. Annual surface leveling between 1982--1992 showed less subsidence occurring than the site average, and a cavern monitoring system, installed in 1984, has revealed no anomalous motion. Repeat sonar surveys in 1992 showed very little, if any, change occurred since 1980 although a small amount of uncertainty exists as a result of changing sonar techniques. We conclude that significant additional solutioning or erosion of the caprock has not occurred and that there is no increased threat to SPR operations.

Neal, J.T.; Todd, J.L.; Linn, J.K. [Sandia National Labs., Albuquerque, NM (United States); Magorian, T.R. [Magorian (Thomas R.), Amherst, NY (United States)

1993-09-01T23:59:59.000Z

240

Questions and Answers - Why are the Halls in bio-dome shapes?  

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

How much does it cost a yearto run Jefferson Lab? How much does it cost a year<br>to run Jefferson Lab? Previous Question (How much does it cost a year to run Jefferson Lab?) Questions and Answers Main Index Next Question (Why did it take so long to build Jefferson Lab?) Why did it take so longto build Jefferson Lab? Why are the Halls in bio-dome shapes? The answer to your question is the answer to many questions... money. The shape of our experimental halls was that which could do the job and spend the least amount of money. There are several reasons for this that you won't quite understand unless you have taken geometry. If you need to enclose a certain amount of AREA, but have to pay for the LENGTH of wall you use, you want to build whatever type of building will enclose that AREA with the shortest LENGTH of wall. It just so happens a circle encloses the

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


241

FLUID MODERATED REACTOR  

DOE Patents (OSTI)

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

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

1957-10-22T23:59:59.000Z

242

REACTOR UNLOADING  

DOE Patents (OSTI)

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

Leverett, M.C.

1958-02-18T23:59:59.000Z

243

Removal of Retired Alkali Metal Test Systems  

SciTech Connect

This paper describes the successful effort to remove alkali metals, alkali metal residues, and piping and structures from retired non-radioactive test systems on the Hanford Site. These test systems were used between 1965 and 1982 to support the Fast Flux Test Facility and the Liquid Metal Fast Breeder Reactor Program. A considerable volume of sodium and sodium-potassium alloy (NaK) was successfully recycled to the commercial sector; structural material and electrical material such as wiring was also recycled. Innovative techniques were used to safely remove NaK and its residues from a test system that could not be gravity-drained. The work was done safely, with no environmental issues or significant schedule delays.

Brehm, W. F.; Church, W. R.; Biglin, J. W.

2003-02-26T23:59:59.000Z

244

Neutronic reactor  

DOE Patents (OSTI)

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

Lewis, Warren R. (Richland, WA)

1978-05-30T23:59:59.000Z

245

NUCLEAR REACTORS  

DOE Patents (OSTI)

An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

1961-12-01T23:59:59.000Z

246

REACTOR CONTROL  

DOE Patents (OSTI)

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

Ruano, W.J.

1957-12-10T23:59:59.000Z

247

Validation of the RVACS (Reactor Vessel Auxiliary Cooling System)/RACS (Reactor Air Cooling System) model in SASSYS-1  

SciTech Connect

The SASSYS-1 LMR systems analysis code contains a model for transient analysis of heat removal by a RVACS (Reactor Vessel Auxiliary Cooling System) or a RACS (Reactor Air Cooling System) in an LMR (Liquid Metal Reactor). This model has been validated by comparisons of model predictions with experimental data from a large scale RVACS/RACS simulation experiment performed at Argonne National Laboratory. 4 refs., 1 fig.

Dunn, F.E.

1987-01-01T23:59:59.000Z

248

Light Water Reactors Technology Development - Nuclear Reactors  

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

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

249

Biological Nitrogen Removal in a Gravity Flow Biomass Concentrator Reactor.  

E-Print Network (OSTI)

??Membrane technology is a well established method for wastewater treatment with advantages including: relatively easy to operate; low turbidity and BOD in the effluent; and… (more)

Scott, Daniel

2011-01-01T23:59:59.000Z

250

Removable feedwater sparger assembly  

DOE Patents (OSTI)

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.

Challberg, R.C.

1994-10-04T23:59:59.000Z

251

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

SciTech Connect

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.

A. B. Culp

2007-01-26T23:59:59.000Z

252

Ceramic membrane reactor with two reactant gases at different pressures  

DOE Patents (OSTI)

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.

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

2001-01-01T23:59:59.000Z

253

Novel Catalytic Membrane Reactors  

DOE Green Energy (OSTI)

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.

Stuart Nemser, PhD

2010-10-01T23:59:59.000Z

254

DISMANTLING OF THE REACTOR BLOCK OF THE FRJ-1 RESEARCH REACTOR (MERLIN)  

SciTech Connect

This report describes the past procedure in dismantling the reactor block of the FRJ-1 research reactor (MERLIN). Furthermore, it gives an outlook on future activities up to the final removal of the reactor block. MERLIN is an abbreviation for Medium Energy Research Light Water Moderated Industrial Nuclear Reactor. The FRJ-1 (MERLIN) was shut down in 1985 and the fuel elements removed from the facility. After dismantling the coolant loops and removing the reactor tank internals with subsequent draining of the reactor tank water, the first activities for dismantling the reactor block were carried out in summer 2001. The relevant license was granted in late July 2001 by the licensing authority specifying 8 incidental provisions. After dismantling the reactor extension (gates of the thermal columns and steel platforms surrounding the reactor block), a heavy-load platform including a casing around the reactor block was constructed. Two ventilation systems with a volume flow of 10,000 and 2 ,000 m3/h will, moreover, serve to avoid a spread of contamination. The reactor block will be dismantled in three phases divided according to upper, central and bottom sections. Dismantling the upper section started in August 2002. This section as well as the bottom section can probably be completely measured for clearance. For this reason, the activities have so far been carried out manually using mechanical and thermal techniques. The central section will probably have to be largely disposed of as radioactive waste. This is the region of the former reactor core in which the experimental devices are also integrated. Most of this work will probably have to be carried out by remote handling. More than 80 % of the dismantled materials of the reactor block can probably be measured for clearance. For this purpose, a clearance measurement device was taken into operation in the FRJ-1. On this occasion, the limits of clearance measurement have become evident. For concrete, which constitutes the largest portion of the dismantled materials by volume, an additional conditioning step has become necessary to fulfill the clearance criteria, whereas waste packages with steel components largely have to be reconditioned once more at a later stage. Material measured for clearance will be disposed of conventionally (recycling, landfill) after inspection by the official expert and clearance by the regulatory authority. Dismantled parts that cannot be measured for clearance will be transferred to the Decontamination Department of the Research Centre. From the present perspective, the dismantling of the reactor block will be completed within the first six months of 2003.

Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

2003-02-27T23:59:59.000Z

255

ELECTRONUCLEAR REACTOR  

DOE Patents (OSTI)

An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

1960-04-19T23:59:59.000Z

256

Photocatalytic reactor  

DOE Patents (OSTI)

A photocatalytic reactor 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.

Bischoff, Brian L. (Knoxville, TN); Fain, Douglas E. (Oak Ridge, TN); Stockdale, John A. D. (Knoxville, TN)

1999-01-01T23:59:59.000Z

257

Membrane reactor advantages for methanol reforming and similar reactions  

Science Conference Proceedings (OSTI)

Membrane reactors achieve efficiencies by combining in one unit a reactor that generates a product with a semipermeable membrane that extracts it. One well-known benefit of this is greater conversion, as removal of a product drives reactions toward completion, but there are several potentially larger advantages that have been largely ignored. Because a membrane reactor tends to limit the partial pressure of the extracted product, it fundamentally changes the way that total pressure in the reactor affects equilibrium conversion. Thus, many gas-phase reactions that are preferentially performed at low pressures in a conventional reactor are found to have maximum conversion at high pressures in a membrane reactor. These higher pressures and reaction conversions allow greatly enhanced product extraction as well. Further, membrane reactors provide unique opportunities for temperature management which have not been discussed previously. These benefits are illustrated for methanol reforming to hydrogen for use with PEM (polymer electrolyte membrane) fuel cells.

Buxbaum, R.E. [REB Research and Consulting Co., Ferndale, MI (United States)

1999-07-01T23:59:59.000Z

258

DECAY HEAT CONDITIONS OF CURRENT AND NEXT GENERATION REACTORS  

E-Print Network (OSTI)

Decay heat is an important parameter in reactor design. Fission products generate heat in the reactor core even when the reactor has shut down. This heat has potential to melt the core if heat removal is not sufficient, and it is what caused the accident in Japan last year. Thus, decay heat must be considered in reactor design for safety. The research focused on decay heat conditions of current and next generation reactors. US-APWR, ABWR, VHTR, and ABR were modeled and simulated using the program SCALE. When the reactors were simulated to operate for two years and cool down for one year, the ABR produced the most decay heat power during operation and cooling time, and the US-APWR, VHTR, and ABWR followed respectfully. Therefore, the ABR requires more coolant and cooling time than other reactors, and the ABWR requires the least.

Choe, JongSoo 1985-

2012-05-01T23:59:59.000Z

259

Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment |  

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

Obtains Patent for Nuclear Reactor Sodium Cleanup Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment March 28, 2013 - 12:00pm Addthis CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. Piping in the east boiler basement of the sodium processing building was color coded for easy identification. Orange indicates sodium and green identifies cooling water.

260

Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment |  

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

Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment March 28, 2013 - 12:00pm Addthis CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. Piping in the east boiler basement of the sodium processing building was color coded for easy identification. Orange indicates sodium and green identifies cooling water.

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261

CONTROL MEANS FOR REACTOR  

DOE Patents (OSTI)

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

Manley, J.H.

1961-06-27T23:59:59.000Z

262

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM  

DOE Patents (OSTI)

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

Moore, W.T.

1958-09-01T23:59:59.000Z

263

DOE Removes Brookhaven Contractor  

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

DOE Removes DOE Removes Brookhaven Contractor Peña sends a message to DOE facilities nationwide INSIDE 2 Accelerator Rx 4 FermiKids 6 Spring at Fermilab Photos courtesy of Brookhaven National Laboratory by Judy Jackson, Office of Public Affairs Secretary of Energy Federico Peña announced on Thursday, May 1, that the Department of Energy would immediately terminate the current management contract with Associated Universities, Inc. at Brookhaven National Laboratory in Upton, New York. Peña said that he made the decision after receiving the results of a laboratory safety management review conducted by the independent oversight arm of DOE's Office of Environment, Safety and Health. In addition, the Secretary said he found unacceptable "the continued on page 8 Volume 20 Friday, May 16, 1997

264

Pneumatic soil removal tool  

DOE Patents (OSTI)

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.

Neuhaus, John E. (Newport News, VA)

1992-01-01T23:59:59.000Z

265

Pneumatic soil removal tool  

Science Conference Proceedings (OSTI)

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.

Neuhaus, J.E.

1992-10-13T23:59:59.000Z

266

Removal design report for the 108-F Biological Laboratory  

Science Conference Proceedings (OSTI)

Most of the 100-F facilities were deactivated with the reactor and have since been demolished. Of the dozen or so reactor-related structures, only the 105-F Reactor Building and the 108-F Biology Laboratory remain standing today. The 108-F Biology Laboratory was intended to be used as a facility for the mixing and addition of chemicals used in the treatment of the reactor cooling water. Shortly after F Reactor began operation, it was determined that the facility was not needed for this purpose. In 1949, the building was converted for use as a biological laboratory. In 1962, the lab was expanded by adding a three-story annex to the original four-story structure. The resulting lab had a floor area of approximately 2,883 m{sup 2} (main building and annex) that operated until 1973. The building contained 47 laboratories, a number of small offices, a conference room, administrative section, lunch and locker rooms, and a heavily shielded, high-energy exposure cell. The purpose of this removal design report is to establish the methods of decontamination and decommissioning and the supporting functions associated with facility removal and disposal.

NONE

1997-09-01T23:59:59.000Z

267

Nuclear Reactor Accidents  

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

Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...

268

Principles of Reactor Physics  

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

Nuclear Reactor Physics M A Smith Argonne National Laboratory Nuclear Engineering Division Phone: 630-252-9747, Email: masmith@anl.gov Abstract: Nuclear reactor physics deals with...

269

Fission converter heat removal and safety under accident conditions  

Science Conference Proceedings (OSTI)

The design and safety of the heat removal system of the Massachusetts Institute of Technology (MIT) design for a fission converter-based epithermal beam is discussed in this paper. Plate-type reactor fuel elements, used in the MIT research reactor (MITR-II), are also used for the fission converter. This fission converter-based beam provides epithermal neutron fluxes at the patient position in excess of 10{sup 10} n/cm{sup 2}s with very low contamination of fast neutrons and gamma rays.

Sutharshan, B.; Todreas, N.E.; Harling, O.K. [Massachusetts Inst. of Technology, Cambridge, MA (United States)

1996-12-31T23:59:59.000Z

270

Summary of advanced LMR (Liquid Metal Reactor) evaluations: PRISM (Power Reactor Inherently Safe Module) and SAFR (Sodium Advanced Fast Reactor)  

Science Conference Proceedings (OSTI)

In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) (Berglund, 1987) and the Sodium Advanced Fast Reactor (SAFR) (Baumeister, 1987), were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II (NED, 1986). The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs.

Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G. (Brookhaven National Lab., Upton, NY (USA))

1989-10-01T23:59:59.000Z

271

Flexible Conversion Ratio Fast Reactor Systems Evaluation  

Science Conference Proceedings (OSTI)

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.

Neil Todreas; Pavel Hejzlar

2008-06-30T23:59:59.000Z

272

Reactor and method of operation  

DOE Patents (OSTI)

A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

Wheeler, John A. (Princeton, NJ)

1976-08-10T23:59:59.000Z

273

Metal Cutting for Large Component Removal  

Science Conference Proceedings (OSTI)

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.

Hulick, Robert M. [Bluegrass Concrete Cutting Inc., 107 Mildred Street PO Box 427, Greenville, Alabama 36037 (United States)

2008-01-15T23:59:59.000Z

274

Reactor safety method  

DOE Patents (OSTI)

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.

Vachon, Lawrence J. (Clairton, PA)

1980-03-11T23:59:59.000Z

275

NEUTRONIC REACTOR MANIPULATING DEVICE  

DOE Patents (OSTI)

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

Ohlinger, L.A.

1962-08-01T23:59:59.000Z

276

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

DOE Green Energy (OSTI)

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.

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

1987-09-01T23:59:59.000Z

277

Waste tank 241-SY-101 dome airspace and ventilation system response to a flammable gas plume burn  

SciTech Connect

A series of flammable gas plume burn and transient pressure analyses have been completed for a nuclear waste tank (241-SY-101) and associated tank farm ventilation system at the U.S. Department of Energy`s Hanford facility. The subject analyses were performed to address issues concerning the effects of transient pressures resulting from igniting a small volume of concentrated flammable gas just released from the surface of the waste as a plume and before the flammable gas concentration could be reduced by mixing with the dome airspace by local convection and turbulent diffusion. Such a condition may exist as part of an in progress episode gas release (EGR) or gas plume event. The analysis goal was to determine the volume of flammable gas that if burned within the dome airspace would result in a differential pressure, after propagating through the ventilation system, greater than the current High Efficiency Particulate Filter (HEPA) limit of 2.49 KPa (10 inches of water or 0. 36 psi). Such a pressure wave could rupture the tank ventilation system inlet and outlet HEPA filters leading to a potential release of contaminants to the environment

Heard, F.J.

1995-11-01T23:59:59.000Z

278

Engine Removal Projection Tool  

DOE Green Energy (OSTI)

The US Navy has over 3500 gas turbine engines used throughout the surface fleet for propulsion and the generation of electrical power. Past data is used to forecast the number of engine removals for the next ten years and determine engine down times between removals. Currently this is done via a FORTRAN program created in the early 1970s. This paper presents results of R&D associated with creating a new algorithm and software program. We tested over 60 techniques on data spanning 20 years from over 3100 engines and 120 ships. Investigated techniques for the forecast basis including moving averages, empirical negative binomial, generalized linear models, Cox regression, and Kaplan Meier survival curves, most of which are documented in engineering, medical and scientific research literature. We applied those techniques to the data, and chose the best algorithm based on its performance on real-world data. The software uses the best algorithm in combination with user-friendly interfaces and intuitively understandable displays. The user can select a specific engine type, forecast time period, and op-tempo. Graphical displays and numerical tables present forecasts and uncertainty intervals. The technology developed for the project is applicable to other logistic forecasting challenges.

Ferryman, Thomas A.; Matzke, Brett D.; Wilson, John E.; Sharp, Julia L.; Greitzer, Frank L.

2005-06-02T23:59:59.000Z

279

Fuel provision for nonbreeding deuterium-tritium fusion reactors  

SciTech Connect

Nonbreeding D-T reactors have decisive advantages in minimum size, unit cost, variety of applications, and ease of heat removal over reactors using any other fusion cycle, and significant advantages in environmental and safety characteristics over breeding D-T reactors. Considerations of relative energy production demonstrate that the most favorable source of tritium for a widely deployed system of nonbreeding D-T reactors is the very large (approx. 10 GW thermal) semi-catalyzed-deuterium (SCD), or sub-SCD reactor, where none of the escaping /sup 3/He (> 95%) or tritium (< 25%) is reinjected for burn-up. Feasibility of the ignited SCD tokamak reactor requires spatially averaged betas of 15 to 20% with a magnetic field at the TF coils of 12 to 13 Tesla.

Jassby, D.L.; Katsurai, M.

1980-01-01T23:59:59.000Z

280

PRISM: An innovative liquid metal fast breeder reactor  

SciTech Connect

This paper describes an innovative sodium-cooled reactor concept employing small certified reactor modules coupled with a standardized steam generator system. The total plant employs nine PRISM reactors (power reactor inherently safe module) in three 415 MWe power blocks. The PRISM design concept utilizes inherent safety characteristics and modularity to improve licensability, reduce owner's risk, and reduce costs. The relatively small size of each reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. It is proposed that a single PRISM module be used in a full-scale integrated reactor safety test. Results from the test would be used to obtain NRC certification of the standard design.

Kruger, G.B.; Boardman, C.E.; Olich, E.E.; Switick, D.M.

1986-01-01T23:59:59.000Z

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


281

Method and apparatus for enhancing reactor air-cooling system performance  

DOE Patents (OSTI)

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.

Hunsbedt, A.

1996-03-12T23:59:59.000Z

282

Method and apparatus for enhancing reactor air-cooling system performance  

DOE Patents (OSTI)

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.

Hunsbedt, Anstein (Los Gatos, CA)

1996-01-01T23:59:59.000Z

283

An Engineering Test Reactor  

SciTech Connect

A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.

Fahrner, T.; Stoker, R.L.; Thomson, A.S.

1951-03-16T23:59:59.000Z

284

MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)  

SciTech Connect

The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

GERBER MS

2009-04-28T23:59:59.000Z

285

Program on Technology Innovation: Assessment of Tritium Removal Technologies  

Science Conference Proceedings (OSTI)

Tritium is produced during nuclear power plant operation, with the majority of production stemming from a reaction with boron in pressurized water reactors (PWRs) in which neutrons from the plant’s chain reaction are absorbed. This report reviews separation technology development for the removal of tritium from wastewaters to meet environmental release limits.BackgroundThe designation “tritiated water” refers to water in which ...

2013-06-26T23:59:59.000Z

286

Reactor Pressure Vessel Task of Light Water Reactor Sustainability...  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure...

287

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

DOE Patents (OSTI)

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.

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

1997-01-01T23:59:59.000Z

288

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

DOE Patents (OSTI)

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.

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

1997-12-16T23:59:59.000Z

289

Geothermal hydrogen sulfide removal  

DOE Green Energy (OSTI)

UOP Sulfox technology successfully removed 500 ppM hydrogen sulfide from simulated mixed phase geothermal waters. The Sulfox process involves air oxidation of hydrogen sulfide using a fixed catalyst bed. The catalyst activity remained stable throughout the life of the program. The product stream composition was selected by controlling pH; low pH favored elemental sulfur, while high pH favored water soluble sulfate and thiosulfate. Operation with liquid water present assured full catalytic activity. Dissolved salts reduced catalyst activity somewhat. Application of Sulfox technology to geothermal waters resulted in a straightforward process. There were no requirements for auxiliary processes such as a chemical plant. Application of the process to various types of geothermal waters is discussed and plans for a field test pilot plant and a schedule for commercialization are outlined.

Urban, P.

1981-04-01T23:59:59.000Z

290

Pneumatic soil removal tool  

Science Conference Proceedings (OSTI)

A soil tool is provided for removing radioactive soil, rock and 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 on the pipe. The actuator rod of the pneumatic cylinder is connected to a collar which is slidably 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 value mounted on the handle, to provide movement of the movable jaw.

Neuhaus, J.F.

1991-01-23T23:59:59.000Z

291

Process for removing technetium from iron and other metals  

DOE Patents (OSTI)

Technetium is a radioactive product of the nuclear fission process. During reprocessing of spent or partially spent fuel from nuclear reactors, the technetium can be released and contaminate other, otherwise good, metals. A specific example is equipment in gaseous diffusion uranium enrichment cascades which have been used to process fuel which was returned from reactors, so-called reactor returns. These returns contained volatile technetium compounds which contaminated the metals in the equipment. Present regulations require that technetium be removed before the metal can be re-used at non-radioactive sites. Removing the technetium from contaminated metals has two desirable results. First, the large amount of nonradioactive metal produced by the process herein described can be recycled at a much lower cost than virgin metal can be produced. Second, large amounts of radioactively contaminated metal can be reduced to relatively small amounts of radioactive slag and large amounts of essentially uncontaminated metal. A new and improved process for removing technetium from iron and other metals is described in which between 1/10 atom % and 5 atom % of manganese is added to the contaminated metal in order to replace the technetium.

Leitnaker, James M.; Trowbridge, Lee D.

1997-12-01T23:59:59.000Z

292

IMPROVED PROCESSES TO REMOVE NAPHTHENIC ACIDS  

SciTech Connect

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.

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

2004-04-28T23:59:59.000Z

293

Memorandum on Chemical Reactors and Reactor Hazards  

SciTech Connect

Two important problems in the investigation of reactor hazards are the chemical reactivity of various materials employed in reactor construction and the chracteristics of heat transfer under transient conditions, specifically heat transfer when driven by an exponentially increasing heat source (exp t/T). Although these problems are independent of each other, when studied in relation to reactor hazards they may occur in a closely coupled sequence. For example the onset of a dangerous chemical reactor may be due to structural failure of various reactor components under an exponentially rising heat source originating with a runaway nuclear reactor. For this reason, these two problems should eventually be studied together after an exploratory experimental survey has been made in which they are considered separately.

Mills, M.M.; Pearlman, H.; Ruebsamen, W.; Steele, G., Chrisney, J.

1951-07-05T23:59:59.000Z

294

Coal hydrogenation and deashing in ebullated bed catalytic reactor  

SciTech Connect

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.

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

1983-01-01T23:59:59.000Z

295

Design options for a bunsen reactor.  

SciTech Connect

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.

Moore, Robert Charles

2013-10-01T23:59:59.000Z

296

Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations  

SciTech Connect

Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated cost of decommissioning a PWR is lowest for ENTOMB and highest for SAFSTOR • the estimated cost of decommissioning a BWR is lowest for OECON and highest for SAFSTOR. In all cases, SAFSTOR has the lowest occupational radiation dose and the highest cost.

Wittenbrock, N. G.

1982-01-01T23:59:59.000Z

297

Progress in understanding of direct containment heating phenomena in pressurized light water reactors  

DOE Green Energy (OSTI)

Progress is described in development of a mechanistic understanding of direct containment heating phemonena arising during high-pressure melt ejection accidents in pressurized water reactor systems. The experimental data base is discussed which forms the basis for current assessments of containment pressure response using current lumped-parameter containment analysis methods. The deficiencies in available methods and supporting data base required to describe major phenomena occurring in the reactor cavity, intermediate subcompartments and containment dome are highlighted. Code calculation results presented in the literature are cited which demonstrate that the progress in understanding of DCH phenomena has also resulted in current predictions of containment pressure loadings which are significantly lower than are predicted by idealized, thermodynamic equilibrium calculations. Current methods are, nonetheless, still predicting containment-threatening loadings for large participating melt masses under high-pressure ejection conditions. Recommendations for future research are discussed. 36 refs., 5 figs., 1 tab.

Ginsberg, T.; Tutu, N.K.

1988-01-01T23:59:59.000Z

298

Process for treating effluent from a supercritical water oxidation reactor  

DOE Patents (OSTI)

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.

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

1997-01-01T23:59:59.000Z

299

Process for treating effluent from a supercritical water oxidation reactor  

DOE Patents (OSTI)

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.

Barnes, C.M.; Shapiro, C.

1997-11-25T23:59:59.000Z

300

Basic and Applied Science Research Reactors - Reactors designed...  

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

BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th...

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


301

Non-oxidative conversion of methane with continuous hydorgen removal  

SciTech Connect

The objective is to overcome the restrictions of non-oxidative methane pyrolysis and oxidative coupling of methane by transferring hydrogen across a selective inorganic membrane between methane and air streams, without simultaneous transport of hydrocarbon reactants or products. This will make the overall reaction system exothermic, remove the thermodynamic barrier to high conversion, and eliminate the formation of carbon oxides. Our approach is to couple C-H bond activation and hydrogen removal by passage of hydrogen atoms through a dense ceramic membrane. In our membrane reactor, catalytic methane pyrolysis produces C2+ hydrogen carbons and aromatics on the one side of the membrane and hydrogen is removed through an oxide film and combusted with air on the opposite side. This process leads to a net reaction with the stoichiometry and thermodynamic properties of oxidative coupling, but without contact between the carbon atoms and oxygen species.

Borry, R.W. III [California Univ., Berkeley, CA (United States). Dept. of Chemical Engineering; Iglesia, E. [California Univ., Berkeley, CA (United States). Lawrence Berkeley Lab.

1997-12-31T23:59:59.000Z

302

Geologic technical assessment of the Stratton Ridge salt dome, Texas, for potential expansion of the U.S. strategic petroleum reserve.  

SciTech Connect

The Stratton Ridge salt dome is a large salt diapir located only some ten miles from the currently active Strategic Petroleum Reserve Site at Bryan Mound, Texas. The dome is approximately 15 miles south-southwest of Houston. The Stratton Ridge salt dome has been intensively developed, in the desirable central portions, with caverns for both brine production and product storage. This geologic technical assessment indicates that the Stratton Ridge salt dome may be considered a viable, if less-than-desirable, candidate site for potential expansion of the Strategic Petroleum Reserve (SPR). Past development of underground caverns significantly limits the potential options for use by the SPR. The current conceptual design layout of proposed caverns for such an expansion facility is based upon a decades-old model of salt geometry, and it is unacceptable, according to this reinterpretation of salt dome geology. The easternmost set of conceptual caverns are located within a 300-ft buffer zone of a very major boundary shear zone, fault, or other structural feature of indeterminate origin. This structure transects the salt stock and subdivides it into an shallow western part and a deeper eastern part. In places, the distance from this structural boundary to the design-basis caverns is as little as 150 ft. A 300-ft distance from this boundary is likely to be the minimum acceptable stand-off, from both a geologic and a regulatory perspective. Repositioning of the proposed cavern field is possible, as sufficient currently undeveloped salt acreage appears to be available. However, such reconfiguration would be subject to limitations related to land-parcel boundaries and other existing infrastructure and topographic constraints. More broadly speaking, the past history of cavern operations at the Stratton Ridge salt dome indicates that operation of potential SPR expansion caverns at this site may be difficult, and correspondingly expensive. Although detailed information is difficult to come by, widely accepted industry rumors are that numerous existing caverns have experienced major operational problems, including salt falls, sheared casings, and unintended releases of stored product(s). Many of these difficulties may be related to on-going differential movement of individual salt spines or to lateral movement at the caprock-salt interface. The history of operational problems, only some of which appear to be a matter of public record, combined with the potential for encountering escaped product from other operations, renders the Stratton Ridge salt dome a less-than-desirable site for SPR purposes.

Rautman, Christopher Arthur; Snider, Anna C.; Looff, Karl M. (Geologic Consultant, Lovelady, TX)

2006-11-01T23:59:59.000Z

303

Removal to Maximum Extent Practical  

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

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

304

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

DOE Patents (OSTI)

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.

Hunsbedt, A.; Boardman, C.E.

1995-04-11T23:59:59.000Z

305

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

SciTech Connect

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.

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

1995-01-01T23:59:59.000Z

306

Attrition reactor system  

DOE Patents (OSTI)

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.

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

1993-01-01T23:59:59.000Z

307

Guidebook to nuclear reactors  

SciTech Connect

A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen.

Nero, A.V. Jr.

1976-05-01T23:59:59.000Z

308

Reactor Sharing Program  

Science Conference Proceedings (OSTI)

Progress achieved at the University of Florida Training Reactor (UFTR) facility through the US Department of Energy's University Reactor Sharing Program is reported for the period of 1991--1992.

Vernetson, W.G.

1993-01-01T23:59:59.000Z

309

Attrition reactor system  

DOE Patents (OSTI)

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.

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

1993-09-28T23:59:59.000Z

310

Medium Power Lead Alloy Reactors: Missions for this Reactor Technology  

Science Conference Proceedings (OSTI)

A multiyear project at the Idaho National Engineering and Environmental Laboratory and the Massachusetts Institute of Technology investigated the potential of medium-power lead-alloy-cooled technology to perform two missions: (1) the production of low-cost electricity and (2) the burning of actinides from light water reactor (LWR) spent fuel. The goal of achieving a high power level to enhance economic performance simultaneously with adoption of passive decay heat removal and modularity capabilities resulted in designs in the range of 600-800 MW(thermal), which we classify as a medium power level compared to the lower [~100 MW(thermal)] and higher [2800 MW(thermal)] power ratings of other lead-alloy-cooled designs. The plant design that was developed shows promise of achieving all the Generation-IV goals for future nuclear energy systems: sustainable energy generation, low overnight capital cost, a very low likelihood and degree of core damage during any conceivable accident, and a proliferation-resistant fuel cycle. The reactor and fuel cycle designs that evolved to achieve these missions and goals resulted from study of the following key trade-offs: waste reduction versus reactor safety, waste reduction versus cost, and cost versus proliferation resistance. Secondary trade-offs that were also considered were monolithic versus modular design, active versus passive safety systems, forced versus natural circulation, alternative power conversion cycles, and lead versus lead-bismuth coolant. These studies led to a selection of a common modular design with forced convection cooling, passive decay heat removal, and a supercritical CO2 power cycle for all our reactor concepts. However, the concepts adopt different core designs to optimize the achievement of the two missions. For the low-cost electricity production mission, a design approach based on fueling with low enriched uranium operating without costly reprocessing in a once-through cycle was pursued to achieve a long operating cycle length by enhancing in-core breeding. For the actinide-burning mission three design variants were produced: (1) a fertile-free actinide burner, i.e., a single-tier strategy, (2) a minor actinide burner with plutonium burned in the LWR fleet, i.e., a two-tier strategy, and (3) an actinide burner with characteristics balanced to also favor economic electricity production.

Neil E. Todreas; Philip E. MacDonald; Pavel Hejzlar; Jacopo Buongiorno; Eric Loewen

2004-09-01T23:59:59.000Z

311

Fuel rod retention device for a nuclear reactor  

DOE Patents (OSTI)

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

Hylton, Charles L. (Madison Heights, VA)

1984-01-01T23:59:59.000Z

312

High solids fermentation reactor  

DOE Patents (OSTI)

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.

Wyman, Charles E. (Lakewood, CO); Grohmann, Karel (Littleton, CO); Himmel, Michael E. (Littleton, CO); Richard, Christopher J. (Lakewood, CO)

1993-01-01T23:59:59.000Z

313

Improved vortex reactor system  

DOE Patents (OSTI)

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.

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

1995-01-01T23:59:59.000Z

314

FAST NEUTRON REACTOR  

DOE Patents (OSTI)

A reactor comprising fissionable material in concentration sufficiently high so that the average neutron enengy within the reactor is at least 25,000 ev is described. A natural uranium blanket surrounds the reactor, and a moderating reflector surrounds the blanket. The blanket is thick enough to substantially eliminate flow of neutrons from the reflector.

Soodak, H.; Wigner, E.P.

1961-07-25T23:59:59.000Z

315

The Allegheny Power Service Constructed Wetland at Springdale: The Role of Plants in the Removal of Trace Elements  

Science Conference Proceedings (OSTI)

Constructed wetlands are proving an effective technology for the removal of many aqueous contaminants. The ability of wetlands to remove contaminants such as trace elements appears to be a function of both the physical trapping of suspended materials and the biological and chemical processes occurring within the wetlands. Thus, wetlands are commonly described as "biogeochemical reactors." This report details a study of trace element removal and sequestration within a highly engineered wetland designed to...

2001-11-05T23:59:59.000Z

316

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

SciTech Connect

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.

Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma (Computational Engineering Analysis, Albuquerque, NM); Al Rashdan, Ahmad (Texas A& M University, College Station, TX); Tsvetkov, Pavel Valeryevich (Texas A& M University, College Station, TX); Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

2011-05-01T23:59:59.000Z

317

Shielding considerations for advanced space nuclear reactor systems  

SciTech Connect

To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

Angelo, J.P. Jr.; Buden, D.

1982-01-01T23:59:59.000Z

318

Shielded fluid stream injector for particle bed reactor  

DOE Patents (OSTI)

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 inline 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.

Notestein, J.E.

1991-12-31T23:59:59.000Z

319

Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico  

SciTech Connect

The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program.

Salazar, M.; Elder, J.

1992-08-01T23:59:59.000Z

320

Design Considerations for Economically Competitive Sodium Cooled Fast Reactors  

SciTech Connect

The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, 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.

Hongbin Zhang; Haihua Zhao

2009-05-01T23:59:59.000Z

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


321

The Argonaut Reactor - Reactors designed/built by Argonne National  

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

Achievements > Achievements > Argonne Reactors > Training Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

322

Nuclear reactor overflow line  

DOE Patents (OSTI)

The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

Severson, Wayne J. (Pittsburgh, PA)

1976-01-01T23:59:59.000Z

323

Reactor vessel support system  

DOE Patents (OSTI)

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.

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

1982-01-01T23:59:59.000Z

324

Spinning fluids reactor  

SciTech Connect

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.

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

2012-11-20T23:59:59.000Z

325

Fission reactors and materials  

SciTech Connect

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions.

Frost, B.R.T.

1981-12-01T23:59:59.000Z

326

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

Cao, Jun

2011-01-01T23:59:59.000Z

327

Reactor water cleanup system  

DOE Patents (OSTI)

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.

Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

1994-01-01T23:59:59.000Z

328

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

Jun Cao

2011-01-12T23:59:59.000Z

329

Sodium Reactor Experiment decommissioning. Final report  

Science Conference Proceedings (OSTI)

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.

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

1983-08-15T23:59:59.000Z

330

Modeling and performance of the MHTGR (Modular High-Temperature Gas-Cooled Reactor) reactor cavity cooling system  

SciTech Connect

The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab.

Conklin, J.C. (Oak Ridge National Lab., TN (USA))

1990-04-01T23:59:59.000Z

331

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

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

332

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

SciTech Connect

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)

Reyna, D. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bernstein, A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Lund, J.; Kiff, S.; Cabrera-Palmer, B. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bowden, N. S.; Dazeley, S.; Keefer, G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

2011-07-01T23:59:59.000Z

333

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

Dickson, J.J.

1963-09-24T23:59:59.000Z

334

Gas-cooled nuclear reactor  

DOE Patents (OSTI)

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.

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

1985-01-01T23:59:59.000Z

335

Analysis of decay heat removal capabilities at Susquehanna  

Science Conference Proceedings (OSTI)

Since the Susquehanna station began operating in 1982, the Nuclear Safety Assessment Group (NSAG) has conducted independent outage safety reviews. In 1990, NSAG published The Outage Planning Information document which consolidated some of the knowledge gained during the outage safety review process. The report discusses the requirements in the technical specifications and it analyzes the decay heat removal problem during the various stages of an outage. The appendices contain facts concerning decay heat generation and the capabilities of the decay heat removal systems. To the maximum extent possible, design documents and certified test data have been used to establish the capabilities of the systems. The Susquehanna station has two General Electric (GE) boiling water reactors (BWR)-4 GE BWR-4 reactors with Mark-2 containments. The safety systems are composed of two independent divisions. The possible heat sinks are the cooling towers via the circulating water and service water systems, the spray pond via the residual heat removal systems, and the atmosphere via boiling.

Miltenberger, J.R. (Pennsylvania Power and Light Co., Allentown (United States)); Henry, R.A. (Pennsylvania Power and Light Co., Berwick (United States))

1992-01-01T23:59:59.000Z

336

Decontamination and decommissioning experience at Experimental Breeder Reactor No. 1  

SciTech Connect

A description is presented of the experience obtained from decontamination and decommissioning of the Experimental Breeder Reactor No. 1 located at the Idaho Nuclear Engineering Laboratory, Scottsville, Idaho. Included are the planning, preparation, and operations associated with the removal and processing of radioactively contaminated sodium-potassium eutectic alloy (NaK) and decontamination and decommissioning of the facility. (auth)

Kendall, E.W.

1975-09-01T23:59:59.000Z

337

REACTOR DEVELOPMENT QUARTERLY PROGRESS REPORT  

DOE Green Energy (OSTI)

The design and construction program of the Bolling Experimental Reactor is reviewed. A number of preliminary experinnents were performed with Borax-II at pressures between 75 and 300 psi. The most corrosion-resistant U-Zr--Nb alloy developed so far is produced by heating in a vacuum to the gamma phase, quenching, and aging for 2 hr at 400 deg C. Special attention is given to the removal of H/sub 2/ from the material. Unclad and unirradiated samples of U--Nb and U--Nb--Zr alloys were corrosion tested in H/sub 2/O. Corrosion rates were also measured under irradiation conditions in CP-5. Elongation measurements of irradiated wrought and cast U--Zr material suggested no way for treating the wrought fuel so that stability comparable to the cast material could be obtained. Natural circulation boiling density tests at 600 psia were made in order to determine the effects of channel cross section and subcooling on the steam void fraction. Results of autoclave and dynamic corrosion studies of 2-S Al in H/sub 2/O are reported. These results include the testing of Ni-clad samples. A large number of criticality calculations were performed for the EBR-Il and the PBR. The solubility of ThO/sub 2/ pellets containing various concentrations of U/sub 3/ O/sub 8/ was tested in water at 316 deg C for periods of 672 to 744 hr. None of the samples disintegrated, although at least one sample developed cracks. Solutions of reactor kinetic equations were attempted for the purpose of studying transients in reactors with lifetimes of 7 x 10/sup -9/ 10/sup -7/, and 6 x 10/ sup -5/ sec. Ignition experiments were performed on Th, Cu, Al, Fe, Mg, Zr, and fluorothene when contacted with fluorides. Except for Zr and fluorothene, the materials did not ignite. (C.H.)

None

1955-01-31T23:59:59.000Z

338

Article removal device for glovebox  

DOE Patents (OSTI)

An article removal device for a glovebox is described comprising a conduit extending through a glovebox wall which may be closed by a plug within the glovebox, and a fire-resistant container closing the outer end of the conduit and housing a removable container for receiving pyrophoric or otherwise hazardous material without disturbing the interior environment of the glovebox or adversely affecting the environment outside of the glovebox. (Official Gazette)

Guyer, R.H.; Leebl, R.G.

1973-12-01T23:59:59.000Z

339

Regenerable Sorbent Development for Sulfur, Chloride and Ammonia Removal from Coal-Derived Synthesis Gas  

DOE Green Energy (OSTI)

A large number of components in coal form corrosive and toxic compounds during coal gasification processes. DOE’s NETL aims to reduce contaminants to parts per billion in order to utilize gasification gas streams in fuel cell applications. Even more stringent requirements are expected if the fuel is to be utilized in chemical production applications. Regenerable hydrogen sulfide removal sorbents have been developed at NETL. These sorbents can remove the hydrogen sulfide to ppb range at 316 °C and at 20 atmospheres. The sorbent can be regenerated with oxygen. Reactivity and physical durability of the sorbent did not change during the multi-cycle tests. The sorbent development work has been extended to include the removal of other major impurities, such as HCl and NH3. The sorbents for HCl removal that are available today are not regenerable. Regenerable HCl removal sorbents have been developed at NETL. These sorbents can remove HCl to ppb range at 300 °C to 500 °C. The sorbent can be regenerated with oxygen. Results of TGA and bench-scale flow reactor tests with both regenerable and non-regenerable HCl removal sorbents will be discussed in the paper. Bench-scale reactor tests were also conducted with NH3 removal sorbents. The results indicated that the sorbents have a high removal capacity and good regenerability during the multi-cycle tests. Future emphasis of the NETL coal gasification/cleanup program is to develop multi-functional sorbents to remove multiple impurities in order to minimize the steps involved in the cleanup systems. To accomplish this goal, a regenerable sorbent capable of removing both HCl and H2S was developed. The results of the TGA conducted with the sorbent to evaluate the feasibility of both H2S and HCl sorption will be discussed in this paper.

Siriwardane, R.V.; Tian, H.; Simonyi, T.; Webster, T.

2007-08-01T23:59:59.000Z

340

ADVANCES IN HEXAVALENT CHROMIUM REMOVAL AT HANFORD  

SciTech Connect

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.

NESHEM DO; RIDDELLE J

2012-01-30T23:59:59.000Z

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


341

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

DOE Patents (OSTI)

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.

Walsh, John V. (Glendora, CA)

1987-12-15T23:59:59.000Z

342

Design for Ceramic Membrane Reactor with two Reactant Gases at Different Pressures  

DOE Patents (OSTI)

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.

Balachandran, Uthamalingam; Mieville, Rodney L.

1998-11-18T23:59:59.000Z

343

Generation -IV Reactor Concepts  

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

Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),

344

NUCLEAR REACTORS AND EARTHQUAKES  

SciTech Connect

A book is presented which supplies pertinent seismological information to engineers in the nuclear reactor field. Data are presented on the occurrence, intensity, and wave shapes. Techniques are described for evaluating the response of structures to such events. Certain reactor types and their modes of operation are described briefly. Various protection systems are considered. Earthquake experience in industrial and reactor plants is described. (D.L.C.)

1961-01-01T23:59:59.000Z

345

THERMAL NEUTRONIC REACTOR  

DOE Patents (OSTI)

A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

Spinrad, B.I.

1960-01-12T23:59:59.000Z

346

Improved vortex reactor system  

DOE Patents (OSTI)

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.

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

1995-05-09T23:59:59.000Z

347

Recycle of LWR (Light Water Reactor) actinides to an IFR (Integral Fast Reactor)  

SciTech Connect

A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs.

Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

1991-01-01T23:59:59.000Z

348

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

DOE Patents (OSTI)

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.

Yunker, Wayne H. (Richland, WA); Christiansen, David W. (Kennewick, WA)

1987-01-01T23:59:59.000Z

349

Advanced Nuclear Research Reactor  

SciTech Connect

This report describes technical modifications implemented by INVAP to improve the safety of the Research Reactors the company designs and builds.

Lolich, J.V.

2004-10-06T23:59:59.000Z

350

Pressurized fluidized bed reactor  

DOE Patents (OSTI)

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.

Isaksson, Juhani (Karhula, FI)

1996-01-01T23:59:59.000Z

351

Pressurized fluidized bed reactor  

DOE Patents (OSTI)

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.

Isaksson, J.

1996-03-19T23:59:59.000Z

352

Tokamak reactor first wall  

DOE Patents (OSTI)

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.

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

1984-11-20T23:59:59.000Z

353

HOMOGENEOUS NUCLEAR POWER REACTOR  

DOE Patents (OSTI)

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

King, L.D.P.

1959-09-01T23:59:59.000Z

354

Public comment sought on hot cell removal project at the Idaho Site�s  

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

Public comment sought on hot cell removal project at the Idaho Site�s Advanced Test Reactor Complex Public comment sought on hot cell removal project at the Idaho Site�s Advanced Test Reactor Complex The U.S. Department of Energy (DOE) is seeking public comment on a project to remove three unused hot cells and the 1950s era laboratory building that contains them at the Idaho Site�s Advanced Test Reactor complex. An Engineering Evaluation/Cost Analysis (EE/CA) document with three proposed alternatives for the final end state of the building and hot cells is under evaluation by DOE, the U.S. Environmental Protection Agency, and Idaho�s Department of Environmental Quality. The TRA-632 building and the hot cells were built in 1952 for assembly, disassembly and examination of nuclear test reactor components. The 13,000 sq. foot building contains three shielded hot cells with lathes, power saws, grinders, and other remote handling equipment. In addition to the examination of test reactor components, the hot cells have been used during the production of radioisotopes for medical use like cobalt-60 and iridium-192. The last active work in the hot cells took place in 2004, and the aging facility was placed on standby in 2006.

355

DEVELOPMENT OF A 250 kW AQUEOUS HOMOGENEOUS SINGLE REGION SUSPENSION REACTOR  

DOE Green Energy (OSTI)

The reactor (20% enriched U/sup 235/) is pressurized with hydrogen up to 20 atm overpressure and has a maximum working temperature of about 250 deg C. The process flowsheet, including the systems for fusion product removal, is presented. Xenon is removed directly by gas stripping and indirectly by removal of iodine. The neutron flux is measured but not used for control purposes. (W.D.M.)

Kreyger, P.J.; van der Plas, Th.; van der Schee, B.L.A.; Went, J.J.; van Zolingen, J.J.

1958-06-11T23:59:59.000Z

356

LLNL-Generated Content for the California Academy of Sciences, Morrison Planetarium Full-Dome Show: Earthquake  

SciTech Connect

The California Academy of Sciences (CAS) Morrison Planetarium is producing a 'full-dome' planetarium show on earthquakes and asked LLNL to produce content for the show. Specifically the show features numerical ground motion simulations of the M 7.9 1906 San Francisco and a possible future M 7.05 Hayward fault scenario earthquake. The show also features concepts of plate tectonics and mantle convection using images from LLNL's G3D global seismic tomography. This document describes the data that was provided to the CAS in support of production of the 'Earthquake' show. The CAS is located in Golden Gate Park, San Francisco and hosts over 1.6 million visitors. The Morrison Planetarium, within the CAS, is the largest all digital planetarium in the world. It features a 75-foot diameter spherical section projection screen tilted at a 30-degree angle. Six projectors cover the entire field of view and give a three-dimensional immersive experience. CAS shows strive to use scientifically accurate digital data in their productions. The show, entitled simply 'Earthquake', will debut on 26 May 2012. They are working on graphics and animations based on the same data sets for display on LLNL powerwalls and flat-screens as well as for public release.

Rodgers, A J; Petersson, N A; Morency, C E; Simmons, N A; Sjogreen, B

2012-01-23T23:59:59.000Z

357

LLNL-Generated Content for the California Academy of Sciences, Morrison Planetarium Full-Dome Show: Earthquake  

SciTech Connect

The California Academy of Sciences (CAS) Morrison Planetarium is producing a 'full-dome' planetarium show on earthquakes and asked LLNL to produce content for the show. Specifically the show features numerical ground motion simulations of the M 7.9 1906 San Francisco and a possible future M 7.05 Hayward fault scenario earthquake. The show also features concepts of plate tectonics and mantle convection using images from LLNL's G3D global seismic tomography. This document describes the data that was provided to the CAS in support of production of the 'Earthquake' show. The CAS is located in Golden Gate Park, San Francisco and hosts over 1.6 million visitors. The Morrison Planetarium, within the CAS, is the largest all digital planetarium in the world. It features a 75-foot diameter spherical section projection screen tilted at a 30-degree angle. Six projectors cover the entire field of view and give a three-dimensional immersive experience. CAS shows strive to use scientifically accurate digital data in their productions. The show, entitled simply 'Earthquake', will debut on 26 May 2012. They are working on graphics and animations based on the same data sets for display on LLNL powerwalls and flat-screens as well as for public release.

Rodgers, A J; Petersson, N A; Morency, C E; Simmons, N A; Sjogreen, B

2012-01-23T23:59:59.000Z

358

Prismatic modular reactor analysis with melcor  

E-Print Network (OSTI)

Hydrogen, a more sustainable source of energy, is a potential substitute for hydrocarbon fuel for power generation. The Very High Temperature gas-cooled Reactor (VHTR) concept can produce hydrogen with high efficiency and in large quantities. The US Department of Energy plans to build a VHTR as a next-generation hydrogen/electricity production plant. This reactor concept is very different from that of commercial reactors in the US. In order to acquire licensing eligibility for VHTRs, analysis tools need to be validated and applied to design and evaluate VHTRs under operation conditions and accident scenarios. In this thesis, MELCOR, a severe accident code, was used to analyze one of the VHTR designs – a prismatic core Next Generation Nuclear Plant (NGNP). The NGNP is based on General Atomics‘ (GA) Gas Turbine – Modular Helium Reactor (GT-MHR) 600 MW design. According to the current literature survey, more data is available for the GT-MHR than for the NGNP. Therefore, for the purposes of extending MELCOR capabilities and code validation, a model of the GT-MHR reactor pressure vessel (RPV) was developed. Based on the currently available data, a model of the NGNP RPV was then developed through modifying the GT-MHR RPV model. For both RPV models, coolant outlet temperature under normal operating conditions corresponds well to the data from literature. The reactor cavity cooling systems (RCCS), which passively removes heat from the RPV wall to the outside atmosphere, was then added to this GT-MHR RPV model. With this model addition, the heat removal rate of the RCCS under normal operating conditions was calculated to correspond well to the data from references. Pressurized conduction cooldown (PCC), one of the important postulated accident scenarios for a prismatic core reactor, was simulated with the complete model. MELCOR has been demonstrated to have the ability of modeling a prismatic core VHTR. The calculated outlet temperature and mass flow rate under normal operation correspond well to references. However, the calculation for the heat distribution in the graphite and fuel is unsatisfactory which requires MELCOR modification for the PCC simulation. For future work, a complete model of the NGNP under normal operation conditions will be developed when additional data becomes available.

Zhen, Ni

2008-12-01T23:59:59.000Z

359

Microsoft Word - EC Sodium coolant removal.doc  

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

1 1 SECTION A. Project Title: MFC - EBR-II Sodium Removal/RCRA Closure Activities SECTION B . Project Description The proposed action will remove the sodium from the Experimental Breeder Reactor (EBR)-II piping system and tanks to achieve clean-closure for eventual decommissioning, deactivation and demolition (DD&D). The clean-closure will be completed in compliance with the EBR-II Hazardous Waste Management Act/Resource Conservation and Recovery Act (HWMA/RCRA) Storage and Treatment Permit PER-120, which includes the closure plan. EBR-II is located at the Materials and Fuels Complex at the Idaho National Laboratory. The EBR-II DD&D actions will be addressed under the Comprehensive Environmental Response Compensation, and Liability Act, specifically, the Engineering Evaluation/Cost

360

METHOD FOR REMOVING SODIUM OXIDE FROM LIQUID SODIUM  

DOE Patents (OSTI)

A method is described for removing sodium oxide from a fluent stream of liquid sodium by coldtrapping the sodium oxide. Apparatus utilizing this method is disclosed in United States Patent No. 2,745,552. Sodium will remain in a molten state at temperatures below that at which sodium oxide will crystallize out and form solid deposits, therefore, the contaminated stream of sodium is cooled to a temperature at which the solubility of sodium oxide in sodium is substantially decreased. Thereafter the stream of sodium is passed through a bed of stainless steel wool maintained at a temperature below that of the stream. The stream is kept in contact with the wool until the sodium oxide is removed by crystal growth on the wool, then the stream is reheated and returned to the system. This method is useful in purifying reactor coolants where the sodium oxide would otherwise deposit out on the walls and eventually plug the coolant tubes.

Bruggeman, W.H.; Voorhees, B.G.

1957-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor dome removal" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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361

High removal rate laser-based coating removal system  

DOE Patents (OSTI)

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.

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)

1999-11-16T23:59:59.000Z

362

Removal - An alternative to clearance  

SciTech Connect

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)

Feinhals, J.; Kelch, A. [TUV NORD SysTec GmbH and Co. KG, Hamburg (Germany); Kunze, V. [Federal Office for Radiation Protection, Salzgitter (Germany)

2007-07-01T23:59:59.000Z

363

Arsenic removal and stabilization by synthesized pyrite  

E-Print Network (OSTI)

Arsenic is ubiquitous whether it is naturally occurring or produced by humans. It is found at sites on the National Priority List and at sites operated by DOE, where it is the second most commonly found contaminant. More wastes containing arsenic will be produced due to the lowering of the Maximum Contaminant Level (MCL) for arsenic in drinking water which will result in more treatment facilities for arsenic removal that will generate residuals. Furthermore, arsenic can be released from such wastes under the reduced conditions that are found in landfills. Pyrite (FeS2) is believed to be a compound that has a high affinity for arsenic and is stable under anoxic conditions. The first task of this research was to develop a method for making pyrite crystals of defined size with minimal reaction time and at high yield. Effects on the synthesis of pyrite particles of pH, the ratio of Fe/S, temperature and reaction time were investigated in batch reactor systems. Pyrite was synthesized within 24 hours at pH values ranging from pH 3.6 through pH 5.6, and at a ratio of Fe/S of 0.5. X-ray diffraction and scanning electron microscopy were used to size and characterize the pyrite particles. Experimental and analytical procedures developed for this work, included a hydride generation atomic absorption spectrometry method for measuring arsenic species (As(III), As(V)). The synthesized pyrite was applied to remove arsenic and its maximum capacity for arsenic removal was measured in batch adsorption experiments to be 3.23 ?mol/g for As(III) and 113 ?mol/g for As(V). Information obtained on the characteristics of chemical species before and after the reaction with arsenic showed that iron and sulfur were oxidized. Last, how strongly arsenic was bound to pyrite was investigated and it was determined that release of arsenic from As(III)-pyrite is not affected by pH, but release from As(V)-pyrite is affected by pH with minimum release in the range pH 5 to pH 8.

Song, Jin Kun

2008-12-01T23:59:59.000Z

364

Multipollutant Removal with WOWClean® System  

E-Print Network (OSTI)

WOW Energy built and tested its multipollutant removal WOWClean® system in a development program with the Texas Commission on Environmental Quality (TCEQ). A 2,500 ACFM mobile unit was built to field test the removal of air pollution constituents from the flue gas of a power plant and demonstrate the technology. The system integrates proven emission reduction techniques into a single, multi-pollutant reduction system and is designed to remove Mercury, SOx, NOx, particulates, heavy metals and other pollutants from low temperature flue gases. Its advantages include robust operation, lower investment, lower operating cost and high removal rates. The WOWClean® system has been tested on flue gases resulting from the combustion of diverse fuels such as petcoke, coal, wood, diesel and natural gas. In addition to significant removal of CO2, test results demonstrate the capability to reduce 99.5% SOx (from levels as high as 2200 ppm), 90% reduction of NOx, and > 90% heavy metals. The paper will include details of the multi-pollutant system along with test results.

Romero, M.

2010-01-01T23:59:59.000Z

365

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

National Nuclear Security Administration (NNSA)

Research ReactorDomestic Research Reactor Receipt Coordinator, Savannah River Nuclear Solutions | National Nuclear Security Administration Our Mission Managing the Stockpile...

366

ADMINISTRATION OF ORNL RESEARCH REACTORS  

SciTech Connect

Organization of the ORNL Operations division for administration of the Oak Ridge Research Reactor, the Low Intensity Testing Reactor, and the Oak Ridge Graphite Reactor is described. (J.R.D.)

Casto, W.R.

1962-08-20T23:59:59.000Z

367

Vietnam HEU Removal | National Nuclear Security Administration  

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

Removal NNSANews posted a photo: Vietnam HEU Removal A truck carrying the last highly enriched uranium in Vietnam winds through the Vietnamese countryside. Facebook Twitter Youtube...

368

Production reactor characteristics  

SciTech Connect

Reactors for the production of special nuclear materials share many similarities with commercial nuclear power plants. Each relies on nuclear fission, uses uranium fuel, and produces large quantities of thermal power. However, there are some important differences in production reactor characteristics that may best be discussed in terms of mission, role, and technology.

Thiessen, C.W.; Hootman, H.E.

1990-01-01T23:59:59.000Z

369

Advanced converter reactors  

SciTech Connect

Advanced converter reactors (ACRs) of primary US interest are those which can be commercialized within about 20 years, and are: Advanced Light-Water Reactors, Spectral-Shift-Control Reactors, Heavy-Water Reactors (CANDU type), and High-Temperature Gas-Cooled Reactors. These reactors can operate on uranium, thorium, or uranium-thorium fuel cycles, but have the greatest fuel utilization on thorium type cycles. The water reactors tend to operate more economically on uranium cycles, while the HTGR is more economical on thorium cycles. Thus, the HTGR had the greatest practical potential for improving fuel utilization. If the US has 3.4 to 4 million tons U/sub 3/O/sub 8/ at reasonable costs, ACRs can make important contributions to maintaining a high nuclear power level for many decades; further, they work well with fast breeder reactors in the long term under symbiotic fueling conditions. Primary nuclear data needs of ACRs are integral measurements of reactivity coefficients and resonance absorption integrals.

Kasten, P.R.

1979-01-01T23:59:59.000Z

370

NEUTRONIC REACTOR SYSTEM  

DOE Patents (OSTI)

A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

Treshow, M.

1959-02-10T23:59:59.000Z

371

NEUTRONIC REACTOR BURIAL ASSEMBLY  

DOE Patents (OSTI)

A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

Treshow, M.

1961-05-01T23:59:59.000Z

372

The Integral Fast Reactor  

SciTech Connect

Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

Till, C.E.; Chang, Y.I. (Argonne National Lab., IL (USA)); Lineberry, M.J. (Argonne National Lab., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

373

Process for Removing Radioactive Wastes from Liquid Streams  

SciTech Connect

The process is under development at Mound Laboratory to remove radioactive waste (principally plutonium-238) from process water prior to discharge of the water to the Miami river. The contaminated water, as normally received, is at a pH between 6 and 90. Under these conditions, plutonium in all its oxidation states is hydrolyzed; however, the level of the radioactive solids varies from about 50ppm down to about 50 ppb and the plutonium remains in a colloidal or subcolloidal condition. The permissible concentration for discharge to the river is about 50 parts per trillion. Pilot plant test show that 95-99% of the radioactive material is removed by adsorption on diatomaceous earth. The remainder is removed by passage through a bed of either dibasic or tribasic calcium phosphate. Ground phosphate rock is equally effective in removing the radioactive material if the flow rate is controlled to permit sufficient contact time. Parameters for optimizing the process are now under study. Future plans include application of the process to wastes from reactor fuels reprocessing.

Kirby, H. W.; Blane, D. E.; Smolin, R. I.

1972-10-01T23:59:59.000Z

374

Plasma generators, reactor systems and related methods  

SciTech Connect

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.

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

2007-06-19T23:59:59.000Z

375

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab  

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

UPTON, N.Y. - American Recovery and Reinvestment Act 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 end Office of Environmental Management legacy cleanup activities at the Lab. The neutron shields were located on the north and south sides of a 700-ton graphite pile. The three-inch-thick shields absorbed neutrons that escaped from the graphite pile. The shields also limited movement of the pile when the reactor was in opera-

376

The Neutrino Mass Hierarchy from Nuclear Reactor Experiments  

E-Print Network (OSTI)

10 years from now reactor neutrino experiments will attempt to determine which neutrino mass eigenstate is the most massive. In this letter we present the results of more than seven million detailed simulations of such experiments, studying the dependence of the probability of successfully determining the mass hierarchy upon the analysis method, the neutrino mass matrix parameters, reactor flux models, geoneutrinos and, in particular, combinations of baselines. We show that a recently reported spurious dependence of the data analysis upon the high energy tail of the reactor spectrum can be removed by using a weighted Fourier transform. We determine the optimal baselines and corresponding detector locations. For most values of the CP-violating, leptonic Dirac phase delta, a degeneracy prevents NOvA and T2K from determining either delta or the hierarchy. We determine the confidence with which a reactor experiment can determine the hierarchy, breaking the degeneracy.

Emilio Ciuffoli; Jarah Evslin; Xinmin Zhang

2013-02-04T23:59:59.000Z

377

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

Science Conference Proceedings (OSTI)

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.

Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

2012-06-01T23:59:59.000Z

378

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT  

SciTech Connect

Progress on reactor programs and in general engineering research and development programs is summarized. Research and development are reported on water-cooled reactors including EBWR and Borax-V, sodium-cooled reactors including ZPR-III, IV, and IX, Juggernaut, and EBR-I and II. Other work included a review of fast reactor technology, and studies on nuclear superheat, thermal and fast reactor safety, and reactor physics. Effort was also devoted to reactor materials and fuels development, heat engineering, separation processes and advanced reactor concepts. (J.R.D.)

1961-04-01T23:59:59.000Z

379

Thermal-hydraulic aspects of flow inversion in a research reactor  

SciTech Connect

PARET, a neutronics and thermal-hydraulics computer code, has been modified to account for natural convection in a reactor core. The code was then used to analyze the flow inversion that occurs in a reactor with heat removal by forced convection in the downward direction after a pump failure. Typical results are shown for a number of parameters. Research reactors normally operating much above ten MW are predicted to experience nucleate boiling in the event of a flow inversion. Comparison with experimental results from the Belgian BR2 reactor indicated general agreement although nucleate boiling that was analytically predicted was not noted in the BR2 data.

Smith, R.S.; Woodruff, W.L.

1986-11-01T23:59:59.000Z

380

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Davidson, J.K.

1963-11-19T23:59:59.000Z

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


381

Nuclear reactor fuel rod attachment system  

DOE Patents (OSTI)

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.

Christiansen, David W. (Kennewick, WA)

1982-01-01T23:59:59.000Z

382

Nuclear reactor control column  

DOE Patents (OSTI)

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.

Bachovchin, Dennis M. (Plum Borough, PA)

1982-01-01T23:59:59.000Z

383

Forecast Technical Document Felling and Removals  

E-Print Network (OSTI)

Forecast Technical Document Felling and Removals Forecasts A document describing how volume fellings and removals are handled in the 2011 Production Forecast system. Tom Jenkins Robert Matthews Ewan Mackie Lesley Halsall #12;PF2011 ­ Felling and removals forecasts Background A fellings and removals

384

2011sr01.doc  

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

Tuesday, February 1, 2011 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

385

Improved method for removing metal vapor from gas streams  

DOE Patents (OSTI)

This invention relates to a process for gas cleanup to remove one or more metallic contaminants present as vapor. More particularly, the invention relates to a gas cleanup process using mass transfer to control the saturation levels such that essentially no particulates are formed, and the vapor condenses on the gas passage surfaces. It addresses the need to cleanup an inert gas contaminated with cadmium which may escape from the electrochemical processing of Integral Fast Reactor (IFR) fuel in a hot cell. The IFR is a complete, self-contained, sodium-cooled, pool-type fast reactor fueled with a metallic alloy of uranium, plutonium and zirconium, and is equipped with a close-coupled fuel cycle. Tests with a model have shown that removal of cadmium from argon gas is in the order of 99.99%. The invention could also apply to the industrial cleanup of air or other gases contaminated with zinc, lead, or mercury. In addition, the invention has application in the cleanup of other gas systems contaminated with metal vapors which may be toxic or unhealthy.

Ahluwalia, R.K.; Im, K.H.

1994-09-19T23:59:59.000Z

386

NUCLEAR REACTOR FUEL SYSTEMS  

DOE Patents (OSTI)

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.

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

1959-09-15T23:59:59.000Z

387

Spherical torus fusion reactor  

DOE Patents (OSTI)

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.

Martin Peng, Y.K.M.

1985-10-03T23:59:59.000Z

388

CONTROL FOR NEUTRONIC REACTOR  

DOE Patents (OSTI)

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.

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

1959-03-31T23:59:59.000Z

389

Fast Breeder Reactor studies  

Science Conference Proceedings (OSTI)

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.

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

1980-07-01T23:59:59.000Z

390

Microfluidic electrochemical reactors  

DOE Patents (OSTI)

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.

Nuzzo, Ralph G. (Champaign, IL); Mitrovski, Svetlana M. (Urbana, IL)

2011-03-22T23:59:59.000Z

391

High removal rate laser-based coating removal system  

Science Conference Proceedings (OSTI)

A compact laser system is disclosed that removes surface coatings (such as paint, dirt, etc.) at a removal rate as high as 1,000 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.

Matthews, D.L.; Celliers, P.M.; Hackel, L.; Da Silva, L.B.; Dane, C.B.; Mrowka, S.

1999-11-16T23:59:59.000Z

392

Savannah River Site | Department of Energy  

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

Savannah River Site Savannah River Site Savannah River Site Work is under way to decommission the Heavy Water Components Test Reactor, which had been used to test experimental fuel assemblies for commercial heavy-water power reactors. SRS is scheduled to remove the dome of the reactor this month (January 2011). Workers also will displace the reactor vessel and steam generators, grout the remaining structure in place, and install a concrete cover over the reactor's footprint Work is under way to decommission the Heavy Water Components Test Reactor, which had been used to test experimental fuel assemblies for commercial heavy-water power reactors. SRS is scheduled to remove the dome of the reactor this month (January 2011). Workers also will displace the reactor vessel and steam generators, grout the remaining structure in place, and

393

Investigating the Use of Biosorbents to Remove Arsenic from Water  

E-Print Network (OSTI)

Evaluating the ability of biosorbents to remove arsenic from water has global significance due to the widespread availability and low cost of biosorbent materials. In this study, the ability of coffee grounds and coconut substrate (two previously unreported biosorbents) to remove arsenic from water was compared against the performance of arsenic removal on rice husk (a recognized and widely tested biosorbent). The three biosorbents were individually screened for their ability to remove arsenite, As (III), and arsenate, As (V), from water. Batch reactors were employed to assess the percent removal, reaction kinetics, adsorption capacity, and desorption of each arsenic species onto/from biosorbents under pH buffered and non?buffered conditions. The resulting experimental data was statistically interpreted using analysis of variance and ttesting of the means. The experimental results were also fit to existing kinetic and isotherm models to provide kinetic rate constants, the maximum adsorption capacity, and to help interpret the nature of the reactions on the biosorbent surface. While all three biosorbents removed arsenic with similar initial reaction kinetics (pseudo 1st order reaction rate constant for As (III) was 0.13 hr^?1 for all three biosorbents and for As (V) was 0.17 hr^?1 for coffee grounds and rice husk and 0.15 hr^?1 for coconut substrate), the amount of arsenite and arsenate removed was highest for coffee grounds (84 and 91 percent, respectively), followed by rice husk (68 and 72 percent, respectively), and then coconut substrate (26 and 24 percent, respectively). The maximum adsorption capacity of arsenite and arsenate was determined for coffee grounds (0.66 and 0.70 mg/g, respectively) and rice husk (0.55 and 0.66 mg/g, respectively). While desorption was observed for both coffee grounds and rice husk, the total amount of desorption accounted for less than 15 percent of the total retained mass. The results of this thesis work reveal that coffee can be used as an effective biosorbent when compared to rice husk; however, coconut substrate is less effective than rice husk at removing As (III) and As (V).

Erapalli, Shreyas

2010-12-01T23:59:59.000Z

394

Pebble Bed Reactor Dust Production Model  

SciTech Connect

The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors’ PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production.

Abderrafi M. Ougouag; Joshua J. Cogliati

2008-09-01T23:59:59.000Z

395

Reactor hot spot analysis  

SciTech Connect

The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

Vilim, R.B.

1985-08-01T23:59:59.000Z

396

Fission product behavior in the Molten Salt Reactor Experiment  

SciTech Connect

Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with $sup 235$U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using $sup 233$U fuel over a period of about 15 months (more than 5100 effective full- power hours). (auth)

Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

1975-10-01T23:59:59.000Z

397

Molten metal reactors  

SciTech Connect

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.

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

2013-11-05T23:59:59.000Z

398

Compact power reactor  

DOE Patents (OSTI)

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.

Wetch, Joseph R. (Woodland Hills, CA); Dieckamp, Herman M. (Canoga Park, CA); Wilson, Lewis A. (Canoga Park, CA)

1978-01-01T23:59:59.000Z

399

Storage of spent fuel from the nation`s nuclear reactors: Status, technology, and policy options  

SciTech Connect

Since the beginning of the commercial nuclear electric power industry, it has been recognized that spent nuclear reactor fuel must be able to be readily removed from the reactor vessel in the plant and safely stored on-site. The need for adjacent ready storage is first for safety. In the event of an emergency, or necessary maintenance that requires the removal of irradiated fuel from the reactor vessel, cooled reserve storage capacity for the full amount of fuel from the reactor core must be available. Also, the uranium fuel in the reactor eventually reaches the point where its heat generation is below the planned efficiency for steam production which drives the turbines and generators. It then must be replaced by fresh uranium fuel, with the ``spent fuel`` elements being removed to a safe and convenient storage location near the reactor vessel. The federal nuclear waste repository program, even without delays in the current schedule of disposal becoming available in 2003, will result in a large percentage of the 111 existing operable commercial reactors requiring expansion of their spent fuel storage capacity. How that need can and will be met raises issues of both technology and policy that will be reviewed in this report.

1989-10-01T23:59:59.000Z

400

Chloride removal from plutonium alloy  

Science Conference Proceedings (OSTI)

SRP is evaluating a program to recover plutonium from a metallic alloy that will contain chloride salt impurities. Removal of chloride to sufficiently low levels to prevent damaging corrosion to canyon equipment is feasible as a head-end step following dissolution. Silver nitrate and mercurous nitrate were each successfully used in laboratory tests to remove chloride from simulated alloy dissolver solution containing plutonium. Levels less than 10 ppM chloride were achieved in the supernates over the precipitated and centrifuged insoluble salts. Also, less than 0.05% loss of plutonium in the +3, +4, or +6 oxidation states was incurred via precipitate carrying. These results provide impetus for further study and development of a plant-scale process to recover plutonium from metal alloy at SRP.

Holcomb, H.P.

1983-01-01T23:59:59.000Z

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


401

Removing Barriers to Interdisciplinary Research  

E-Print Network (OSTI)

A significant amount of high-impact contemporary scientific research occurs where biology, computer science, engineering and chemistry converge. Although programmes have been put in place to support such work, the complex dynamics of interdisciplinarity are still poorly understood. In this paper we interrogate the nature of interdisciplinary research and how we might measure its "success", identify potential barriers to its implementation, and suggest possible mechanisms for removing these impediments.

Naomi Jacobs; Martyn Amos

2010-12-19T23:59:59.000Z

402

METHOD OF REMOVING STRONTIUM IONS  

DOE Patents (OSTI)

A method is given for removing trace amounts of Sr/sup 90/ from solutions. Phosphate ion is added to the solution and it is then brought into contact with a solid salt such as calcium carbonate which will react methathetically with the phosphate ion to form a salt such as calcium phosphate. During this reaction, strontium will be absorbed to a high degree within the newly formed lattice. (AEC)

Rhodes, D.W.; McHenry, J.R.; Ames, L.L. Jr.

1962-05-01T23:59:59.000Z

403

Hanford Railcars Make Final Stop at B Reactor: Move Enhances Visitor  

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

Hanford Railcars Make Final Stop at B Reactor: Move Enhances Hanford Railcars Make Final Stop at B Reactor: Move Enhances Visitor Experience at Historic Reactor Hanford Railcars Make Final Stop at B Reactor: Move Enhances Visitor Experience at Historic Reactor May 10, 2011 - 12:00pm Addthis Media Contacts Cameron Hardy, DOE (509) 376-5365 Cameron.Hardy@rl.doe.gov Andre Armstrong, CH2M HILL (509) 376-6773 Andre_l_Armstrong@rl.gov RICHLAND, WASH. - Two locomotives that hauled irradiated fuel around the Hanford Site for a half-century will reach their final stop this week when they are delivered to the Historic B Reactor for preservation and public display. The two locomotives are among 16 railcars from Hanford's 200 North Area being removed by Department of Energy (DOE) contractor CH2M HILL Plateau Remediation Company (CH2M HILL).

404

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

DOE Patents (OSTI)

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.

Grossman, Mark W. (Belmont, MA)

1991-01-01T23:59:59.000Z

405

NEUTRONIC REACTOR CONSTRUCTION AND OPERATION  

DOE Patents (OSTI)

A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

West, J.M.; Weills, J.T.

1960-03-15T23:59:59.000Z

406

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

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

to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. 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 remove up to 200 metric tons (MT) of Highly Enriched Uranium (HEU), in the coming decades, from further use as fissile material in U.S. nuclear weapons and prepare this material for other uses. Secretary Bodman made this announcement while addressing the 2005 Carnegie International Nonproliferation Conference in Washington, DC.

407

HIGH FLUX ISOTOPE REACTOR PRELIMINARY DESIGN STUDY  

SciTech Connect

A comparison of possible types of research reactors for the production of transplutonium elements and other isotopes indicates that a flux-trap reactor consisting of a beryllium-reflecteds light-water-cooled annular fuel region surrounding a light-water island provides the required thermal neutron fluxes at minimum cost. The preliminary desigu of such a reactor was carried out on the basis of a parametric study of the effect of dimensions of the island and fuel regions heat removal rates, and fuel loading on the achievable thermal neutmn fluxes in the island and reflector. The results indicate that a 12- to 14-cm- diam. island provides the maximum flux for a given power density. This is in good agreement with the US8R critical experiments. Heat removal calculations indicate that average power densities up to 3.9 Mw/liter are achievable with H/ sub 2/O-cooled, platetype fuel elements if the system is pressurized to 650 psi to prevent surface boiling. On this basis, 100 Mw of heat can be removed from a 14-cm-ID x 36-cm-OD x 30.5-cm-long fuel regions resulting in a thermal neutron flux of 3 x 10/sup 15/ in the island after insertion of 100 g of Cm/sup 244/ or equivalent. The resulting production of Cf/sup 252/ amounts to 65 mg for a 1 1/2- year irradiation. Operation of the reactor at the more conservative level of 67 Mw, providing an irradiation flux of 2 x 10/sup 15/ in the islands will result in the production of 35 mg of Cf/sup 252/ per 18 months from 100 g of Cm/sup 244/. A development program is proposed to answer the question of the feasibility of the higher power operation. In addition to the central irradiation facility for heavyelement productions the HFIR contains ten hydraulic rabbit tubes passing through the beryllium reflector for isotope production and four beam holes for basic research, Preliminary estimates indicate that the cost of the facility, designed for an operating power level of 100 Mw, will be approximately 2 million. (auth)

Lane, J.A.; Cheverton, R.D.; Claiborne, G.C.; Cole, T.E.; Gambill, W.R.; Gill, J.P.; Hilvety, N.; McWherther, J.R.; Vroom, D.W.

1959-03-20T23:59:59.000Z

408

Safety aspects of Particle Bed Reactor plutonium burner system  

SciTech Connect

An assessment is made of the safety aspects peculiar to using the Particle Bed Reactor (PBR) as the burner in a plutonium disposal system. It is found that a combination of the graphitic fuel, high power density possible with the PBR and engineered design features results in an attractive concept. The high power density potentially makes it possible to complete the plutonium burning without requiring reprocessing and remanufacturing fuel. This possibility removes two hazardous steps from a plutonium burning complex. Finally, two backup cooling systems depending on thermo-electric converters and heat pipes act as ultimate heat removal sinks in the event of accident scenarios which result in loss of fuel cooling.

Powell, J.R.; Ludewig, H.; Todosow, M.

1993-08-01T23:59:59.000Z

409

Liquid metal reactor deactivation as applied to the experimental breeder reactor - II.  

DOE Green Energy (OSTI)

The Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-W) was shutdown in September, 1994. This sodium cooled reactor had been in service since 1964, and by the US Department of Energy (DOE) mandate, 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 (SPF) 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 SPF.

Earle, O. K.; Michelbacher, J. A.; Pfannenstiel, D. F.; Wells, P. B.

1999-05-28T23:59:59.000Z

410

REACTOR GROUT THERMAL PROPERTIES  

DOE Green Energy (OSTI)

Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

2011-01-28T23:59:59.000Z

411

Applications for reactor-pumped lasers  

Science Conference Proceedings (OSTI)

Nuclear reactor-pumped lasers (RPLs) have been developed in the US by the Department of Energy for over two decades, with the primary research occurring at Sandia National Laboratories and Idaho National Engineering Laboratory. The US program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1,271, 1,733, 1,792, 2,032, 2,630, 2,650, and 3,370 nm with intrinsic efficiency as high as 2.5%. The major strengths of a reactor-pumped laser are continuous high-power operation, modular construction, self-contained power, compact size, and a variety of wavelengths (from visible to infrared). These characteristics suggest numerous applications not easily accessible to other laser types. The continuous high power of an RPL opens many potential manufacturing applications such as deep-penetration welding and cutting of thick structures, wide-area hardening of metal surfaces by heat treatment or cladding application, wide-area vapor deposition of ceramics onto metal surfaces, production of sub-micron sized particles for manufacturing of ceramics, and 3-D ceramic lithography. In addition, a ground-based RPL could beam its power to space for such activities as illuminating geosynchronous communication satellites in the earth`s shadow to extend their lives, beaming power to orbital transfer vehicles, removing space debris, and providing power (from earth) to a lunar base during the long lunar night.

Lipinski, R.J.; McArthur, D.A. [Sandia National Labs., Albuquerque, NM (United States). Nuclear Systems Research

1994-10-01T23:59:59.000Z

412

B Reactor | Department of Energy  

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

Operational Management » History » Manhattan Project » Signature Operational Management » History » Manhattan Project » Signature Facilities » B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first 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 initial design output of 1,000 kilowatts, the B Reactor was designed to operate at 250,000 kilowatts. Consisting of a 28- by 36-foot, 1,200-ton graphite cylinder lying on its side, the reactor was penetrated through its

413

Design and Manufacture of the Storage Cask for the Old Reactor Internals  

SciTech Connect

Mitsubishi Heavy Industries, Ltd. (MHI) completed replacement work of upper reactor internals (UCI) and lower reactor internals (LCI) of the pressurized water reactor in Shikoku Electric Power Company's Ikata Unit No.1 by 'the all-in-one-piece extraction method' introduced in the document of [ICONE14-89233]. In the pressurized water reactor (PWR) plant, the UCI are usually removed from the reactor vessel (RV) independently and reinstalled into the RV again every refueling outage. The LCI are independently able to be removed from the RV and reinstalled again during in-service inspection, too. In the boiling water reactor (BWR) plant, there are several cases of replacing BWR shrouds by cutting small and containing in a container. But no replacement of all reactor internals (CI) for the PWR, in one piece without splitting or cutting, has been reported. The purpose of this paper is to introduce the key points about the design and manufacture of the storage cask for old reactor internals in the replacement work by 'the all-in-one-piece extraction method'. (author)

Yasuhiro Tomiita [Mitsubishi Heavy Industries, Ltd. (Japan)

2006-07-01T23:59:59.000Z

414

C^ DECOMMISSIONING OF EIGHT SURPLUS PRODUCTION REACTORS AT THE HANFORD SITE,  

E-Print Network (OSTI)

To request copies of the draft environmental impact statement, contact Ms. WheeLess at the above address. e.`? ABSTRACT: The purpose of this Draft Envirormental Impact Statement ( DEIS) is to provide environmental information to assist the U.S. Department of Energy ( DOE) in the selection of a decomaissioning alternative for the eight surplus production reactors at the Hanford Site, Richland, Washington. Five alternatives are considered in this DEIS: 1) No Action, in which the reactors are left in place and the present maintenance and surveillance programs are continued; 2) Immediate One-Piece Removal, in which the reactor buildings are demolished and the reactor blocks are transported in one piece on a tractor-transporter across the Site along a predetermined route to an onsite low-Level ^* ( waste burial area; 3) Safe Storage Followed by Deferred One-Piece Removal, in which the reactors are temporarily stored in a safe, secure status for 75 years, after which the reactor buildings are demolished and the reactor blocks are transported in one piece on a tractor-transporter across the Site along a predetermined route to an onsite low-LeveL waste-buriat area; 4) Safe Storage Followed by Deferred Dismantlement, in which the reactors are temporarily stored in a safe, secure status for

Karen J. Wheeless

1989-01-01T23:59:59.000Z

415

SMAHTR - A Concept for a Small, Modular Advanced High Temperaure Reactor  

SciTech Connect

Several new high temperature reactor concepts, referred to as Fluoride Salt Cooled High Temperature Reactors (FHRs), have been developed over the past decade. These FHRs use a liquid salt coolant combined with high temperature gas-cooled reactor fuels (TRISO) and graphite structural materials to provide a reactor that operates at very high temperatures and is scalable to large sizes perhaps exceeding 2400 MWt. This paper presents a new small FHR the Small Modular Advanced High Temperature Reactor or SmAHTR . SmAHTR is targeted at applications that require compact, high temperature heat sources either for high efficiency electricity production or process heat applications. A preliminary SmAHTR concept has been developed that delivers 125 MWt of energy in an integral primary system design that places all primary and decay heat removal heat exchangers inside the reactor vessel. The current reactor baseline concept utilizes a prismatic fuel block core, but multiple removable fuel assembly concepts are under evaluation as well. The reactor vessel size is such that it can be transported on a standard tractor-trailer to support simplified deployment. This paper will provide a summary of the current SmAHTR system concept and on-going technology and system architecture trades studies.

Gehin, Jess C [ORNL; Greene, Sherrell R [ORNL; Holcomb, David Eugene [ORNL; Carbajo, Juan J [ORNL; Cisneros, Anselmo T [ORNL; Corwin, William R [ORNL; Ilas, Dan [ORNL; Wilson, Dane F [ORNL; Varma, Venugopal Koikal [ORNL; Bradley, Eric Craig [ORNL; Yoder, III, Graydon L [ORNL

2010-01-01T23:59:59.000Z

416

Heat exchanger for reactor core and the like  

DOE Patents (OSTI)

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.

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

1986-01-01T23:59:59.000Z

417

Advanced Reactor Development and Technology - Nuclear Engineering...  

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

Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Nuclear Data Program Advanced Reactor Development Overview Advanced Fast Reactor...

418

Zero Power Reactor simulation | Argonne National Laboratory  

NLE Websites -- All DOE Office Websites (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...

419

THE MATERIALS OF FAST BREEDER REACTORS  

E-Print Network (OSTI)

jet aircraft engines, and nuclear reactor fuel elements. Ancomponents of a nuclear reactor core are susceptible tothe nuclear physics of the thermal and fast neutron reactors

Olander, Donald R.

2013-01-01T23:59:59.000Z

420

Gas-cooled reactors  

SciTech Connect

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

Schulten, R.; Trauger, D.B.

1976-01-01T23:59:59.000Z

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421

Natural gas: Removing the obstacles  

Science Conference Proceedings (OSTI)

Defining and then figuring out how to remove whatever obstacles may be blocking the wider use of natural gas was the purpose of a conference held early last month in Phoenix, Arizona. The unique, three-day event was jointly sponsored by the U.S. Department of Energy (DOE) and the National Association of Regulatory Utility Commissioners (NARUC). It drew an overflow crowd of more than 500, with a registration list that read like a Who's Who of the natural gas industry. This article summarizes some of the main points of this conference.

Romo, C.

1992-03-15T23:59:59.000Z

422

Material Removal and Disposition | National Nuclear Security...  

National Nuclear Security Administration (NNSA)

Removal and Disposition | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...

423

Nuclear reactor safety device  

DOE Patents (OSTI)

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.

Hutter, Ernest (Wilmette, IL)

1986-01-01T23:59:59.000Z

424

REACTOR CONTROL DEVICE  

DOE Patents (OSTI)

A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

Graham, R.H.

1962-09-01T23:59:59.000Z

425

A NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

Luebke, E.A.; Vandenberg, L.B.

1959-09-01T23:59:59.000Z

426

Reactor for exothermic reactions  

DOE Patents (OSTI)

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.

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

1993-03-02T23:59:59.000Z

427

MERCHANT MARINE SHIP REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01T23:59:59.000Z

428

Heat dissipating nuclear reactor  

DOE Patents (OSTI)

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.

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

1987-01-01T23:59:59.000Z

429

Heat dissipating nuclear reactor  

DOE Patents (OSTI)

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.

Hunsbedt, A.; Lazarus, J.D.

1985-11-21T23:59:59.000Z

430

MERCHANT MARINE SHIP REACTOR  

DOE Patents (OSTI)

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

Sankovich, M.F.; Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Gestson, D.K.

1961-05-01T23:59:59.000Z

431

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from

432

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the operation of commercial nuclear power plants, require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including

433

Passive cooling system for top entry liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

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

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

1992-01-01T23:59:59.000Z

434

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

DOE Patents (OSTI)

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.

Sturm, W.J.

1958-12-01T23:59:59.000Z

435

Reactor and Material Supply | Y-12 National Security Complex  

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

Reactor and Reactor and Material Supply Reactor and Material Supply Y-12 has processed highly enriched uranium for more than 60 years in support of the nation's defense. The end of the Cold War and ensuing strategic arms control treaties have resulted in an excess of HEU materials. In 1994, approximately 174 metric tons of weapons-usable HEU was declared surplus to defense needs. The HEU disposition program was established to make the surplus HEU unsuitable for use in weapons by blending it down to low-enriched uranium and to recover the economic value of the materials to the extent practical. In 2005, the Secretary of Energy announced that an additional 200 metric tons of HEU would be removed from further use as fissile material in U.S. nuclear weapons. Approximately 20 metric tons of this material will

436

TURRET: A HIGH TEMPERATURE GAS-CYCLE REACTOR PROPOSAL  

SciTech Connect

A nitrogen-cooled graphite-moderated nuclear reactor experiment is proposed to drive a closed-cycle gas turbine power plant at 1300 deg F. The annular core of the reactor can be rotated inside the reflector to permit fuel loading and discharge while operating at full power. Small cylindrical fuel elements of graphite are solutionimpregnated with partially enriched uranium. The fuel is recycled by incineration of the elements, chemical fresh graphite tn a small batch process. The unclad, uncoated fuel should permit high burn-up and simple fuel processing, but allows fission product diffusion into the gas stream. While methods are proposed for the removal of these from the gas, the Song-term consequences on turbine operation are unknown. The compatibility of nitrogen gas with the fuel has been studied experimentally. The radial movement of fuel gives a reactor with a constant power profile and no excess reactivity. The temperature is regulated by the fuel charging rate. (auth)

Hammond, R.P.; Busey, H.M.; Chapman, K.R.; Durham, F.P.; Rogers, J.D.; Wykoff, W.R.

1958-01-23T23:59:59.000Z

437

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

DOE Patents (OSTI)

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.

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

1996-12-31T23:59:59.000Z

438

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

SciTech Connect

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.

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

1998-01-01T23:59:59.000Z

439

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

DOE Patents (OSTI)

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.

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

1998-01-13T23:59:59.000Z

440

Nuclear reactor apparatus  

DOE Patents (OSTI)

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.

Wade, Elman E. (Ruffs Dale, PA)

1978-01-01T23:59:59.000Z

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


441

Fusion reactor pumped laser  

DOE Patents (OSTI)

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.

Jassby, Daniel L. (Princeton, NJ)

1988-01-01T23:59:59.000Z

442

Fast quench reactor method  

DOE Patents (OSTI)

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.

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)

1999-01-01T23:59:59.000Z

443

Perspectives on reactor safety  

SciTech Connect

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.

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

1994-03-01T23:59:59.000Z

444

Reactor Neutrino Experiments  

E-Print Network (OSTI)

Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measurement are also briefed.

Jun Cao

2007-12-06T23:59:59.000Z

445

Diagnostics for hybrid reactors  

SciTech Connect

The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

Orsitto, Francesco Paolo [ENEA Unita' Tecnica Fusione , Associazione ENEA-EURATOM sulla Fusione C R Frascati v E Fermi 45 00044 Frascati (Italy)

2012-06-19T23:59:59.000Z

446

THERMAL NUCLEAR REACTOR  

DOE Patents (OSTI)

Nuclear reactors of the graphite moderated air cooled type in which canned slugs or rods of fissile material are employed are discussed. Such a reactor may be provided with a means for detecting dust particles in the exhausted air. The means employed are lengths of dust absorbent cord suspended in vertical holes in the shielding structure above each vertical coolant flow channel to hang in the path of the cooling air issuing from the channels, and associated spindles and drive motors for hauling the cords past detectors, such as Geiger counters, for inspecting the cords periodically. This design also enables detecting the individual channel in which a fault condition may have occurred.

Fenning, F.W.; Jackson, R.F.

1957-09-24T23:59:59.000Z

447

MEANS FOR SHIELDING REACTORS  

DOE Patents (OSTI)

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.

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