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Note: This page contains sample records for the topic "reactor dome removal" from the National Library of EnergyBeta (NLEBeta).
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1

Heavy Water Test Reactor Dome Removal  

SciTech Connect (OSTI)

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

None

2011-01-01T23:59:59.000Z

2

Reactor for removing ammonia  

DOE Patents [OSTI]

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

Luo, Weifang (Livermore, CA); Stewart, Kenneth D. (Valley Springs, CA)

2009-11-17T23:59:59.000Z

3

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect (OSTI)

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

4

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

SciTech Connect (OSTI)

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

Savannah River Site Removes Dome, Opening Reactor for Recovery Act  

Office of Environmental Management (EM)

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6

TMI-2 reactor vessel head removal  

SciTech Connect (OSTI)

This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1984-12-01T23:59:59.000Z

7

TMI-2 reactor vessel head removal  

SciTech Connect (OSTI)

This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1985-09-01T23:59:59.000Z

8

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

SciTech Connect (OSTI)

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

Mackey, Thomas C.; Sutey, Michael J.

2013-06-10T23:59:59.000Z

9

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

SciTech Connect (OSTI)

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

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

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

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. The River Corridor is a 220-square-mile section of...

11

Shutdown heat removal system reliability in thermal reactors  

SciTech Connect (OSTI)

An analysis of the failure probability per year of the shutdown heat removal system (SHRS) at hot standby conditions for two thermal reactor designs is presented. The selected reactor designs are the Pressurized Water Reactor and the Nonproliferation Alternative System Assessment Program Heavy Water Reactor. Failures of the SHRS following the initiating transients of loss of offsite power and loss of main feedwater system are evaluated. Common mode failures between components are incorporated in this anlaysis via the ..beta..-factor method and the sensitivity of the system reliability to common mode failures is investigated parametrically.

Sun, Y.H.; Bari, R.A.

1980-01-01T23:59:59.000Z

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

Detroit Edison's Fermi 1 - Preparation for Reactor Removal  

SciTech Connect (OSTI)

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

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

2008-01-15T23:59:59.000Z

14

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

Office of Environmental Management (EM)

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15

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion | Department ofT ib l L dDepartment ofList?Department09 Section 9990 RecoverDepartment of

16

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

SciTech Connect (OSTI)

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

17

Removable check valve for use in a nuclear reactor  

DOE Patents [OSTI]

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

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

1988-01-01T23:59:59.000Z

18

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

19

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

Colburn, Richard P. (Pasco, WA)

1986-01-01T23:59:59.000Z

20

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

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

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

22

100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT  

SciTech Connect (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

23

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

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. 5 figures.

Ekeroth, D.E.; Orr, R.

1993-12-07T23:59:59.000Z

25

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

26

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

SciTech Connect (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

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

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

29

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

SciTech Connect (OSTI)

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

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

2003-09-01T23:59:59.000Z

30

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

31

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

SciTech Connect (OSTI)

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

32

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

SciTech Connect (OSTI)

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

33

Method and means for remote removal of guide balls from nuclear reactor control rods  

SciTech Connect (OSTI)

This patent describes a method of remotely removing guide balls from nuclear reactor control rods using a punch mechanism, comprising: (a) providing attachment means in the punch mechanism for attaching the punch mechanism to means for reversibly lowering the punch mechanism over the top of one of the control rods; (b) providing a die within the punch mechanism; (c) providing cylinder means within the punch mechanism operatively connected to the die for axially moving the die in a back-and-forth direction; (d) providing a die block within the punch mechanism cooperating with the die; (e) providing guide means within the punch mechanism for self-aligning the punch mechanism so that the die and the die block are automatically aligned with a first one of the guide balls therebetween when the punch mechanism is lowered over the top of the control rod; (f) lowering the punch mechanism over the control rod so that the die, the die block, and the first guide ball are in alignment; and (g) then operating the cylinder means so that the die advances into the die block, thereby removing the first guide ball from the control rod.

Krieg, A.H.

1988-11-29T23:59:59.000Z

34

Analysis of Rod Removal Transient Experiments in VVER Reactors at Zero Power  

SciTech Connect (OSTI)

Within the context of the Fissile Materials Disposition Program of the U.S. Department of Energy we analyzed rod removal transient experiments performed at the Kurchatov Institute in a full-scale mockup of VVER reactors, The transients were started (via water inlet) in slightly (few cents) supercritical configurations with all the control rods withdrawn. After a few minutes, control rods banks or individual control rods were f and t inserted and later withdrawn (returning to the initial state). Available experimental data include the relative time profiles of nine incore and excore detectors. Because of the mild nature of the transients (very low power and no more than 2 $ reactivities) we decided to use a quasistatic approach. The time-dependent flux is factorized into two terms: a function of phase space, given by the solution of the static equation with parametric excitation; and a function of time, given by the solution of the point kinetic equations with time-dependent kinetics para meters. Due to the nature of the experiment, cold conditions, control rods withdrawn and critical state with water level, the power distributions, measured and calculated, are quite unusual, with the inner part of the core heavily shielded. Measured power levels at the center of the reactor are almost 20 times smaller than similar regions at the periphery. Transport and diffusion calculations of the power distributions are in reasonable agreement, so the division code BOLD-VENTURE was used to calculate the kinetics parameters and the relative changes of the detector field of view. The numerical integration of the time-dependent part of the solution was made with the LSODE package using ENDFIB-V and VI delayed neutron data. Very good results were obtained for the nine lime profiles.

Difillippo, F.C.

2000-05-07T23:59:59.000Z

35

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

SciTech Connect (OSTI)

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

Dolan, T.J.; Longhurst, G.R. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Garcia-Otero, E. (Missouri Univ., Columbia, MO (United States). Dept. of Nuclear Engineering)

1992-01-01T23:59:59.000Z

36

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

SciTech Connect (OSTI)

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

Dolan, T.J.; Longhurst, G.R. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Garcia-Otero, E. [Missouri Univ., Columbia, MO (United States). Dept. of Nuclear Engineering

1992-09-01T23:59:59.000Z

37

Removal of volatile organic compounds from polluted air in a reverse flow reactor: An experimental study  

SciTech Connect (OSTI)

An experimental study of the reverse flow reactor for the purification of contaminated air has been carried out. An experimental reactor with an inner diameter of 0.145 m has been constructed. It almost completely reached the goal of an adiabatically operating system. The influence of several operating parameters such as gas velocity, cycle period, chemical character, and concentration of the pollutants and reactor pressure are discussed. The reactor could be operated autothermally provided that the inlet concentrations were sufficiently high. If a mixture of contaminants is fed to the reactor, it might be necessary to increase the total hydrocarbon concentration to assure an autothermal process. Increasing the reactor pressure will hardly change the axial temperature profiles, if the mass flux is kept constant. Increasing the mass flow rate will lead to a higher plateau temperature. Not only the reactor behavior at fixed operating conditions, but also the response of the reactor toward variations in inlet conditions is reported.

Beld, B. van de; Borman, R.A.; Derkx, O.R.; Woezik, B.A.A. van; Westerterp, K.R. (Univ. of Twente, Enschede (Netherlands). Dept. of Chemical Engineering)

1994-12-01T23:59:59.000Z

38

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

39

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

SciTech Connect (OSTI)

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

40

Equilibrium analysis of masonry domes  

E-Print Network [OSTI]

This thesis developed a new method to analyze the structural behavior of masonry domes: the modified thrust line analysis. This graphical-based method offers several advantages to existing methods. It is the first to account ...

Lau, Wanda W

2006-01-01T23:59:59.000Z

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

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

E-Print Network [OSTI]

/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...

Cocheme, Francois Guilhem

2005-02-17T23:59:59.000Z

42

Nitrogen removal via nitrite in a sequencing batch reactor treating sanitary landfill leachate  

E-Print Network [OSTI]

Ammonium rich wastewater Fuzzy logic Biological nutrient removal a b s t r a c t The present paper reports and biological pro- cesses (Lema et al., 1988). Among several biological treatment sys- tems, sequencing batch confirm the effectiveness of the nitrite route for nitrogen removal optimisation in leachate treatment

43

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

Office of Environmental Management (EM)

so the plant can be torn down. Nearly 2,000 capsules of highly radioactive cesium and strontium need to be removed from water-filled storage basins and placed in dry storage....

44

A simplified model of aerosol removal by natural processes in reactor containments  

SciTech Connect (OSTI)

Simplified formulae are developed for estimating the aerosol decontamination that can be achieved by natural processes in the containments of pressurized water reactors and in the drywells of boiling water reactors under severe accident conditions. These simplified formulae were derived by correlation of results of Monte Carlo uncertainty analyses of detailed models of aerosol behavior under accident conditions. Monte Carlo uncertainty analyses of decontamination by natural aerosol processes are reported for 1,000, 2,000, 3,000, and 4,000 MW(th) pressurized water reactors and for 1,500, 2,500, and 3,500 MW(th) boiling water reactors. Uncertainty distributions for the decontamination factors and decontamination coefficients as functions of time were developed in the Monte Carlo analyses by considering uncertainties in aerosol processes, material properties, reactor geometry and severe accident progression. Phenomenological uncertainties examined in this work included uncertainties in aerosol coagulation by gravitational collision, Brownian diffusion, turbulent diffusion and turbulent inertia. Uncertainties in aerosol deposition by gravitational settling, thermophoresis, diffusiophoresis, and turbulent diffusion were examined. Electrostatic charging of aerosol particles in severe accidents is discussed. Such charging could affect both the coagulation and deposition of aerosol particles. Electrostatic effects are not considered in most available models of aerosol behavior during severe accidents and cause uncertainties in predicted natural decontamination processes that could not be taken in to account in this work. Median (50%), 90 and 10% values of the uncertainty distributions for effective decontamination coefficients were correlated with time and reactor thermal power. These correlations constitute a simplified model that can be used to estimate the decontamination by natural aerosol processes at 3 levels of conservatism. Applications of the model are described.

Powers, D.A.; Washington, K.E.; Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States); Burson, S.B. [Nuclear Regulatory Commission, Washington, DC (United States)

1996-07-01T23:59:59.000Z

45

Salt dome discoveries mounting in Mississippi  

SciTech Connect (OSTI)

Exploratory drilling around piercement salt domes in Mississippi has met with a string of successes in recent months. Exploration of these salt features is reported to have been initiated through the review of non-proprietary, 2D seismic data and subsurface control. This preliminary data and work were then selectively upgraded by the acquisition of additional, generally higher quality, conventional 2D seismic lines. This current flurry of successful exploration and ensuing development drilling by Amerada Hess Corp. on the flanks of salt domes in Mississippi has resulted in a number of significant Hosston discoveries/producers at: Carson salt dome in Jefferson Davis County; Dry Creek salt dome in Covington County, Midway salt dome in lamar County, Monticello salt dome in Lawrence County, and Prentiss salt dome in Jefferson Davis County. The resulting production from these fields is gas and condensate, with wells being completed on 640 acre production units.

Ericksen, R.L. [Mississippi Office of Geology, Jackson, MS (United States)

1996-06-17T23:59:59.000Z

46

Heat removal aspects of Liquid Metal Fast Breeder Reactor safety in light of the Three Mile Island incident  

SciTech Connect (OSTI)

The safety aspects of the Liquid Metal Fast Breeder Reactor (LMFBR) loop design are compared with those of the Light Water Reactor (LWR), in light of the Three Mile Island (TMI) incident. The events at TMI are briefly described, the fundamental differences between the LWR water coolant and the LMFBR sodium coolant are presented, and the design of analogous LMFBR safety systems under similar events as those at TMI is discussed. A preliminary qualitative evaluation of a TMI-equivalent accident for an LMFBR indicates that there is likely to be: (1) negligible pressure transients in the primary loop, (2) no core damage, (3) isolation of the incident at the steam generator, and (4) no radiation release to the environment, except a negligible amount of tritium from the secondary sodium. Furthermore, with the absence of the ECCS (Emergency Core Cooling System), pressurizer, and other pressure-related components in the LMFBR design, operator action for a LMFBR should be much simpler in dealing with the coolant upset condition and the decay heat removal problems.

Victor, H.R.; Graf, D.G.

1980-12-01T23:59:59.000Z

47

Lava Dome | Open Energy Information  

Open Energy Info (EERE)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia: Energy Resources Jump to:46 - 429Lacey, Washington:Lakeville, MN)Lauderhill, Florida: EnergyLaurelDome Jump to:

48

Radar investigation of the Hockley salt dome  

E-Print Network [OSTI]

: Geophysics RADAR INVESTIGATION OF THE HOCKLEY SALT DOME A Thesis by UAMES ANDREW HLUCHANEK A'pproved as to style and content by: (Head of Departme t ? Member) May 1. 973 ABSTRACT Radar investigation of the Hockley Salt Dome. . (Nay, 1973) James.... THE PROBLEM. Page A. Probing into Unknown Areas in Salt. . B. Equipment Used. II. BACKGROUND MATERIAL. A. Geology of the Hockley Area. . . B. Economic History of the Hockley Dome Area. . 6 1. Oil 2. Gypsum. 3. Salt C. Geophysical Surveys Over...

Hluchanek, James Andrew

1973-01-01T23:59:59.000Z

49

Development of a countercurrent multistage fluidized-bed reactor and mathematical modeling for prediction of removal efficiency of sulfur dioxide from flue gases  

SciTech Connect (OSTI)

A bubbling countercurrent multistage fluidized-bed reactor for the sorption of sulfur dioxide by hydrated lime particles was simulated employing a two-phase model, with the bubble phase assumed to be in plug flow and with the emulsion phase either in plug flow (EGPF model) or in perfectly mixed flow (EGPM model). The model calculations were compared with experimental data in term of percentage removal efficiency of sulfur dioxide. Both models were applied to understand the influence of operating parameters on the reactor performance. The comparison showed that the EGPF model agreed well with the experimental data. From the perspective of use of a multistage fluidized-bed reactor as air pollution control equipment in industry, the model could be considered general enough for predicting the performance of reactors for gas-solid treatment.

Mohanty, C.R.; Malavia, G.; Meikap, B.C. [Indian Institute of Technology, Kharagpur (India). Dept. of Chemical Engineering

2009-02-15T23:59:59.000Z

50

Dome Tech | Open Energy Information  

Open Energy Info (EERE)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 NoPublic Utilities Address:011-DNA Jump to:52c8ff988c1DeringDolgeville, New York: Energy Resources JumpDolton,Dome

51

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING  

SciTech Connect (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

52

Nuclear reactor engineering  

SciTech Connect (OSTI)

Chapters are presented concerning energy from nuclear fission; nuclear reactions and radiations; diffusion and slowing-down of neutrons; principles of reactor analysis; nuclear reactor kinetics and control; energy removal; non-fuel reactor materials; the reactor fuel system; radiation protection and environmental effects; nuclear reactor shielding; nuclear reactor safety; and power reactor systems.

Glasstone, S.; Sesonske, A.

1981-01-01T23:59:59.000Z

53

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 (OSTI)

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

54

Decommissioning of the high flux beam reactor at Brookhaven Lab  

SciTech Connect (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

55

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

56

Geology of Damon Mound Salt Dome, Texas  

SciTech Connect (OSTI)

Geological investigation of the stratigraphy, cap-rock characteristics, deformation and growth history, and growth rate of a shallow coastal diapir. Damon Mound salt dome, located in Brazoria County, has salt less than 600 feet and cap rock less than 100 feet below the surface; a quarry over the dome provides excellent exposures of cap rock as well as overlying Oligocene to Pleistocene strata. These conditions make it ideal as a case study for other coastal diapirs that lack bedrock exposures. Such investigations are important because salt domes are currently being considered by chemical waste disposal companies as possible storage and disposal sites. In this book, the author reviews previous research, presents additional data on the subsurface and surface geology at Damon Mound, and evaluates Oligocene to post-Pleistocene diapir growth.

Collins, E.W.

1989-01-01T23:59:59.000Z

57

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

SciTech Connect (OSTI)

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

58

User-Defined Experiments for DOME Version 1.6  

E-Print Network [OSTI]

). Furthermore, buses often come within WiFi proximity of each other. The DOME testbed is designed to allow for distributing the experiments to each of the buses. The DOME #12;2 portal provides a service to schedule when1 User-Defined Experiments for DOME Version 1.6 University of Massachusetts, Amherst June 2010

Massachusetts at Amherst, University of

59

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

SciTech Connect (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

60

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

SciTech Connect (OSTI)

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

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

2002-02-26T23:59:59.000Z

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

Nuclear reactor engineering  

SciTech Connect (OSTI)

A book is reviewed which emphasizes topics directly related to the light water reactor power plant and the fast reactor power system. Current real-world problems are addressed throughout the text, and a chapter on safety includes much of the postThree Mile Island impact on operating systems. Topics covered include Doppler broadening, neutron resonances, multigroup diffusion theory, reactor kinetics, reactor control, energy removal, nonfuel materials, reactor fuel, radiation protection, environmental effects, and reactor safety.

Glasstone, S.; Sesonske, A.

1982-07-01T23:59:59.000Z

62

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

SciTech Connect (OSTI)

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

Kehle, R.

1980-08-01T23:59:59.000Z

63

Radar investigation of the Cote Blanche salt dome  

E-Print Network [OSTI]

THE COTE BLANCHE SALT DOME. Geology of the Cote Blanche Salt-Dome Azea. . Economic History of the Cote BLanche Salt-Dome Azea, Salt. . Oil and gas. III. ELECTROMAGNETIC WAVE PROPAGATION. . . Radar Speed in Air and in Salt. . . Velocity...' zznd i'r. mzznz 1959) . The east, south, end west flanks are nearly vertical. However, the north side oi the dome is characterised by a massive overhang. A well drilled by Shell Oil Company, Caffrey No. 1, confirmed the presence of a minimum of 3300...

Stewart, Robert Donald

1974-01-01T23:59:59.000Z

64

Salt dome gas storage solves curtailment threat  

SciTech Connect (OSTI)

In November 1981, Valero Transmission Co. (San Antonio, TX) opened two salt-dome storage caverns with a combined capacity of 5 billion CF (1.5 billion of cushion gas, 3.5 of working gas). The facility's maximum deliverability is 400 million CF/day for 9 days; when two more caverns are finished in late 1982, the $55 million complex will be able to sustain that level for 18 days, making Valero less dependent on linepacking and spot sales to avoid curtailing deliveries to its customers.

Watts, J.

1982-04-01T23:59:59.000Z

65

SOLUTION MINING IN SALT DOMES OF THE GULF COAST EMBAYMENT  

SciTech Connect (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

66

Information Sheet 2011/1/29 DOT Folddown Protective Domes  

E-Print Network [OSTI]

to protect solar telescopes by providing a dome that can be folded down flat to minimize temperature and wind effects, this concept can also be used for other situations. The two solar telescope domes that were built are located on top of towers at an altitude of approximately 2300 m on mountains

Rutten, Rob

67

UPWELLING IN THE COSTA RICA DOME BY KLAUS WYRTKI  

E-Print Network [OSTI]

. This current, the Costa Rica Coastal Current, and parts of the North Equatorial Current form a cyclonic circu'the horizontal circula~ion. The upwelling in the dome is caused by the cyclonic flow around the dome. When survey a deep-reaching" eddy transporting lOX 1018 cm.a/sec. "appeared to be separated and to drift north

68

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

69

Neutron behavior, reactor control, and reactor heat transfer. Volume four  

SciTech Connect (OSTI)

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant).

Not Available

1986-01-01T23:59:59.000Z

70

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

SciTech Connect (OSTI)

The primary objective of a remediation project is the safe extraction and disposition of diverse waste forms and materials. Remediation of a solid waste burial ground containing reactor hardware and irradiated debris involves handling waste with the potential to expose workers to significantly elevated dose rates. Therefore, a major challenge confronted by any remediation project is developing work processes that facilitate compliant waste management practices while at the same time implementing controls to protect personnel. Traditional burial ground remediation is accomplished using standard excavators to remove materials from trenches and other excavation configurations often times with minimal knowledge of waste that will be encountered at a specific location. In the case of the 118-K-1 burial ground the isotopic activity postulated in historic documents to be contained in vertical cylindrical silos was sufficient to create the potential for a significant radiation hazard to project personnel. Additionally, certain reported waste forms posed an unacceptably high potential to contaminate the surrounding environment and/or workers. Based on process knowledge, waste management requirements, historic document review, and a lack of characterization data it was determined that traditional excavation techniques applied to remediation of vertical silos would expose workers to unacceptable risk. The challenging task for the 118-K-1 burial ground remediation project team then became defining an acceptable replacement technology or modification of an existing technology to complete the silo remediation. Early characterization data provided a good tool for evaluating the location of potential high exposure rate items in the silos. Quantitative characterization was a different case and proved difficult because of the large diameter of the silos and the potential for variable density of attenuating soils and waste forms in the silo. Consequently, the most relevant information supporting job planning and understanding of the conditions was the data obtained from the gross gamma meter that was inserted into each casing to provide a rough estimate of dose rates in the tubes. No added value was realized in attempting to quantify the source term and/or associate the isotopic activity with a particular actual waste form (e.g., sludge). Implementing the WRM system allowed monitoring of worker and boundary exposure rates from a distance, maintaining compliance with ALARA principles. This system also provided the project team early knowledge of items being removed that had high exposure rates associated with them, thus creating an efficient method of acknowledging an issue and arriving at a solution prior to having an upset condition. An electronic dosimeter with telemetry capability replaced the excavator mounted AMP-100 system approximately half way through remediation of the silos. Much higher connectivity efficiency was derived from this configuration. Increasing the data feed efficiency additionally led to less interruption of the remediation effort. Early in system testing process a process handicap on the excavator operator was acknowledged. A loss of depth perception resulted when maneuvering the excavator and bucket using the camera feed to an in-cab monitor. Considerable practice and mock-up testing allowed this handicap to be overcome. The most significant equipment failures involved the cable connection to the camera mounted between the clamshell bucket jaws and the video splitter in the excavator cab. Rotation of the clamshell bucket was identified as the cause of cable connection failures because of the cyclic twisting motion and continuous mechanical jarring of the connection. In-cab vibration was identified as the culprit in causing connection failures of the video splitter. While these failures were repaired, substantial production time was lost. Ultimately, the decision was made to purchase a second cable and higher quality video splitter eliminate the down time. An engineering improvement for future operations would be i

Teachout, Douglas B. [Vista Engineering Technologies, LLC, Richland, Washington, 99352 (United States); Adamson, Clinton J.; Zacharias, Ames [Washington Closure Hanford, LLC, Richland, Washington, 99352 (United States)

2012-07-01T23:59:59.000Z

71

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

Open Energy Info (EERE)

Geology and Evolution of the Redondo Dome, Valles Caldera, New Mexico Jump to: navigation, search OpenEI Reference LibraryAdd to library Journal Article: Internal Geology and...

72

Icosadeltahedral geometry of fullerenes, viruses and geodesic domes  

E-Print Network [OSTI]

I discuss the symmetry of fullerenes, viruses and geodesic domes within a unified framework of icosadeltahedral representation of these objects. The icosadeltahedral symmetry is explained in details by examination of all of these structures. Using Euler's theorem on polyhedra, it is shown how to calculate the number of vertices, edges, and faces in domes, and number of atoms, bonds and pentagonal and hexagonal rings in fullerenes. Caspar-Klug classification of viruses is elaborated as a specific case of icosadeltahedral geometry.

Antonio Siber

2007-11-22T23:59:59.000Z

73

Effects of H{sub 2}O and particles on the simultaneous removal of SO{sub 2} and fly ash using a fluidized-bed sorbent/catalyst reactor  

SciTech Connect (OSTI)

This study investigated the potential of a fluidized-bed sorbent/catalyst reactor for the simultaneous removals of SO{sub 2} and fly ash from a simulated flue gas containing different H{sub 2}O and particles. Experimental results showed that the removal efficiency of particles and SO{sub 2} was 85%-96% and 5.75-2.97 mg SO{sub 2}/g, respectively, as the H{sub 2}O content was 1.5-5.3%. The activities of sorbent/catalysts for simultaneous removals of SO{sub 2} and particles were inhibited by H{sub 2}O and particles, and the inhibition effects increased with the content of H{sub 2}O. As the H{sub 2}O content increased, the particle size distribution (PSD) of fine particles shifted to the coarse particles. The results of BET analysis show that the obstruction phenomenon of the sorbent/catalyst caused by the particles was diminished with the increased content of H{sub 2}O. The results showed this aggregation phenomenon of fine particles shifted to the coarse particles may cause increased water vapor content in fluidized-bed sorbent/catalyst reactor.

Rau, J.Y.; Chen, J.C.; Wey, M.Y.; Lin, M.D. [National Chung Hsing University, Taichung (Taiwan). Dept. of Environmental Engineering

2009-12-15T23:59:59.000Z

74

Progress Update: Reactor Disassembly Grouting  

SciTech Connect (OSTI)

Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

Cody, Tom

2010-01-01T23:59:59.000Z

75

Progress Update: Reactor Disassembly Grouting  

ScienceCinema (OSTI)

Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

Cody, Tom

2012-06-14T23:59:59.000Z

76

A mechanical model of early salt dome growth  

E-Print Network [OSTI]

of Department) December 1988 A Mechanical Analysis of Early Salt Dome Growth. (December 1988) Frank Albert Irwin, B. S. , Texas A&M University Chair of Advisory Committee: Dr. Raymond C. Fletcher A two-layer superposition model, the lower layer representing... of the sediments results in growth rates much higher than those observed. Analysis of the case with a diffusivity of 104m2/Ka agrees with all observa- tions. A range of diffusivities which will produce a realistic salt dome model is then determined. The lower...

Irwin, Frank Albert

1988-01-01T23:59:59.000Z

77

Site testing for submillimetre astronomy at Dome C, Antarctica  

E-Print Network [OSTI]

Over the past few years a major effort has been put into the exploration of potential sites for the deployment of submillimetre astronomical facilities. Amongst the most important sites are Dome C and Dome A on the Antarctic Plateau, and the Chajnantor area in Chile. In this context, we report on measurements of the sky opacity at 200 um over a period of three years at the French-Italian station, Concordia, at Dome C, Antarctica. We also present some solutions to the challenges of operating in the harsh polar environ- ment. Dome C offers exceptional conditions in terms of absolute atmospheric transmission and stability for submillimetre astron- omy. Over the austral winter the PWV exhibits long periods during which it is stable and at a very low level (0.1 to 0.3 mm). Higher values (0.2 to 0.8 mm) of PWV are observed during the short summer period. Based on observations over three years, a transmission of around 50% at 350 um is achieved for 75% of the time. The 200-um window opens with a typical transmission...

Tremblin, P; Schneider, N; Durand, G Al; Ashley, M C B; Lawrence, J S; Luong-Van, D M; Storey, J W V; Durand, G An; Reinert, Y; Veyssiere, C; Walter, C; Ade, P; Calisse, P G; Challita, Z; Fossat, E; Sabbatini, L; Pellegrini, A; Ricaud, P; Urban, J

2011-01-01T23:59:59.000Z

78

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

SciTech Connect (OSTI)

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

none,

1981-10-01T23:59:59.000Z

79

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.

80

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-

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

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

82

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

SciTech Connect (OSTI)

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

83

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

SciTech Connect (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

84

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

SciTech Connect (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

85

Fractures of the Dammam Dome Carbonate Outcrops: Their Characterization, Development, and Implications for Subsurface Reservoirs.  

E-Print Network [OSTI]

??The exposed Tertiary carbonates of the Dammam Dome present an opportunity to study fractures in outcrops within the oil-producing region of Eastern Saudi Arabia. The… (more)

Al-Fahmi, Mohammed M

2012-01-01T23:59:59.000Z

86

Salt tectonism and seismic stratigraphy of the Upper Jurassic in the Destin Dome Region, northeastern Gulf of Mexico  

E-Print Network [OSTI]

in the Norphlet Formation (Bearden and Mink, 1989). Comprehensive high-quality multifold seismic reflection data have not previously been available for detailed seismic stratigraphic analysis in the Destin Dome region. Establishment of a seismic stratigraphic... DD-2 Destin Dome Exxon 162 17 938 P, V, G, S, SP, R, D DD-3 Destin Dome Sun Oil 166 17 608 P, V, G, S, SP, D DD-4 Destin Dome Gulf 360 20 988 P, V, G, S, SP, R, D DD-5 Destin Dome Chevron 422 22 222 V, G, S, SP, R, D DD-6 Destin Dome Sohio...

MacRae, Grant

1990-01-01T23:59:59.000Z

87

Field Survey of Cactus Crater Storage Facility (Runit Dome)  

SciTech Connect (OSTI)

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

Douglas Miller, Terence Holland

2008-10-31T23:59:59.000Z

88

Large Component Removal/Disposal  

SciTech Connect (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

89

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

90

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

91

Cold domes over the warm pool: a study of the properties of cold domes produced by mesoscale convective systems during TOGA COARE  

E-Print Network [OSTI]

Mesoscale convective systems (MCSs) are known to cool the subcloud layer by the introduction of penetrative downdrafts to the surface, resulting in the formation of cold domes (also known as cold pools). Five MCSs sampled during the Tropical Ocean...

Caesar, Kathy-Ann Lois

1995-01-01T23:59:59.000Z

92

Mesozoic magmatism and granitic dome in the Wugongshan Massif, Jiangxi province and their genetical  

E-Print Network [OSTI]

Mesozoic magmatism and granitic dome in the Wugongshan Massif, Jiangxi province and their genetical,CNRS - Université d'Orleans, 45067 Orleans 2, France Abstract In SE China, a Mesozoic granitic dome coeval and granitic gneisses, and the E­W-trending Late-Paleozoic­Mesozoic Pingxiang and Anfu basins are located along

Boyer, Edmond

93

Fast foldable tent domes Aswin P.L. Jgers a,b  

E-Print Network [OSTI]

-gust accelerations around large obstacles. This applies also to future large solar telescopes. At present two over the whole dome size. Simultaneously, a variety of wind-speed and -direction sensors measure the wind field around the dome. In addition, fast sensitive air- pressure sensors placed on the supporting

Rutten, Rob

94

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

SciTech Connect (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

95

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

SciTech Connect (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

96

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

SciTech Connect (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

97

Selective purge for hydrogenation reactor recycle loop  

SciTech Connect (OSTI)

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

98

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

99

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. 3 figs.

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

1995-04-25T23:59:59.000Z

100

Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion | Department of EnergyDevelopment Accident TolerantDepartment of Energyon Trackremoved

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

Reactor Physics  

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

Reactor Physics Reactor and nuclear physics is a key area of research at INL. Much of the research done in reactor physics can be separated into one of three categories:...

102

Gravity and seismic reflection studies over the Ferguson Crossing Salt Dome, Grimes and Brazos Counties, Texas  

E-Print Network [OSTI]

and not com- pletely understood f orm. The study of salt domes is important because: (1) more than four-fift'ns or all oil and gas accumulations in fields in the Gulf Coast province have been geologically affected by the growth of the salt domes (Halbouty... to obtain a Bouguer anomaly map of the Ferguson Crossing Salt Dome, (2) a seismic reflection study in order to obtain a seismic section of the area of investigation, and (3) an analysis and interpretation of these studies to obtain a reasonable model...

Cordero Ardila, Vladimir Francisco

1977-01-01T23:59:59.000Z

103

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

104

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

105

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 km˛) 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

106

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

Broader source: Energy.gov [DOE]

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

107

Three-dimensional representations of salt-dome margins at four active strategic petroleum reserve sites.  

SciTech Connect (OSTI)

Existing paper-based site characterization models of salt domes at the four active U.S. Strategic Petroleum Reserve sites have been converted to digital format and visualized using modern computer software. The four sites are the Bayou Choctaw dome in Iberville Parish, Louisiana; the Big Hill dome in Jefferson County, Texas; the Bryan Mound dome in Brazoria County, Texas; and the West Hackberry dome in Cameron Parish, Louisiana. A new modeling algorithm has been developed to overcome limitations of many standard geological modeling software packages in order to deal with structurally overhanging salt margins that are typical of many salt domes. This algorithm, and the implementing computer program, make use of the existing interpretive modeling conducted manually using professional geological judgement and presented in two dimensions in the original site characterization reports as structure contour maps on the top of salt. The algorithm makes use of concepts of finite-element meshes of general engineering usage. Although the specific implementation of the algorithm described in this report and the resulting output files are tailored to the modeling and visualization software used to construct the figures contained herein, the algorithm itself is generic and other implementations and output formats are possible. The graphical visualizations of the salt domes at the four Strategic Petroleum Reserve sites are believed to be major improvements over the previously available two-dimensional representations of the domes via conventional geologic drawings (cross sections and contour maps). Additionally, the numerical mesh files produced by this modeling activity are available for import into and display by other software routines. The mesh data are not explicitly tabulated in this report; however an electronic version in simple ASCII format is included on a PC-based compact disk.

Rautman, Christopher Arthur; Stein, Joshua S.

2003-01-01T23:59:59.000Z

108

Variability of bottom water domes and geostrophic currents in the eastern Gulf of Maine  

E-Print Network [OSTI]

entering Georges Bas- in, the slope water mixes with the endemic, less-dense bottom water (Hopkins and Gar- field, 1979). The newly formed bottom water accumulates in Georges Basin, resulting in an upward doming of the interface (found between 50 and 200...VARIABILITY OF BOTTOM WATER DOMES AND GEOSTROPHIC CURRENTS IN THE EASTERN GULF OF MAINE A Thesis by ERIK SAUL GQTTLIEB Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirement for the degree...

Gottlieb, Erik S

1987-01-01T23:59:59.000Z

109

An investigation of the subsurface Bouguer anomaly in the vicinity of shallow salt domes  

E-Print Network [OSTI]

'itted to the anomaly data in vertical profiles. Analysis of the anomalous vertical gravity gradients indicates that such gradients are too minute for purposes of salt dome exploration. However, calculations of the Bouguer anomaly reveal data which would be easily... detected in the field and amenable to geological interpretation. The empirical curves are fourd to be useful in estimating the Bouguer anomaly for salt domes not explicitly represented by the models. ACKNONLEDGEMENTS The author wishes to express his...

Barnes, William Charles

1977-01-01T23:59:59.000Z

110

A comparison between semi-spheroid- and dome-shaped quantum dots coupled to wetting layer  

SciTech Connect (OSTI)

During the epitaxial growth method, self-assembled semi-spheroid-shaped quantum dots (QDs) are formed on the wetting layer (WL). However for sake of simplicity, researchers sometimes assume semi-spheroid-shaped QDs to be dome-shaped (hemisphere). In this work, a detailed and comprehensive study on the difference between electronic and transition properties of dome- and semi-spheroid-shaped quantum dots is presented. We will explain why the P-to-S intersubband transition behaves the way it does. The calculated results for intersubband P-to-S transition properties of quantum dots show two different trends for dome-shaped and semi-spheroid-shaped quantum dots. The results are interpreted using the probability of finding electron inside the dome/spheroid region, with emphasis on the effects of wetting layer. It is shown that dome-shaped and semi-spheroid-shaped quantum dots feature different electronic and transition properties, arising from the difference in lateral dimensions between dome- and semi-spheroid-shaped QDs. Moreover, an analogy is presented between the bound S-states in the quantum dots and a simple 3D quantum mechanical particle in a box, and effective sizes are calculated. The results of this work will benefit researchers to present more realistic models of coupled QD/WL systems and explain their properties more precisely.

Shahzadeh, Mohammadreza; Sabaeian, Mohammad, E-mail: Sabaeian@scu.ac.ir [Physics Department, Faculty of Science, Shahid Chamran University of Ahvaz, Ahvaz, 61357-43135 (Iran, Islamic Republic of)

2014-06-15T23:59:59.000Z

111

Percutaneous Ethanol Injection via an Artificially Induced Right Hydrothorax for Hepatocellular Carcinoma in the Hepatic Dome  

SciTech Connect (OSTI)

To evaluate the efficacy of sonographically (US) guided percutaneous ethanol injection (PEI) via an artificially induced right hydrothorax (transthoracic PEI) to treat US-invisible hepatocellular carcinoma (HCC) in the hepatic dome. Five cirrhotic patients with US-invisible HCC in the hepatic dome, who were poor surgical candidates, underwent transthoracic PEI. An artificial right hydrothorax was created by instilling 500 ml saline, and absolute ethanol was injected transhydrothoracically into the hepatic dome lesion under local anesthesia. The success and complications were assessed radiologically. The patients were followed up serologically and radiologically for 12-44 (mean 28.4) months. Twenty-five hydrothoraces were induced. All hydrothoraces enabled US visualization of the entire hepatic dome. Eight of the nine small lesions were treated successfully by the treatment. Two of the three local recurrences were eradicated by repeat transthoracic PEI. One large lesion was treated by a combination of transthoracic and regular PEI. The only complication was one clinically insignificant pneumothorax. Induction of a right hydrothorax is feasible and safe. The hydrothorax enables US visualization of the entire hepatic dome and permits US-guided PEI for HCC in the hepatic dome that otherwise would not be possible.

Kume, Akimichi, E-mail: kumea@med.nagoya-u.ac.jp; Nimura, Yuji; Kamiya, Junichi; Nagino, Masato; Kito, Yasushi [Nagoya University Graduate School of Medicine, 65 Tsurumai-cho, Showa-ku, Nagoya, 466-8550, Division of Surgical Oncology, Department of Surgery (Japan)

2003-11-15T23:59:59.000Z

112

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

SciTech Connect (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

113

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

SciTech Connect (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

114

F Reactor Inspection  

SciTech Connect (OSTI)

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29T23:59:59.000Z

115

F Reactor Inspection  

ScienceCinema (OSTI)

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-11-24T23:59:59.000Z

116

Nuclear divisional reactor  

SciTech Connect (OSTI)

A nuclear divisional reactor including a reactor core having side and top walls, a heat exchanger substantially surrounding the core, the heat exchanger including a plurality of separate fluid holding and circulating chambers each in contact with a portion of the core, control rod means associated with the core and external of the heat exchanger including control rods and means for moving said control rods, each of the chambers having separate means for delivering and removing fluid therefrom, separate means associated with each of the delivering and removing means for producing useable energy external of the chambers, each of the means for producing useable energy having separate variable capacity energy outputs thereby making available a plurality of individual sources of useable energy of varying degrees.

Administratrix, A.P.; Rugh, J.L.

1982-11-02T23:59:59.000Z

117

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

118

Degradation of dome cutting minerals in Hanford waste  

SciTech Connect (OSTI)

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

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

2013-01-11T23:59:59.000Z

119

Degradation of Dome Cutting Minerals in Hanford Waste - 13100  

SciTech Connect (OSTI)

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

Reynolds, Jacob G.; Cooke, Gary A.; Huber, Heinz J. [Washington River Protection Solutions, LLC, P.O. Box 850, Richland, WA 99352 (United States)] [Washington River Protection Solutions, LLC, P.O. Box 850, Richland, WA 99352 (United States)

2013-07-01T23:59:59.000Z

120

Liquid metal cooled nuclear reactors with passive cooling system  

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 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

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

Submillimetre/TeraHertz Astronomy at Dome C with CEA filled bolometer array  

E-Print Network [OSTI]

Submillimetre/TeraHertz (e.g. 200, 350, 450 microns) astronomy is the prime technique to unveil the birth and early evolution of a broad range of astrophysical objects. A major obstacle to carry out submm observations from ground is the atmosphere. Preliminary site testing and atmospheric transmission models tend to demonstrate that Dome C could offer the best conditions on Earth for submm/THz astronomy. The CAMISTIC project aims to install a filled bolometer-array camera with 16x16 pixels on IRAIT at Dome C and explore the 200-$\\mu$m windows for potential ground-based observations.

Vincent Minier; Gilles Durand; Pierre-Olivier Lagage; Michel Talvard; Tony Travouillon; Maurizio Busso; Gino Tosti

2007-02-19T23:59:59.000Z

122

Capillary-Pumped Passive Reactor Concept for Space Nuclear Power  

SciTech Connect (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

123

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

124

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

125

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

126

Dome-shaped microresonators and the Born-Oppenheimer Jens U. Nockela and David H. Fosterb  

E-Print Network [OSTI]

Dome-shaped microresonators and the Born-Oppenheimer method Jens U. N¨ockela and David H. Fosterb a explore the Born-Oppenheimer method as an alternative to the paraxial approximation. The conditions the major results of paraxial theory can also be derived from the Born-Oppenheimer method. We discuss how

Nöckelm, Jens

127

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...

Gaviria Garcia, Ricardo

2006-04-12T23:59:59.000Z

128

Architecture is frozen music. Bragdon This publication accompanies the traveling exhibition Pulse Dome Project: Art &  

E-Print Network [OSTI]

a form of sustainable architecture that was in harmony with natural processes--a structure he calledpulse dome #12;Architecture is frozen music. Bragdon #12;This publication accompanies the traveling architecture, wombs, and such natural forms as caves, tunnels, and volcanoes to learn what had been done

Kunkle, Tom

129

Pinch-the-Sky Dome: Freehand Multi-Point Interactions with  

E-Print Network [OSTI]

simple gestures and without on-body trackers. We also aimed to highlight the increasing availability around the Solar system, visit the outskirts of the known universe, and observe the incredible imagery different applications shown in the dome: a) World Wide Telescope (e.g., Solar System visualization), b

Benko, Hrvoje

130

Supplemental DOME Documentation for Researchers with Bricks Version 1.6  

E-Print Network [OSTI]

, the script /diesel/init.pl is executed. This is DOME's bootstrap script; it configures the brick, starts address so that it can be inherited by a VM. o Start a process to manage the 3G modem (/diesel/bustracker3gx.py). #12;2 o Start the gpsd daemon. o Start a process to install any updates (/diesel

Massachusetts at Amherst, University of

131

UPb SHRIMP zircon geochronology and Ttd history of the Kampa Dome, southern Tibet  

E-Print Network [OSTI]

U­Pb SHRIMP zircon geochronology and T­t­d history of the Kampa Dome, southern Tibet M.C. Quigley a al., 2004; Aoya et al., 2005, 2006; Quigley et al., 2006; Lee et al., 2006). Several workers have

Sandiford, Mike

132

Ar thermochronology of the Kampa Dome, southern Tibet: Implications for tectonic evolution  

E-Print Network [OSTI]

of the North Himalayan gneiss domes Mark Quigley a,, Yu Liangjun b , Liu Xiaohan b , Christopher J.L. Wilson: +61 3 8344 7761. E-mail address: m.quigley2@pgrad.unimelb.edu.au (M. Quigley). 0040-1951/$ - see front

Sandiford, Mike

133

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

134

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

135

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

SciTech Connect (OSTI)

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

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

2006-03-01T23:59:59.000Z

136

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

137

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

138

TA-2 Water Boiler Reactor Decommissioning Project  

SciTech Connect (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

139

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites  

SciTech Connect (OSTI)

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. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Best, Ralph E.; Ross, Steven B. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River National Laboratory, Aiken, SC (United States); McConnell, Paul E. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

2013-09-30T23:59:59.000Z

140

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

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

Reactor refueling containment system  

DOE Patents [OSTI]

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

1995-01-01T23:59:59.000Z

142

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

143

The origin of the structural depression above Gulf coast salt domes with particular reference to Clay Creek dome, Washington County, Texas  

E-Print Network [OSTI]

Creek, indioatos that the ~ ouroe layer is at least 17, 000 feet below the surfaoe and is older than F. E. Heath, J. A. Waters, and W. B. Ferguson, op. oit. c p, A3. 8, C. W. Saith, "Gulf Coast Oil Fields", The World Oil, Vol. 130, Eo, 7 {June, 1950... information on salt dome geology published sinoe 1936. However, muoh of the pertinent literature since that date consists of field development data with little to no discussion of struotural prooesses ~ An impsrtant exoeption to this apparently diminished...

McDowell, Alfred Norman

1951-01-01T23:59:59.000Z

144

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

Broader source: Energy.gov [DOE]

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

145

TMI defueling project fuel debris removal system  

SciTech Connect (OSTI)

The three mile Island Unit 2 (TMI-2) pressurized water reactor loss-of-coolant accident on March 28, 1979, presented the nuclear community with many challenging remediation problems; most importantly, the removal of the fission products within the reactor containment vessel. To meet this removal problem, an air-lift system (ALS) can be used to employ compressed air to produce the motive force for transporting debris. Debris is separated from the transport stream by gravity separation. The entire method does not rely on any moving parts. Full-scale testing of the ALS at the Idaho National Engineering Laboratory (INEL) has demonstrated the capability of transporting fuel debris from beneath the LCSA into a standard fuel debris bucket at a minimum rate of 230 kg/min.

Burdge, B.

1992-01-01T23:59:59.000Z

146

TMI defueling project fuel debris removal system  

SciTech Connect (OSTI)

The three mile Island Unit 2 (TMI-2) pressurized water reactor loss-of-coolant accident on March 28, 1979, presented the nuclear community with many challenging remediation problems; most importantly, the removal of the fission products within the reactor containment vessel. To meet this removal problem, an air-lift system (ALS) can be used to employ compressed air to produce the motive force for transporting debris. Debris is separated from the transport stream by gravity separation. The entire method does not rely on any moving parts. Full-scale testing of the ALS at the Idaho National Engineering Laboratory (INEL) has demonstrated the capability of transporting fuel debris from beneath the LCSA into a standard fuel debris bucket at a minimum rate of 230 kg/min.

Burdge, B.

1992-08-01T23:59:59.000Z

147

Quasi-static rock mechanics data for rocksalt from three Strategic Petroleum Reserve domes  

SciTech Connect (OSTI)

Triaxial compression and extension experiments have been run on rocksalt samples from three Strategic Petroleum Reserve (SPR) domes. Seventeen quasi-static tests were loaded at mean stress rates of .66 to 1.04 psi/sec (4.5 to 7.2 kPa/sec), confining pressures of 14.5 to 2000 psi (0.1 to 13.8 MPa) and temperatures of 22 to 100/sup 0/C. Eleven of the test specimens were from Bryan Mound, Texas, and three each were from Bayou Choctaw, Louisiana, and West Hackberry, Louisiana. In general, the resulting mechanical data from the three domes are similar, and they are consistent with previously published data. Ultimate sample strengths are directly related to confining pressure (least principal stress) and indirectly related to temperature, while ductility increases with both pressure and temperature.

Price, R.H.; Wawersik, W.R.; Hannum, D.W.; Zirzow, J.A.

1981-12-01T23:59:59.000Z

148

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

149

Research reactors - an overview  

SciTech Connect (OSTI)

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

150

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

151

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

152

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

153

Light Water Reactor Sustainability  

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

3 Light Water Reactor Sustainability Program ACCOMPLISHMENTS REPORT 2013 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

154

Light Water Reactor Sustainability  

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

4 Light Water Reactor Sustainability ACCOMPLISHMENTS REPORT 2014 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

155

Reactor pressure vessel head vents and methods of using the same  

DOE Patents [OSTI]

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

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

2014-10-28T23:59:59.000Z

156

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

SciTech Connect (OSTI)

As a methodology demonstration for the Office of Nuclear Waste Isolation (ONWI), the Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) Program conducted an initial reference site analysis of the long-term effectiveness of a salt dome repository. The Hainesville Salt Dome in Texas was chosen to be representative of the Gulf Coast interior salt domes; however, the Hainesville Site has been eliminated as a possible nuclear waste repository site. The data used for this exercise are not adequate for an actual assessment, nor have all the parametric analyses been made that would adequately characterize the response of the geosystem surrounding the repository. Additionally, because this was the first exercise of the complete AEGIS and WASTE Rock Interaction Technology (WRIT) methodology, this report provides the initial opportunity for the methodology, specifically applied to a site, to be reviewed by the community outside the AEGIS. The scenario evaluation, as a part of the methodology demonstration, involved consideration of a large variety of potentially disruptive phenomena, which alone or in concert could lead to a breach in a salt dome repository and to a subsequent transport of the radionuclides to the environment. Without waste- and repository-induced effects, no plausible natural geologic events or processes which would compromise the repository integrity could be envisioned over the one-million-year time frame after closure. Near-field (waste- and repository-induced) effects were excluded from consideration in this analysis, but they can be added in future analyses when that methodology development is more complete. The potential for consequential human intrusion into salt domes within a million-year time frame led to the consideration of a solution mining intrusion scenario. The AEGIS staff developed a specific human intrusion scenario at 100 years and 1000 years post-closure, which is one of a whole suite of possible scenarios. This scenario resulted in the delivery of radionuclidecontaminated brine to the surface, where a portion was diverted to culinary salt for direct ingestion by the existing population. Consequence analyses indicated calculated human doses that would be highly deleterious. Additional analyses indicated that doses well above background would occur from such a scenario t even if it occurred a million years into the future. The way to preclude such an intrusion is for continued control over the repository sitet either through direct institutional control or through the effective passive transfer of information. A secondary aspect of the specific human intrusion scenario involved a breach through the side of the salt dome t through which radionuclides migrated via the ground-water system to the accessible environment. This provided a demonstration of the geotransport methodology that AEGIS can use in actual site evaluations, as well as the WRIT program's capabilities with respect to defining the source term and retardation rates of the radionuclides in the repository. This reference site analysis was initially published as a Working Document in December 1979. That version was distributed for a formal peer review by individuals and organizations not involved in its development. The present report represents a revisiont based in part on the responses received from the external reviewers. Summaries of the comments from the reviewers and responses to these comments by the AEGIS staff are presented. The exercise of the AEGIS methodology was sUGcessful in demonstrating the methodologyt and thus t in providing a basis for substantive peer review, in terms of further development of the AEGIS site-applications capability and in terms of providing insight into the potential for consequential human intrusion into a salt dome repository.

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

1982-06-01T23:59:59.000Z

157

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

SciTech Connect (OSTI)

As a methodology demonstration for the Office of Nuclear Waste Isolation (ONWI), the Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) Program conducted an initial reference site analysis of the long-term effectiveness of a salt dome repository. The Hainesville Salt Dome in Texas was chosen to be representative of the Gulf Coast interior salt domes; however, the Hainesville Site has been eliminated as a possible nuclear waste repository site. The data used for this exercise are not adequate for an actual assessment, nor have all the parametric analyses been made that would adequately characterize the response of the geosystem surrounding the repository. Additionally, because this was the first exercise of the complete AEGIS and WASTE Rock Interaction Technology (WRIT) methodology, this report provides the initial opportunity for the methodology, specifically applied to a site, to be reviewed by the community outside the AEGIS. The scenario evaluation, as a part of the methodology demonstration, involved consideration of a large variety of potentially disruptive phenomena, which alone or in concert could lead to a breach in a salt dome repository and to a subsequent transport of the radionuclides to the environment. Without waste- and repository-induced effects, no plausible natural geologic events or processes which would compromise the repository integrity could be envisioned over the one-million-year time frame after closure. Near-field (waste- and repository-induced) effects were excluded from consideration in this analysis, but they can be added in future analyses when that methodology development is more complete. The potential for consequential human intrusion into salt domes within a million-year time frame led to the consideration of a solution mining intrusion scenario. The AEGIS staff developed a specific human intrusion scenario at 100 years and 1000 years post-closure, which is one of a whole suite of possible scenarios. This scenario resulted in the delivery of radionuclidecontaminated brine to the surface, where a portion was diverted to culinary salt for direct ingestion by the existing population. Consequence analyses indicated calculated human doses that would be highly deleterious. Additional analyses indicated that doses well above background would occur from such a scenario t even if it occurred a million years into the future. The way to preclude such an intrusion is for continued control over the repository sitet either through direct institutional control or through the effective passive transfer of information. A secondary aspect of the specific human intrusion scenario involved a breach through the side of the salt dome t through which radionuclides migrated via the ground-water system to the accessible environment. This provided a demonstration of the geotransport methodology that AEGIS can use in actual site evaluations, as well as the WRIT program's capabilities with respect to defining the source term and retardation rates of the radionuclides in the repository. This reference site analysis was initially published as a Working Document in December 1979. That version was distributed for a formal peer review by individuals and organizations not involved in its development. The present report represents a revisiont based in part on the responses received from the external reviewers. Summaries of the comments from the reviewers and responses to these comments by the AEGIS staff are presented. The exercise of the AEGIS methodology was successful in demonstrating the methodologyt and thus t in providing a basis for substantive peer review, in terms of further development of the AEGIS site-applications capability and in terms of providing insight into the potential for consequential human intrusion into a salt dome repository.

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

1982-06-01T23:59:59.000Z

158

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

159

Bioconversion reactor  

DOE Patents [OSTI]

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

McCarty, Perry L. (Stanford, CA); Bachmann, Andre (Palo Alto, CA)

1992-01-01T23:59:59.000Z

160

Microsoft Word - 12-9155160-000 - Pebble Bed Reactor Assessment...  

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

water in the tank. No valves or pumps are required to change state to continue cavity heat removal. Preliminary calculations and Document No.: 12-9155160-000 Pebble Bed Reactor...

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

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor  

SciTech Connect (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

162

Computational neutronics analysis of TRIGA reactors during power pulsing  

E-Print Network [OSTI]

Training, Research, Isotopes, General Atomics (TRIGA) reactors have the unique capability of generating high neutron flux environments with the removal of a transient control rod, creating conditions observed in fast fission ...

Bean, Malcolm (Malcolm K.)

2011-01-01T23:59:59.000Z

163

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab  

Broader source: Energy.gov [DOE]

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

164

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

165

Neutronic reactor  

DOE Patents [OSTI]

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

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

1983-01-01T23:59:59.000Z

166

105-H Reactor Interim Safe Storage Project Final Report  

SciTech Connect (OSTI)

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

167

The Thermal Environment of the Fiber Glass Dome for the New Solar Telescope at Big Bear Solar Observatory  

E-Print Network [OSTI]

The New Solar Telescope (NST) is a 1.6-meter off-axis Gregory-type telescope with an equatorial mount and an open optical support structure. To mitigate the temperature fluctuations along the exposed optical path, the effects of local/dome-related seeing have to be minimized. To accomplish this, NST will be housed in a 5/8-sphere fiberglass dome that is outfitted with 14 active vents evenly spaced around its perimeter. The 14 vents house louvers that open and close independently of one another to regulate and direct the passage of air through the dome. In January 2006, 16 thermal probes were installed throughout the dome and the temperature distribution was measured. The measurements confirmed the existence of a strong thermal gradient on the order of 5 degree Celsius inside the dome. In December 2006, a second set of temperature measurements were made using different louver configurations. In this study, we present the results of these measurements along with their integration into the thermal control system (ThCS) and the overall telescope control system (TCS).

A. P. Verdoni; C. Denker; J. R. Varsik; S. Shumko; J. Nenow; R. Coulter

2007-08-04T23:59:59.000Z

168

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

SciTech Connect (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

169

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

SciTech Connect (OSTI)

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

Neal, J.T. [Sandia National Labs., Albuquerque, NM (United States)] [Sandia National Labs., Albuquerque, NM (United States); Magorian, T.R.; Ahmad, S. [Acres International Corp., Amherst, NY (United States)] [Acres International Corp., Amherst, NY (United States)

1994-11-01T23:59:59.000Z

170

Hydrothermal alteration at the Panorama Formation, North Pole Dome, Pilbara Craton, Western Australia  

E-Print Network [OSTI]

An airborne hyperspectral remote sensing dataset was obtained of the North Pole Dome region of the Pilbara Craton in October 2002. It has been analyzed for indications of hydrothermal minerals. Here we report on the identification and mapping of hydrothermal minerals in the 3.459 Ga Panorama Formation and surrounding strata. The spatial distribution of a pattern of subvertical pyrophyllite rich veins connected to a pyrophyllite rich palaeohorizontal layer is interpreted to represent the base of an acid-sulfate epithermal system that is unconformably overlain by the stromatolitic 3.42 Ga Strelley Pool Chert.

Brown, Adrian J; Walter, Malcolm R

2014-01-01T23:59:59.000Z

171

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

SciTech Connect (OSTI)

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

172

Passive shut-down heat removal system  

DOE Patents [OSTI]

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

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

1988-01-01T23:59:59.000Z

173

The development of self aligning geodesic dome geometries and their application at Chernobyl  

SciTech Connect (OSTI)

William M. Hardy and the author have been actively working towards the goal of establishing a non-profit engineering firm to provide remedial answers to environmental crises. Their immediate concerns are nuclear waste. Their first project has been to provide a design for the second sarcophagus planned at the Chernobyl nuclear power plant in the Ukraine. The sarcophagus will provide a habitat in which to robotically dismantle the power plant and dispose of the radioactive material over a 100+ year period. The Chernobyl Ukritiye Dome structure will `float` on a sufficiently wide, concentrically ``toothed`` footing, supported on a bed of crushed rock so that a minimum of vibrational site disturbance is achieved during construction. This will eliminate the need for a conventional poured concrete foundation and the associated excavating. The Chernobyl Ukritiye Dome will potentially be the single largest structure ever erected in the world. A goal and benefit of this project will be a minimum of exposure to radioactivity for the labor pool because of extensive off site fabrication and as much of the construction as possible will be done robotically. Not only will this structure provide superior long term encapsulation, but once completed, it will become the premiere storage location for the entire region during the next several hundred years.

Hoelzen, E.C.

1995-12-31T23:59:59.000Z

174

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

175

An analysis of research procedures used during the restoration of the dome of the Texas State Capitol, Austin, 1989 to 1995  

E-Print Network [OSTI]

on the exterior of the building. One of the least accessible areas proved to be the dome. This thesis examines the archival and physical research procedures undertaken to determine the condition of the dome, how this information was used, and what discrepancies...

Hocker, Emma Elizabeth

1997-01-01T23:59:59.000Z

176

Investigating the Use of Biosorbents to Remove Arsenic from Water  

E-Print Network [OSTI]

, 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...

Erapalli, Shreyas

2011-02-22T23:59:59.000Z

177

Development of Silica/Vanadia/ Titania Catalysts for Removal of  

E-Print Network [OSTI]

mercury (Hg0) from simulated coal-combustion flue gas. Experiments were carried out in fixed-bed reactorsDevelopment of Silica/Vanadia/ Titania Catalysts for Removal of Elemental Mercury from Coal-Combustion the composition and microstructures of SCR (selective catalytic reduction) catalysts for Hg0 oxidation in coal-combustion

Li, Ying

178

Novel Catalytic Membrane Reactors  

SciTech Connect (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

179

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

180

LBB application in the US operating and advanced reactors  

SciTech Connect (OSTI)

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

Wichman, K.; Tsao, J.; Mayfield, M.

1997-04-01T23:59:59.000Z

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.


181

Hybrid adsorptive membrane reactor  

DOE Patents [OSTI]

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

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

2011-03-01T23:59:59.000Z

182

WCH Removes Massive Test Reactor | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear SecurityTensile Strain Switched Ferromagnetism inS-4500IIVasudhaSurface. |VolunteeringMap2-5:Liquids ReserveWCH

183

Optimizing Medium Baseline Reactor Neutrino Experiments  

E-Print Network [OSTI]

10 years from now medium baseline reactor neutrino experiments will attempt to determine the neutrino mass hierarchy from the observed antineutrino spectra. In this letter we present the results of more than four million detailed simulations of such experiments, studying the dependence of the probability of successfully determining the hierarchy upon the analysis method, the neutrino mass matrix parameters, reactor flux models and, in particular, combinations of baselines. We show that the strong dependence of the hierarchy determination upon mass differences and flux models found by Qian et al. results from a spurious dependence of the Fourier analysis upon the high energy tail of the reactor spectrum which can be removed by using a weighted Fourier transform. Such experiments necessarily use flux from multiple reactors at distinct baselines, smearing the oscillation signal and thus impeding the determination of the hierarchy. Using the results of our simulations, we determine the optimal baselines and corre...

Ciuffoli, Emilio; Zhang, Xinmin

2013-01-01T23:59:59.000Z

184

Conductive incubation and the origin of dome-and-keel structure in Archean granite-greenstone terrains  

E-Print Network [OSTI]

Conductive incubation and the origin of dome-and-keel structure in Archean granite August 2003; accepted 24 October 2003; published 27 January 2004. [1] The Archean East Pilbara Granite duration, following the burial of radiogenic granitic crust beneath the accumulated greenstone pile

Sandiford, Mike

185

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

Broader source: Energy.gov [DOE]

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

186

Triaxial creep measurements on rock salt from the Jennings dome, Louisiana, borehole LA-1, core {number_sign}8  

SciTech Connect (OSTI)

Tejas Power Company requested that facilities in the Rock Mechanics Laboratory at Sandia National Laboratories be used to assess the time-dependent properties of rock salt from the Jennings dome in Acadia Parish, Louisiana. Nominally 2.5-inch diameter slat core from borehole LA-1, core 8 (depth 3924.8 to 3837.8 ft; 1196.8--1197.1 m) was provided to accomplish two tasks: (1) Using the smallest possible number of experiments, evaluate the tendency of Jennings salt to undergo time-dependent deformation (creep) under constant applied stresses, and compare the creep of Jennings salt with creep data for rock salt from other locations. (2) Assess the applicability of published laboratory-derived creep properties for rock salt from several bedded and domal sites in finite element analyses concerning the design of new gas storage caverns in the Jennings dome. The characterization of Jennings salt followed the same strategy that was applied in earlier laboratory experiments on core from the Moss Bluff dome near Houston, Texas. This report summarizes the relevant details of five creep experiments on a sample from depth 3927.5 ft, the results obtained, and how these results compared with laboratory creep measurements gathered on rock salt from other locations including the West Hackberry, Bryan Mound and Moss Bluff domes. The report also considers the estimates of specific creep parameters commonly used in numerical engineering design analyses.

Wawersik, W.R.; Zimmerer, D.J.

1994-05-01T23:59:59.000Z

187

R- AND P- REACTOR VESSEL IN-SITU DECOMISSIONING VISUALIZATION  

SciTech Connect (OSTI)

The R- & P- Reactor facilities were constructed in the early 1950's in response to Cold War efforts. The mission of the facilities was to produce materials for use in the nation's nuclear weapons stockpile. R-Reactor was removed from service in 1964 when President Johnson announced a slowdown of he nuclear arms race. PReactor continued operation until 1988 until the facility was taken off-line to modernize the facility with new safeguards. Efforts to restart the reactor ended in 1990 at the end of the Cold War. Both facilities have sat idle since their closure and have been identified as the first two reactors for closure at SRS.

Vrettos, N.; Bobbitt, J.; Howard, M.

2010-06-07T23:59:59.000Z

188

Proceedings of the international ANS/ENS topical meeting on thermal reactor safety  

SciTech Connect (OSTI)

This book presents the papers given at a conference on reactor safety. Topics considered at the conference included containment, fission product release, safety injection, reactor instrumentation, thermal-hydraulic analyses, reactor safety experiments, meltdowns, gas production rates, computerized simulation, occupational safety, human factors engineering, emergency plans, emergency core cooling systems, reactor licensing, after-heat removal, RHR systems, risk assessment, probability, fuel-cladding interactions, safety standards, and molten metal-water interactions.

Not Available

1986-01-01T23:59:59.000Z

189

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

SciTech Connect (OSTI)

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

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

1996-10-01T23:59:59.000Z

190

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

SciTech Connect (OSTI)

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

Wang, Lingzhi; Zhu, Zonghong [Department of Astronomy, Beijing Normal University, Beijing 100875 (China); Macri, Lucas M.; Wang, Lifan [Mitchell Institute for Fundamental Physics and Astronomy, Department of Physics and Astronomy, Texas A and M University, College Station, TX 77843 (United States); Ashley, Michael C. B.; Lawrence, Jon S.; Luong-Van, Daniel; Storey, John W. V. [School of Physics, University of New South Wales, NSW 2052 (Australia); Cui, Xiangqun; Feng, Long-Long; Gong, Xuefei; Liu, Qiang; Shang, Zhaohui; Yang, Huigen; Yang, Ji; Yuan, Xiangyan; Zhou, Xu; Zhu, Zhenxi [Chinese Center for Antarctic Astronomy, Nanjing 210008 (China); Pennypacker, Carl R. [Center for Astrophysics, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); York, Donald G., E-mail: wanglingzhi@bao.ac.cn [Department of Astronomy and Astrophysics and Enrico Fermi Institute, University of Chicago, Chicago, IL 60637 (United States)

2013-12-01T23:59:59.000Z

191

Geologic technical assessment of the Richton salt dome, Mississippi, for potential expansion of the U.S. strategic petroleum reserve.  

SciTech Connect (OSTI)

Technical assessment and remodeling of existing data indicates that the Richton salt dome, located in southeastern Mississippi, appears to be a suitable site for expansion of the U.S. Strategic Petroleum Reserve. The maximum area of salt is approximately 7 square miles, at a subsurface elevation of about -2000 ft, near the top of the salt stock. Approximately 5.8 square miles of this appears suitable for cavern development, because of restrictions imposed by modeled shallow salt overhang along several sides of the dome. The detailed geometry of the overhang currently is only poorly understood. However, the large areal extent of the Richton salt mass suggests that significant design flexibility exists for a 160-million-barrel storage facility consisting of 16 ten-million-barrel caverns. The dome itself is prominently elongated from northwest to southeast. The salt stock appears to consist of two major spine features, separated by a likely boundary shear zone trending from southwest to northeast. The dome decreases in areal extent with depth, because of salt flanks that appear to dip inward at 70-80 degrees. Caprock is present at depths as shallow as 274 ft, and the shallowest salt is documented at -425 ft. A large number of existing two-dimensional seismic profiles have been acquired crossing, and in the vicinity of, the Richton salt dome. At least selected seismic profiles should be acquired, examined, potentially reprocessed, and interpreted in an effort to understand the limitations imposed by the apparent salt overhang, should the Richton site be selected for actual expansion of the Reserve.

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

2006-01-01T23:59:59.000Z

192

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

SciTech Connect (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

193

Arsenic removal from water  

DOE Patents [OSTI]

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

Moore, Robert C. (Edgewood, NM); Anderson, D. Richard (Albuquerque, NM)

2007-07-24T23:59:59.000Z

194

Flexible Conversion Ratio Fast Reactor Systems Evaluation  

SciTech Connect (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

195

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.

Challberg, Roy C. (Livermore, CA)

1994-01-01T23:59:59.000Z

196

Drum lid removal tool  

DOE Patents [OSTI]

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

Pella, Bernard M. (Martinez, GA); Smith, Philip D. (North Augusta, SC)

2010-08-24T23:59:59.000Z

197

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

198

Methods and apparatus for carbon dioxide removal from a fluid stream  

DOE Patents [OSTI]

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

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

2010-01-19T23:59:59.000Z

199

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

200

SRS Small Modular Reactors  

SciTech Connect (OSTI)

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

None

2012-04-27T23:59:59.000Z

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

SRS Small Modular Reactors  

ScienceCinema (OSTI)

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

None

2014-05-21T23:59:59.000Z

202

Removal design report for the 108-F Biological Laboratory  

SciTech Connect (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

203

Investigation of vessel exterior air cooling for a HLMC reactor  

SciTech Connect (OSTI)

The Secure Transportable Autonomous Reactor (STAR) concept under development at Argonne National Laboratory provides a small (300 MWt) reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100%+ natural circulation heat removal from the low power density/low pressure drop ultra-long lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the Reactor Exterior Cooling System (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the Reactor Vessel Auxiliary Cooling System (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink.

Sienicki, J. J.; Spencer, B. W.

2000-01-13T23:59:59.000Z

204

Investigation of vessel exterior air cooling for an HLMC reactor  

SciTech Connect (OSTI)

The secure transportable autonomous reactor (STAR) concept under development at Argonne National Laboratory provides a small [300-MW(thermal)] reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100% + natural-circulation heat removal from the low-power-density/low-pressure-drop ultralong lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the reactor exterior cooling system (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the reactor vessel auxiliary cooling system (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink.

Sienicki, J.J.; Spencer, B.W.

2000-07-01T23:59:59.000Z

205

Nuclear reactor  

DOE Patents [OSTI]

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

Thomson, Wallace B. (Severna Park, MD)

2004-03-16T23:59:59.000Z

206

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

SciTech Connect (OSTI)

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

207

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

208

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

209

Treatment of methyl t-butyl ether contaminated water using a dense medium plasma reactor, a mechanistic and kinetic investigation  

E-Print Network [OSTI]

and oxidation mechanisms of methyl t-butyl ether (MTBE) in a dense medium plasma (DMP) reactor utilizing gas for the removal of MTBE from an aqueous solution in the DMP reactor. Rate constants are also derived for three reactor configurations and two pin array spin rates. The oxidation products from the treatment of MTBE

Dandy, David

210

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

211

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

212

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. 3 figs.

Neuhaus, J.E.

1992-10-13T23:59:59.000Z

213

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

214

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

SciTech Connect (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

215

KKG Group Paraffin Removal  

SciTech Connect (OSTI)

The Rocky Mountain Oilfield Testing Center (RMOTC) has recently completed a test of a paraffin removal system developed by the KKG Group utilizing the technology of two Russian scientists, Gennady Katzyn and Boris Koggi. The system consisting of chemical ''sticks'' that generate heat in-situ to melt the paraffin deposits in oilfield tubing. The melted paraffin is then brought to the surface utilizing the naturally flowing energy of the well.

Schulte, Ralph

2001-12-01T23:59:59.000Z

216

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

Broader source: Energy.gov [DOE]

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

217

Design options for a bunsen reactor.  

SciTech Connect (OSTI)

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

218

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

SciTech Connect (OSTI)

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

219

Key word:Daylight Factor Window Wall Ratio Pendentive dome Lighting design Tropical region Architecture and Interior design  

E-Print Network [OSTI]

Abstract: Daylighting design plays important role in architecture of religious buildings such as churches and mosques where pendentive dome construction is frequently used. In daylighting design, many designers face difficulty in estimating the interior share of light which is usually expressed by daylight factor due to complexity of interior form. This study aims to provide designers with a rather high precision rule of thumb for average daylight factor in pendentive dome building. Thus, it investigates the Daylight Factor [DF] distribution of such buildings with reference to the tropics. It takes the Window Wall Ratio [WWR] into account and seeks its influence on daylight factor. By a 12 X 12 points grid, it examines five different ratios including 0.1, 0.2, 0.3, 0.4 and 0.5 on DF of the floor beneath the dome. The results endorse the direct relation of WWR and DF. The least WWR equal to 0.1 yields an average DF of 0.55 % while the greatest WWR of 0.5 yields in average DF of 2.56%. The intermediate WWR in steps of 0.2, 0.3 and 0.4 correspond to 1.04, 1.56 and 2.07 percent respectively. As a relatively precise rule of thumb, any increment in consequent steps of WWR with 0.1 intervals results in 0.5 % increase in DF. This can be employed by architects and interior designers for lighting design of pendentive dome buildings in tropical region.

Mehrdad Mazloomi

220

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

SciTech Connect (OSTI)

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

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

1999-11-01T23:59:59.000Z

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

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

SciTech Connect (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

222

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

223

Undergraduate reactor control experiment  

SciTech Connect (OSTI)

A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise.

Edwards, R.M.; Power, M.A.; Bryan, M. (Pennsylvania State Univ., University Park (United States))

1992-01-01T23:59:59.000Z

224

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

225

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

226

Reactor Sharing Program  

SciTech Connect (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

227

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

228

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. 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

229

The TMI defueling project fuel debris removal system  

SciTech Connect (OSTI)

The Three Mile Island (TMI) unit 2 pressurized water reactor loss-of-coolant accident on March 28, 1979, presented the nuclear community with many challenging remediation problems. A plethora of techniques, systems, and tools have been employed for the recovery and packaging of the postaccident configuration of the reactor core. Of particular difficulty was the removal of the fuel debris located beneath the lower core support structure. Fuel debris located beneath the lower core support structure was the result of rapid cooling of the previously molten UO{sub 2} and ZrO{sub 2}, causing formation of a ceramic like rubble. Approximately 19,100 kg of this rubble settled beneath the lower core support structure and onto the lower head of the reactor containment vessel. The development and implementation of a debris collection system based on the air lift principle proved to be an effective method for gathering the fuel debris from beneath the lower core support structure.

Burge, B. (EG and G Idaho, Inc., Idaho Falls (United States))

1992-01-01T23:59:59.000Z

230

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

1993-03-02T23:59:59.000Z

231

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

232

Advanced Test Reactor Tour  

SciTech Connect (OSTI)

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

Miley, Don

2011-01-01T23:59:59.000Z

233

Advanced Test Reactor Tour  

ScienceCinema (OSTI)

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

Miley, Don

2013-05-28T23:59:59.000Z

234

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

235

Hypothetical Reactor Accident Study  

E-Print Network [OSTI]

- W 4 DfcSkoollo Rise-R-427 CARNSORE: Hypothetical Reactor Accident Study O. Walmod-Larsen, N. O: HYPOTHETICAL REACTOR ACCIDENT STUDY O. Walmod-Larsen, N.O. Jensen, L. Kristensen, A. Heide, K.L. NedergĂĄrd, P-basis accident and a series of hypothetical core-melt accidents to a 600 MWe reactor are de- scribed

236

IMPROVED PROCESSES TO REMOVE NAPHTHENIC ACIDS  

SciTech Connect (OSTI)

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

237

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

SciTech Connect (OSTI)

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

238

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 in-line reamer are provided for removing deposits from the inside of the pipe. The assembly enables fluid streams to be injected and distributed uniformly into the particle bed with minimized clogging of injector ports. The same design may also be used for extraction of fluid streams from particle bed reactors.

Notestein, John E. (Morgantown, WV)

1993-01-01T23:59:59.000Z

239

Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico  

SciTech Connect (OSTI)

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

240

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

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

Geothermal hydrogen sulfide removal  

SciTech Connect (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

242

Spinning fluids reactor  

DOE Patents [OSTI]

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

243

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

2012-03-08T23:59:59.000Z

244

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. 1 figure.

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

1994-12-20T23:59:59.000Z

245

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

246

Numerical Procedure to Extract Physical Properties from Raman Scattering Data in a Flow Reactor  

E-Print Network [OSTI]

in an up-flow cold-wall CVD reactor was monitored using in situ Raman spectroscopy. A two uncertainties by removing complicating phenomena or sim- plifying model development. An up-flow cold-wall CVD of the process would be useful, e.g., for optimizing an existing process or designing a new reactor. Accurate

Anderson, Timothy J.

247

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

248

An investigation of urea decomposition and selective non-catalytic removal of nitric oxide with urea  

E-Print Network [OSTI]

]................................................................................... 22 17 The removal of NOx as a function of reactor temperature for seven levels of oxygen for 100 ppm carbon monoxide and a heated residence time of 2.11 secs. [42] ............................................................................. 23 18... The removal of NOx as a function of oxygen concentration for three levels of carbon monoxide (100, 600 and 900 ppm) for a reactor temperature of 1150 K and a heated residence time of 2.11 secs. [42]... 24 19 Influence of CO/NO ratio in the feed...

Park, Yong Hun

2004-09-30T23:59:59.000Z

249

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

SciTech Connect (OSTI)

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

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)] [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01T23:59:59.000Z

250

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

251

Hyperspectral imaging spectroscopy of a Mars analogue environment at the North Pole Dome, Pilbara Craton, Western Australia  

E-Print Network [OSTI]

A visible and near infrared (VNIR) to shortwave infrared (SWIR) hyperspectral dataset of the Early Archaean North Pole Dome, Pilbara Craton, Western Australia, has been analysed for indications of hydrothermal alteration. Occurrence maps of hydrothermal alteration minerals were produced. It was found that using a spatial resolution on the ground of approximately 5 m and spectral coverage from 0.4 to 2.5 mm was sufficient to delineate several hydrothermal alteration zones and associated veins, including phyllic, serpentinitic and chloritic alteration. These results suggest this level of spectral and spatial resolution would be ideal for localising shallow epithermal activity, should such activity have existed, on the surface of Mars.

Brown, Adrian J; Cudahy, Thomas

2014-01-01T23:59:59.000Z

252

Removing Arsenic from Drinking Water  

ScienceCinema (OSTI)

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

None

2013-05-28T23:59:59.000Z

253

Mercury and tritium removal from DOE waste oils  

SciTech Connect (OSTI)

This work covers the investigation of vacuum extraction as a means to remove tritiated contamination as well as the removal via sorption of dissolved mercury from contaminated oils. The radiation damage in oils from tritium causes production of hydrogen, methane, and low-molecular-weight hydrocarbons. When tritium gas is present in the oil, the tritium atom is incorporated into the formed hydrocarbons. The transformer industry measures gas content/composition of transformer oils as a diagnostic tool for the transformers` condition. The analytical approach (ASTM D3612-90) used for these measurements is vacuum extraction of all gases (H{sub 2}, N{sub 2}, O{sub 2}, CO, CO{sub 2}, etc.) followed by analysis of the evolved gas mixture. This extraction method will be adapted to remove dissolved gases (including tritium) from the SRS vacuum pump oil. It may be necessary to heat (60{degrees}C to 70{degrees}C) the oil during vacuum extraction to remove tritiated water. A method described in the procedures is a stripper column extraction, in which a carrier gas (argon) is used to remove dissolved gases from oil that is dispersed on high surface area beads. This method appears promising for scale-up as a treatment process, and a modified process is also being used as a dewatering technique by SD Myers, Inc. (a transformer consulting company) for transformers in the field by a mobile unit. Although some mercury may be removed during the vacuum extraction, the most common technique for removing mercury from oil is by using sulfur-impregnated activated carbon (SIAC). SIAC is currently being used by the petroleum industry to remove mercury from hydrocarbon mixtures, but the sorbent has not been previously tested on DOE vacuum oil waste. It is anticipated that a final process will be similar to technologies used by the petroleum industry and is comparable to ion exchange operations in large column-type reactors.

Klasson, E.T. [Oak Ridge National Lab., TN (United States)

1997-10-01T23:59:59.000Z

254

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

SciTech Connect (OSTI)

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

255

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

256

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

257

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

258

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

259

Next Generation Reactors  

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

Nuclear Advances We are coordinating the Generation IV Nuclear Systems Initiative - an international effort to develop the next generation of nuclear power reactors. Skip...

260

TA-2 water boiler reactor decommissioning (Phase 1)  

SciTech Connect (OSTI)

Removal of external structures and underground piping associated with the gaseous effluent (stack) line from the TA-2 Water Boiler Reactor was performed as Phase I of reactor decommissioning. Six concrete structures were dismantled and 435 ft of contaminated underground piping was removed. Extensive soil contamination by /sup 137/Cs was encountered around structure TA-2-48 and in a suspected leach field near the stream flowing through Los Alamos Canyon. Efforts to remove all contaminated soil were hampered by infiltrating ground water and heavy rains. Methods, cleanup guidelines, and ALARA decisions used to successfully restore the area are described. The cost of the project was approximately $320K; 970 m/sup 3/ of low-level solid radioactive waste resulted from the cleanup operations.

Elder, J.C.; Knoell, C.L.

1986-12-01T23:59:59.000Z

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.


261

RELAP5/MOD3 simulation of the loss of residual heat removal during midloop operation experiment conducted at the ROSA-IV/ Large Scale Test Facility  

E-Print Network [OSTI]

The modeling of the complex thermal hydraulics Of reactor systems involves the use Of experimental test systems as well as numerical codes. A simulation of the loss of residual heat removal (RHR) during midloop operations was performed using...

Banerjee, Sibashis Sanatkumar

2012-06-07T23:59:59.000Z

262

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...  

Office of Scientific and Technical Information (OSTI)

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor Design and Feasibility Problem Re-direct Destination: Temp Data Fields Rosen, M. A.; Coburn, D. B.; Flynn, T....

263

Brookhaven Graphite Research Reactor Workshop  

Broader source: Energy.gov [DOE]

The Brookhaven Graphite Research Reactor (BGRR) was the first reactor built in the U.S. for peacetime atomic research following World War II.  Construction began in 1947 and the reactor started...

264

Portfolio for fast reactor collaboration  

SciTech Connect (OSTI)

The development of the LMFBR type reactor in the United Kingdom is reviewed. Design characteristics of a commercial demonstration fast reactor are presented and compared with the Super Phenix reactor.

Rippon, S.

1981-12-01T23:59:59.000Z

265

REACTOR OPERATIONS AND CONTROL  

E-Print Network [OSTI]

REACTOR OPERATIONS AND CONTROL KEYWORDS: core calculations, neural networks, control rod elevation of a control rod, or a group of control rods, is an important parameter from the viewpoint of reactor control DETERMINATION OF PWR CONTROL ROD POSITION BY CORE PHYSICS AND NEURAL NETWORK METHODS NINOS S. GARIS* and IMRE

Pázsit, Imre

266

Beryllium Use in the Advanced Test Reactor  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) began operation in 1967. It makes use of a unique serpentine fuel core design and a beryllium reflector. Reactor control is achieved with rotating beryllium cylinders to which have been fastened plates of hafnium. Over time, the beryllium develops rather high helium content because of nuclear transmutations and begins to swell. The beryllium must be replaced at nominally 10-year intervals. Determination of when the replacement is made is by visual observation using a periscope to examine the beryllium surface for cracking and swelling. Disposition of the irradiated beryllium was once accomplished in the INL’s Radioactive Waste Management Complex, but that is no longer possible. Among contributing reasons are high levels of specific radioactive contaminants including transuranics. The INL is presently considering disposition pathways for this irradiated beryllium, but presently is storing it in the canal adjacent to the reactor. Numerous issues are associated with this situation including (1) Is there a need for ultra-low uranium material? (2) Is there a need to recover tritium from irradiated beryllium either because this is a strategic material resource or in preparation for disposal? (3) Is there a need to remove activation and fission products from irradiated beryllium? (4) Will there be enough material available to meet requirements for research reactors (fission and fusion)? In this paper will be discussed the present status of considerations on these issues.

Glen R. Longhurst

2007-12-01T23:59:59.000Z

267

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

2013-08-14T23:59:59.000Z

268

Method for automatically scramming a nuclear reactor  

DOE Patents [OSTI]

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

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

2005-12-27T23:59:59.000Z

269

Plasma generators, reactor systems and related methods  

DOE Patents [OSTI]

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

270

Reactor & Nuclear Systems Publications | ORNL  

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

Nuclear Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications...

271

Reed Reactor Facility. Final report  

SciTech Connect (OSTI)

This report discusses the operation and maintenance of the Reed Reactor Facility. The Reed reactor is mostly used for education and train purposes.

Frantz, S.G.

1994-12-31T23:59:59.000Z

272

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

273

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

SciTech Connect (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

274

Performance of Decay Heat Removal Systems in the LS-VHTR  

SciTech Connect (OSTI)

Investigations are underway to determine the viability of the Liquid Salt-Cooled - Very High Temperature Reactor (LS-VHTR) concept which combines fuel and moderator similar to gas cooled VHTR concepts but utilizes liquid salt coolant which can operate at low pressures with improved heat transfer properties relative to helium. Analyses have been carried out investigating the viability of two alternative passive approaches for emergency decay heat removal for a 2400 MWt LS-VHTR: RVACS air natural circulation cooling of the exterior of the guard vessel and DRACS Direct Reactor Heat Exchangers (DRHXs) immersed in the liquid salt coolant and connected to natural draft air heat exchangers through secondary and tertiary cooling circuits. Results of first principles and integrated systems analyses of RVACS and DRACS performance are presented for a postulated accident scenario involving loss-of-normal heat removal, loss-of-forced (pumped) liquid salt flow, and successful scram of the reactor. (authors)

Sienicki, James J.; Moisseytsev, Anton; Farmer, Mitchell T.; Dunn, Floyd E.; Cahalan, James E. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL, 60439 (United States)

2006-07-01T23:59:59.000Z

275

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

276

Design study of the deep-sea reactor X  

SciTech Connect (OSTI)

The deep-sea reactor X (DRX) is a small nuclear plant designed to provide undersea power sources. It has the full advantages of nuclear reactors and can provide large power capacity and does not require oxygen for power production. An application conceivable in the near future is that for a submersible. The Japan Atomic Energy Research Institute is conducting a design study of a 150-kW(electric) DRX plant for a deep-sea research vessel. It has a so-called integrated pressurized water reactor,'' having a steam generator inside the reactor vessel. A pressure shell includes a turbine and a generator as well as a reactor vessel, resulting in a very compact electricity producing plant. It should be easy to operate and have high passive safety characteristics; namely, a short startup time, good reactor response to power demand changes, and passive core flooding and decay heat removal in case of an accident. Transient analyses including those for load follow-up, reactor startup, and accidents have been conducted. The results show that the DRX has excellent inherent characteristics satisfying those requirements.

Iida, Hiromasa (Japan Atomic Energy Research Inst., Ibaraki (Japan)); Ishizaka, Yuichi (Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)); Kim, Y.C.; Yamaguchi, Chouichi (Japan Research Inst., Ltd., Tokyo (Japan))

1994-07-01T23:59:59.000Z

277

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

278

Nuclear reactor control column  

SciTech Connect (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 crosssectional 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, D.M.

1982-08-10T23:59:59.000Z

279

Reactor Safety Research Programs  

SciTech Connect (OSTI)

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

Edler, S. K.

1981-07-01T23:59:59.000Z

280

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

DOE Patents [OSTI]

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

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

1995-01-01T23:59:59.000Z

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

Journal of Volcanology and Geothermal Research69 (1995) 105-l 16 Mount St. Helens and Santiaguito lava domes: The effect of short-  

E-Print Network [OSTI]

of silicic lava flows, we studied surface characteristics and obtained water content and hydrogen isotopic for these patterns is most clearly preserved in lavas erupted during early, rapid stages of dome growth. Petrologists to sample flows early in their emplacement while paying attention to surface texture, position relative

Rose, William I.

282

Growth of Dome-Shaped Carbon Nanoislands on Ir(111): The Intermediate between Carbidic Clusters and Quasi-Free-Standing Graphene  

E-Print Network [OSTI]

of hydrocarbon dissociation on transition metal (TM) sur- faces represents a challenging way to its synthesisGrowth of Dome-Shaped Carbon Nanoislands on Ir(111): The Intermediate between Carbidic Clusters coupled carbidic carbon and a quasi-free-standing graphene layer, can provide information for a rational

Alfè, Dario

283

Fast Breeder Reactor studies  

SciTech Connect (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

284

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

285

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

286

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

SciTech Connect (OSTI)

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

287

Preliminary design study of compressed-air energy storage in a salt dome. Volume 4. CAES turbomachinery design. Final report  

SciTech Connect (OSTI)

A summary is presented of the study undertaken by the Turbomachinery Subcontractor on Task 1, according to instructions received from the Middle South Services (MSS), the US Department of Energy (DOE) and the Electric Power Research Institute (EPRI). The subject of this study was to investigate the question of whether it would be possible to build an air storage generating plant capable of operating economically and using leached-out salt domes as air reservoirs. In the course of the work performed on Task 1, the Turbomachinery Subcontractors have on various occasions supplied information on the results obtained, in the form of preliminary subreports. The present summary includes all the previous reports, most of which have been revised to a large extent.

Not Available

1982-06-01T23:59:59.000Z

288

New developments in measurement technology relevant to the studies of deep geological repositories in domed salt and basalt  

SciTech Connect (OSTI)

This report briefly describes recent geophysical and geotechnical instrumentation developments relevant to the studies of deep geologic repositories. Special emphasis has been placed on techniques that appear to minimize measurement problems associated with repositories constructed in basalt or domed salt. Included in the listing are existing measurement capabilities and deficiencies that have been identified by a few authors and instrumentation workshops that have assessed the capabilities of existing instrumentation with respect to repository applications. These deficiencies have been compared with the reported advantages and limitations of the new developments described. Based on these comparisons, areas that merit further research and development have been identified. The report is based on a thorough literature review and on discussions with several instrumentation specialists involved in instrumentation development.

Ramirez, A.L.; Mao, N.H.

1980-05-21T23:59:59.000Z

289

A method of estimating time scales of atmospheric piston and its application at DomeC (Antarctica)  

E-Print Network [OSTI]

Temporal fluctuations of the atmospheric piston are critical for interferometers as they determine their sensitivity. We characterize an instrumental set-up, termed the piston scope, that aims at measuring the atmospheric time constant, tau0, through the image motion in the focal plane of a Fizeau interferometer. High-resolution piston scope measurements have been obtained at the observatory of Paranal, Chile, in April 2006. The derived atmospheric parameters are shown to be consistent with data from the astronomical site monitor, provided that the atmospheric turbulence is displaced along a single direction. Piston scope measurements, of lower temporal and spatial resolution, were for the first time recorded in February 2005 at the Antarctic site of DomeC. Their re-analysis in terms of the new data calibration sharpens the conclusions of a first qualitative examination.

A. Kellerer; M. Sarazin; T. Butterley; R. Wilson

2007-03-06T23:59:59.000Z

290

Method for loading, operating, and unloading a ball-bed nuclear reactor  

SciTech Connect (OSTI)

This patent describes a method of operating a ball-bed nuclear reactor with fuel element balls. Some have a fissionable material content different from that of others of the balls. It consists of: initially partly filling a reactor core with fuel balls of sufficient fissionable material content for establishing criticality and a desired level of power production at the completion of the partial filling and then, without any further filling of the reactor cavern, starting reactor operation; thereafter without any removal of fuel balls from the reactor cavern, filling fuel balls continually or in groups at relatively short intervals into the reactor cavern during increasing burning up of the fuel balls already, for compensation of the diminishing fissionable material content of the reactor core constituted by the fuel balls until a final total quantity of filling is reached; after the final filling quantity is reached and burning up has occurred, shutting down the reactor, cooling it off, releasing the pressure in the cavern, and thereafter unloading all the fuel balls from the reactor cavern, unloading being begun when the reactor is shut down and being completed before the reactor is restarted.

Teuchert, E.; Haas, K.A.; Gerwin, H.

1987-09-22T23:59:59.000Z

291

Reactor hot spot analysis  

SciTech Connect (OSTI)

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

292

P Reactor Grouting  

SciTech Connect (OSTI)

Filling the P Reactor with grout. This seals the radioactive material and reduces the environmental footprint left from the Cold War. Project sponsored by the Recovery Act at the Savannah River Site.

None

2010-01-01T23:59:59.000Z

293

Nuclear reactor control  

SciTech Connect (OSTI)

A liquid metal cooled fast breeder nuclear reactor has power setback means for use in an emergency. On initiation of a trip-signal a control rod is injected into the core in two stages, firstly, by free fall to effect an immediate power-set back to a safe level and, secondly, by controlled insertion. Total shut-down of the reactor under all emergencies is avoided. 4 claims.

Ingham, R.V.

1980-01-01T23:59:59.000Z

294

Polymerization reactor control  

SciTech Connect (OSTI)

The principal difficulties in achieving good control of polymerization reactors are related to inadequate on-line measurement, a lack of understanding of the dynamics of the process, the highly sensitive and nonlinear behavior of these reactors, and the lack of well-developed techniques for the control of nonlinear processes. Some illustrations of these problems and a discussion of potential techniques for overcoming some of these difficulties is provided.

Ray, W.H.

1985-01-01T23:59:59.000Z

295

Molten metal reactors  

DOE Patents [OSTI]

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

296

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

297

Fuel and cladding nano-technologies based solutions for long life heat-pipe based reactors  

SciTech Connect (OSTI)

A novel nuclear reactor concept, unifying the fuel pipe with fuel tube functionality has been developed. The structure is a quasi-spherical modular reactor, designed for a very long life. The reactor module unifies the fuel tube with the heat pipe and a graphite beryllium reflector. It also uses a micro-hetero-structure that allows the fission products to be removed in the heat pipe flow and deposited in a getter area in the cold zone of the heat pipe, but outside the neutron flux. The reactor operates as a breed and burn reactor - it contains the fuel pipe with a variable enrichment, starting from the hot-end of the pipe, meant to assure the initial criticality, and reactor start-up followed by area with depleted uranium or thorium that get enriched during the consumption of the first part of the enriched uranium. (authors)

Popa-Simil, L. [LAVM LLC, Los Alamos (United States)

2012-07-01T23:59:59.000Z

298

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. 10 figures.

Grossman, M.W.

1991-04-30T23:59:59.000Z

299

Reactor Safety Research Programs  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Dotson, CW

1980-08-01T23:59:59.000Z

300

REACTOR GROUT THERMAL PROPERTIES  

SciTech Connect (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

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

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

302

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

SciTech Connect (OSTI)

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

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

1985-01-01T23:59:59.000Z

303

Pressurized water nuclear reactor system with hot leg vortex mitigator  

DOE Patents [OSTI]

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

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

1990-01-01T23:59:59.000Z

304

Methanation assembly using multiple reactors  

DOE Patents [OSTI]

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

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24T23:59:59.000Z

305

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - September 2006PhotovoltaicSeptember 22,Reactor Decommissioning Click here to view

306

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...  

Office of Scientific and Technical Information (OSTI)

A reactor and associated power plant designed to produce 1.05 Mwh and 3.535 Mwh of steam for heating purposes are described. The total thermal output of the reactor is 10 Mwh....

307

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

308

Inexpensive Mini Thermonuclear Reactor  

E-Print Network [OSTI]

This proposed design for a mini thermonuclear reactor uses a method based upon a series of important innovations. A cumulative explosion presses a capsule with nuclear fuel up to 100 thousands of atmospheres, the explosive electric generator heats the capsule/pellet up to 100 million degrees and a special capsule and a special cover which keeps these pressure and temperature in capsule up to 0.001 sec. which is sufficient for Lawson criteria for ignition of thermonuclear fuel. Major advantages of these reactors/bombs is its very low cost, dimension, weight and easy production, which does not require a complex industry. The mini thermonuclear bomb can be delivered as a shell by conventional gun (from 155 mm), small civil aircraft, boat or even by an individual. The same method may be used for thermonuclear engine for electric energy plants, ships, aircrafts, tracks and rockets. Key words: Thermonuclear mini bomb, thermonuclear reactor, nuclear energy, nuclear engine,

Alexander Bolonkin; Alexander Bolonkin

309

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

310

Fusion reactor control  

SciTech Connect (OSTI)

The plasma kinetic temperature and density changes, each per an injected fuel density rate increment, control the energy supplied by a thermonuclear fusion reactor in a power production cycle. This could include simultaneously coupled control objectives for plasma current, horizontal and vertical position, shape and burn control. The minimum number of measurements required, use of indirect (not plasma parameters) system measurements, and distributed control procedures for burn control are to be verifiable in a time dependent systems code. The International Thermonuclear Experimental Reactor (ITER) has the need to feedback control both the fusion output power and the driven plasma current, while avoiding damage to diverter plates. The system engineering of fusion reactors must be performed to assure their development expeditiously and effectively by considering reliability, availability, maintainability, environmental impact, health and safety, and cost.

Plummer, D.A.

1995-12-31T23:59:59.000Z

311

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

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

Schock, Alfred

1994-08-01T23:59:59.000Z

312

Reactor for exothermic reactions  

DOE Patents [OSTI]

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

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

1993-01-01T23:59:59.000Z

313

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

314

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

315

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

316

Preliminary design study of compressed-air energy storage in a salt dome. Volume 6. CAES plant design. Final report  

SciTech Connect (OSTI)

The preliminary plant design for a compressed-air energy storage (CAES) plant located in the Middle South Services, Inc. (MSS), is presented. The design is based upon the facility criteria; the specific site; and the systems, subsystems, and components identified in the other task of this study. The proposed 220-MW(e) plant is located at the Carmichael salt dome near Jackson, Mississippi. The compressed air is stored in two solution-mined caverns in the salt dome. The plant area, exclusive of the remote fuel unloading facility, occupies 20 acres. An equipment list, a plot plan, and general arrangement drawings define the CAES plant. The details concerning the major equipment and the operation of the mechanical systems are described. The capital investment cost (exclusive of owner's cost) of the 220-MW(e) CAES plant is $85.6 million in 1979 dollars or $389/kW. This cost is based on firing the turbines with No. 2 fuel oil. As an alternative, the capital investment cost under the same conditions for a plant firing No. 6 oil is $90.9 million or $413/kW. The project schedule from start of licensing to commercial operation is estimated to be 70 months, with actual construction (including dewatering of the caverns) estimated to be 39 months. Based on the cost estimate developed in this task and the modified financial data and fuel cost projections, the economic introduction of CAES into the MSS system was examined for the No. 2 oil-fired plant. Due to lack of funds, the economic analysis did not extend beyond the year 1988. No system analysis of the No. 6 oil-fired plant was made. The economic introduction of CAES in the MSS system before 1990 is unlikely because the older oil-fired units in the MSS system may be economically used for cycling and peaking, if required. For a system with a different composition of generating units (i.e., low-cost, coal-fired plants), CAES may be economical at an earlier date.

Not Available

1982-04-01T23:59:59.000Z

317

Support arrangements for core modules of nuclear reactors. [PWR  

DOE Patents [OSTI]

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

Bollinger, L.R.

1983-11-03T23:59:59.000Z

318

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

319

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

320

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. 8 figs.

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

1999-08-10T23:59:59.000Z

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.


321

Diagnostics for hybrid reactors  

SciTech Connect (OSTI)

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

322

Perspectives on reactor safety  

SciTech Connect (OSTI)

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

323

Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor  

SciTech Connect (OSTI)

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

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

2009-10-01T23:59:59.000Z

324

Neutronic and thermal design considerations for heat-pipe reactors  

SciTech Connect (OSTI)

SABRE (Space-Arena Baseline Reactor) is a 100-kW/sub e/, heat-pipe-cooled, beryllium-reflected, fast reactor that produces heat at a temperature of 1500/sup 0/K and radiatively transmits it to high-temperature thermoelectric (TE) conversion elements. The use of heat pipes for core heat removal eliminates single-point failure mechanisms in the reactor cooling system, and provides minimal temperature drop radiative coupling to the TE array, as well as automatic, self-actuating removal of reactor afterheat. The question of how the failure of a fuel module heat pipe will affect neighboring fuel modules in the core is discussed, as is fission density peaking that occurs at the core/reflector interface. Results of neutronic calculations of the control margin available are described. Another issue that is addressed is that of helium generation in the heat pipes from neutron reactions in the core with the heat pipe fluid. Finally, the growth potential of the SABRE design to much higher powers is examined.

Ranken, W.A.; Koenig, D.R.

1983-01-01T23:59:59.000Z

325

Innovative design of uranium startup fast reactors  

E-Print Network [OSTI]

Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

Fei, Tingzhou

2012-01-01T23:59:59.000Z

326

Reactor operation environmental information document  

SciTech Connect (OSTI)

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

1989-12-01T23:59:59.000Z

327

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

Broader source: Energy.gov [DOE]

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

328

Fuel Summary Report: Shippingport Light Water Breeder Reactor  

SciTech Connect (OSTI)

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

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

1999-01-01T23:59:59.000Z

329

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

SciTech Connect (OSTI)

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

Scheele, Randall D.; Casella, Andrew M.

2010-09-28T23:59:59.000Z

330

Reactor operation safety information document  

SciTech Connect (OSTI)

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

Not Available

1990-01-01T23:59:59.000Z

331

Reed Reactor Facility Annual Report  

SciTech Connect (OSTI)

This is the report of the operations, experiments, modifications, and other aspects of the Reed Reactor Facility for the year.

Frantz, Stephen G.

2000-09-01T23:59:59.000Z

332

Thermal Reactor Safety  

SciTech Connect (OSTI)

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

Not Available

1980-06-01T23:59:59.000Z

333

Nuclear reactor building  

DOE Patents [OSTI]

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

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

1994-01-01T23:59:59.000Z

334

Nuclear reactor building  

DOE Patents [OSTI]

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

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

1994-04-05T23:59:59.000Z

335

NETL - Chemical Looping Reactor  

ScienceCinema (OSTI)

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

None

2014-06-26T23:59:59.000Z

336

Nuclear Reactors and Technology  

SciTech Connect (OSTI)

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

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

1992-01-01T23:59:59.000Z

337

Fossil fuel furnace reactor  

DOE Patents [OSTI]

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

Parkinson, William J. (Los Alamos, NM)

1987-01-01T23:59:59.000Z

338

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

SciTech Connect (OSTI)

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

339

Mo-99 production at the Annular Core Research Reactor - recent calculative results  

SciTech Connect (OSTI)

Significant progress has been made over the past year in understanding the chemistry and processing challenges associated with {sup 99}Mo production using Cintichem type targets. Targets fabricated at Los Alamos National Laboratory have been successfully irradiated in fuel element locations at the Annular Core Research Reactor (ACRR) and processed at the Sandia Hot Cell Facility. The next goal for the project is to remove the central cavity experiment tube from the reactor core, allowing for the irradiation of up to 37 targets. After the in-core work is complete, the reactor will be capable of producing significant quantities of {sup 99}Mo.

Parma, E.J.

1997-11-01T23:59:59.000Z

340

Development of ceramic membrane reactors for high temperature gas cleanup. Final report  

SciTech Connect (OSTI)

The objective of this project was to develop high temperature, high pressure catalytic ceramic membrane reactors and to demonstrate the feasibility of using these membrane reactors to control gaseous contaminants (hydrogen sulfide and ammonia) in integrated gasification combined cycle (IGCC) systems. Our strategy was to first develop catalysts and membranes suitable for the IGCC application and then combine these two components as a complete membrane reactor system. We also developed a computer model of the membrane reactor and used it, along with experimental data, to perform an economic analysis of the IGCC application. Our results have demonstrated the concept of using a membrane reactor to remove trace contaminants from an IGCC process. Experiments showed that NH{sub 3} decomposition efficiencies of 95% can be achieved. Our economic evaluation predicts ammonia decomposition costs of less than 1% of the total cost of electricity; improved membranes would give even higher conversions and lower costs.

Roberts, D.L.; Abraham, I.C.; Blum, Y.; Gottschlich, D.E.; Hirschon, A.; Way, J.D.; Collins, J.

1993-06-01T23:59:59.000Z

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.


341

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

SciTech Connect (OSTI)

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

Schilperoort, Daryl L.; Faulk, Darrin [Washington Closure Hanford, 2620 Fermi Avenue, Richland, Washington 99352 (United States)] [Washington Closure Hanford, 2620 Fermi Avenue, Richland, Washington 99352 (United States)

2013-07-01T23:59:59.000Z

342

Reactor vessel support system. [LMFBR  

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

1980-05-09T23:59:59.000Z

343

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

344

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

1998-01-01T23:59:59.000Z

345

Mexico HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

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

346

Arsenic removal and stabilization by synthesized pyrite  

E-Print Network [OSTI]

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...

Song, Jin Kun

2009-05-15T23:59:59.000Z

347

Libya HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Libya HEU Removal Libya HEU Removal Location Libya United States 27 34' 9.5448" N, 17 24' 8.4384" E See map: Google Maps Javascript is required to view this map....

348

Canada HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Canada HEU Removal Canada HEU Removal Location Canada United States 53 47' 24.972" N, 104 35' 23.4384" W See map: Google Maps Javascript is required to view this map....

349

Israel HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Israel HEU Removal Israel HEU Removal Location Israel United States 30 53' 18.2328" N, 34 52' 14.178" E See map: Google Maps Javascript is required to view this map....

350

Turkey HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Turkey HEU Removal Turkey HEU Removal Location Turkey United States 38 26' 50.2044" N, 40 15' 14.0616" E See map: Google Maps Javascript is required to view this map...

351

Uzbekistan HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Uzbekistan HEU Removal Uzbekistan HEU Removal Location Uzbekistan United States 42 6' 56.196" N, 63 22' 8.9076" E See map: Google Maps Javascript is required to view this map...

352

France HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Four-Year Plan France HEU Removal France HEU Removal Location United States 45 44' 20.0544" N, 2 17' 6.5616" E See map: Google Maps Javascript is required to view this map...

353

Kazakhstan HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

HEU Removal Kazakhstan HEU Removal Location Kazakhstan United States 48 59' 44.1492" N, 67 3' 37.9692" E See map: Google Maps Javascript is required to view this map....

354

Ukraine HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Ukraine HEU Removal Ukraine HEU Removal Location Ukraine United States 50 12' 24.8688" N, 25 50' 23.4384" E See map: Google Maps Javascript is required to view this map...

355

Chile HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Four-Year Plan Chile HEU Removal Chile HEU Removal Location United States 25 28' 1.4916" S, 69 33' 55.548" W See map: Google Maps Javascript is required to view this map...

356

Taiwan HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Taiwan HEU Removal Taiwan HEU Removal Location Taiwan United States 24 35' 37.4964" N, 120 53' 36.798" E See map: Google Maps Javascript is required to view this map....

357

Romania HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Romania HEU Removal Romania HEU Removal Location Romania United States 45 47' 1.932" N, 24 41' 50.1576" E See map: Google Maps Javascript is required to view this map....

358

Hungary HEU removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Hungary HEU removal Hungary HEU removal Location Hungary United States 47 11' 51.6336" N, 19 41' 15" E See map: Google Maps Javascript is required to view this map....

359

Serbia HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Serbia HEU Removal Serbia HEU Removal Location Serbia United States 44 22' 45.7068" N, 20 26' 4.452" E See map: Google Maps Javascript is required to view this map....

360

Japan HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Japan HEU Removal Japan HEU Removal Location Japan United States 37 36' 59.5872" N, 140 5' 51.5616" E See map: Google Maps Javascript is required to view this map....

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.


361

Poland HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Poland HEU Removal Poland HEU Removal Location Poland United States 53 23' 50.2872" N, 17 50' 30.4692" E See map: Google Maps Javascript is required to view this map....

362

Italy HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Italy HEU Removal Italy HEU Removal Location Italy United States 43 41' 3.4548" N, 11 28' 11.0172" E See map: Google Maps Javascript is required to view this map...

363

Vietnam HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Plan Vietnam HEU Removal Vietnam HEU Removal Location Vietnam United States 13 12' 30.8628" N, 108 19' 30.702" E See map: Google Maps Javascript is required to view this map....

364

Nuclear reactor construction with bottom supported reactor vessel  

DOE Patents [OSTI]

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

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

1987-01-01T23:59:59.000Z

365

Water Distribution and Removal Model  

SciTech Connect (OSTI)

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

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

2005-08-26T23:59:59.000Z

366

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

SciTech Connect (OSTI)

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

Volkov, Victor; Danilovich, Alexey; Zverkov, Yuri; Ivanov, Oleg; Kolyadin, Vyacheslav; Lemus, Alexey; Pavlenko, Vitaly; Semenov, Sergey; Fadin, Sergey; Shisha, Anatoly; Chesnokov, Alexander [National Research Centre 'Kurchatov Institute', Moscow (Russian Federation)] [National Research Centre 'Kurchatov Institute', Moscow (Russian Federation)

2013-07-01T23:59:59.000Z

367

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

DOE Patents [OSTI]

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

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

1983-11-25T23:59:59.000Z

368

Removing Stains from Washable Fabrics.  

E-Print Network [OSTI]

Page Numbers Stain Page Numbers Acne medicine Blueberry Special 9 Wet 8 Adhesive tape Dye 8 Special 9 Butter Alcoholic beverages Dry 8 Wet 8 Oil 8 Tannin 8 Calamine lotion Asphalt Combination 8 Combination 8 Dye 8 Dye 8 Candle wax Automotive... the most gentle to the most harsh, so always stop treatments as soon as the stain has been removed. Dry Type Stains Dissolve the stain with a grease solvent. Lubricate the stain with dry spotter, coconut oil or mineral oil (sold in health food...

Beard, Ann Vanderpoorten

1988-01-01T23:59:59.000Z

369

Automatic Eyeglasses Removal from Face Images  

E-Print Network [OSTI]

Automatic Eyeglasses Removal from Face Images Chenyu Wu, Ce Liu, Heung-Yueng Shum, Member, IEEE an intelligent image editing and face synthesis system that automatically removes eyeglasses from an input frontal face image. Although conventional image editing tools can be used to remove eyeglasses by pixel

Narasayya, Vivek

370

Spherical torus fusion reactor  

DOE Patents [OSTI]

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

Peng, Yueng-Kay M. (Oak Ridge, TN)

1989-01-01T23:59:59.000Z

371

Nuclear reactor safety device  

DOE Patents [OSTI]

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

Hutter, E.

1983-08-15T23:59:59.000Z

372

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. 10 figs.

Jassby, D.L.

1987-09-04T23:59:59.000Z

373

Fluid sampling system for a nuclear reactor  

DOE Patents [OSTI]

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

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

1994-11-22T23:59:59.000Z

374

Fluid sampling system for a nuclear reactor  

DOE Patents [OSTI]

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

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

1994-01-01T23:59:59.000Z

375

Preliminary design study of compressed-air energy storage in a salt dome. Volume 1: executive summary. Final report  

SciTech Connect (OSTI)

The preliminary design and cost estimate of a compressed air energy storage (CAES) plant located in the Middle South Utilities (MSU) system are summarized in this report. The 220 MWe CAES plant which stores air in two solution mined salt caverns, is located at the Carmichael salt dome near Jackson, Mississippi. The facility criteria, site selection and the turbomachinery and auxiliaries, are briefly described together with an outline of the proposed procedure for developing the caverns. Using this information and data, the preliminary CAES plant design was prepared; also the capital cost estimate, cash flow and project schedule were developed. The Environmental Assessment did not reveal any major site impediments to the construction of the plant. However, it is believed that an EIS is required primarily because CAES is a new technology without precedent in the United States. Although a final system planning study was not completed because of lack of funds, from preliminary analysis a CAES plant does not appear to be economic in the MSU system before the mid 1990s. This is due to the unique features of the MSU system. For other systems under more favorable conditions, CAES may be economic at an earlier date. The construction of a CAES plant with salt cavern air storage may by considered ready for use as a commercial electric generating plant. The experience at the Huntorf plant in West Germany demonstrates the technical feasibility of the CAES concept. Certain details of the plant defined in this study are different from the Huntorf plant. Design verification by limited testing and analysis would provide added confidence to those considering a CAES plant.

Not Available

1982-01-01T23:59:59.000Z

376

Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report  

SciTech Connect (OSTI)

This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

1994-02-01T23:59:59.000Z

377

Process and system for removing impurities from a gas  

DOE Patents [OSTI]

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

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

2014-04-15T23:59:59.000Z

378

Method of making thermally removable epoxies  

DOE Patents [OSTI]

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

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

2002-01-01T23:59:59.000Z

379

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

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

Schock, Alfred

1994-06-01T23:59:59.000Z

380

Nuclear reactor cooling system decontamination reagent regeneration  

DOE Patents [OSTI]

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

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

1985-01-01T23:59:59.000Z

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.


381

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two  

SciTech Connect (OSTI)

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future.

Glasstone, S.; Sesonske, A.

1994-12-31T23:59:59.000Z

382

DOE reactor-pumped laser program  

SciTech Connect (OSTI)

FALCON is a high-power, steady-state, nuclear reactor-pumped laser (RPL) concept that is being developed by the Department of Energy. The FALCON program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 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. 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. The compact size and self-contained power also makes an RPL very suitable for ship basing so that power-beaming activities could be situated around the globe. 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, wide-area deposition of diamond-like coatings, and 3-D ceramic lithography.

Felty, J.R. [USDOE, Germantown, MD (United States). Defense Programs; Lipinski, R.J.; McArthur, D.A.; Pickard, P.S. [Sandia National Labs., Albuquerque, NM (United States)

1993-12-31T23:59:59.000Z

383

BROOKHAVEN NATIONAL LABORATORY'S HIGH FLUX BEAM REACTOR  

E-Print Network [OSTI]

irradiation experiments with a variety of neutron energy spectra for neutron activation analysis, producing HISTORY #12;2 Erecting the dome in 1962. The building was designed as "confinement" to prevent leakage of radioactive materials into the environment. The interior of the building is maintained at a pressure slightly

Ohta, Shigemi

384

Designing Reactors to Facilitate Decommissioning  

SciTech Connect (OSTI)

Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

Richard H. Meservey

2006-06-01T23:59:59.000Z

385

Melter Glass Removal and Dismantlement  

SciTech Connect (OSTI)

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

Richardson, BS

2000-10-31T23:59:59.000Z

386

Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant  

SciTech Connect (OSTI)

Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J. [Serco, Rutherford House, Quedgeley, Gloucester, Gl2 4NF (United Kingdom); Seren, T.; Lipponen, M.; Kekki, T. [VTT, Technical Research Centre of Finland, Otakaari 3 K, P.O. BOX 1000, Espoo, FI-02044 (Finland)

2011-07-01T23:59:59.000Z

387

Engineering problems of tandem-mirror reactors  

SciTech Connect (OSTI)

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

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

1981-10-22T23:59:59.000Z

388

Alternate particle removal technologies for the Airborne Activity Confinement System at the Savannah River Site  

SciTech Connect (OSTI)

This report presents a review of the filtration technologies available for the removal of particulate material from a gas stream. It was undertaken to identify alternate filtration technologies that may be employed in the Airborne Activity Confinement System (AACS) at the Savannah River Plant. This report is organized into six sections: (1) a discussion of the aerosol source term and its definition, (2) a short discussion of particle and gaseous contaminant removal mechanisms, (3) a brief overview of particle removal technologies, (4) a discussion of the existing AACS and its potential shortcomings, (5) an enumeration of issues to be addressed in upgrading the AACS, and, (6) a detailed discussion of the identified technologies. The purpose of this report is to identity available options to the existing particle removal system. This system is in continuous operation during routine operation of the reactor. As will be seen, there are a number of options and the selection of any technology or combination of technologies will depend on the design aerosol source term (yet to be appropriately defined) as well as the flow requirements and configuration. This report does not select a specific technology. It focuses on particulate removal and qualitatively on the removal of radio-iodine and mist elimination. Candidate technologies have been selected from industrial and nuclear gas cleaning applications.

Brockmann, J.E.; Adkins, C.L.J.; Gelbard, F. [Sandia National Labs., Albuquerque, NM (United States)

1991-09-01T23:59:59.000Z

389

Neutrino Oscillation Studies with Reactors  

E-Print Network [OSTI]

Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

Petr Vogel; Liangjian Wen; Chao Zhang

2015-03-03T23:59:59.000Z

390

Neutrino Oscillation Studies with Reactors  

E-Print Network [OSTI]

Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

Vogel, Petr; Zhang, Chao

2015-01-01T23:59:59.000Z

391

Thermonuclear Reflect AB-Reactor  

E-Print Network [OSTI]

The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical project. The main point of preference for the offered reactor is its likely cheapness as a power source. Key words: Micro-thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, Self-magnetic AB-thermonuclear reactor, aerospace thermonuclear engine.

Alexander Bolonkin

2008-03-26T23:59:59.000Z

392

Light Water Reactor Sustainability Newsletter  

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

hydraulics software RELAP-7 (which is under development in the Light Water Reactor Sustainability LWRS Program). A novel interaction between the probabilistic part (i.e., RAVEN)...

393

Light Water Reactor Sustainability Newsletter  

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

30-35, August 2012. Clayton, D. A. and M. S. Hileman, 2012, Light Water Reactor Sustainability Non-Destructive Evaluation for Concrete Research and Development Roadmap, ORNLTM-...

394

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

SciTech Connect (OSTI)

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

Kelleher, B. C.; Sellers, S. R.; Anderson, M. H.; Sridharan, K.; Scheele, R. D. [Dept. of Engineering Physics, Univ.of Wisconsin - Madison, 1500 Engineering Drive, Madison, WI 53706 (United States)

2012-07-01T23:59:59.000Z

395

Reactor coolant pump flywheel  

DOE Patents [OSTI]

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

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

2013-11-26T23:59:59.000Z

396

Nuclear reactor control assembly  

SciTech Connect (OSTI)

This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other.

Negron, S.B.

1991-06-11T23:59:59.000Z

397

Nuclear reactor control apparatus  

SciTech Connect (OSTI)

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

Sridhar, B.N.

1981-08-28T23:59:59.000Z

398

Nuclear reactor control apparatus  

DOE Patents [OSTI]

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

Sridhar, Bettadapur N. (Cupertino, CA)

1983-11-01T23:59:59.000Z

399

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect (OSTI)

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

Not Available

1991-04-01T23:59:59.000Z

400

Modular Pebble Bed Reactor High Temperature Gas Reactor  

E-Print Network [OSTI]

Modular Pebble Bed Reactor High Temperature Gas Reactor Andrew C Kadak Massachusetts Institute For 1150 MW Combined Heat and Power Station Oil Refinery Hydrogen Production Desalinization Plant VHTR/Graphite Discrimination system Damaged Sphere ContainerGraphiteReturn FuelReturn Fresh Fuel Container Spent Fuel Tank #12

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

Fast Reactor Fuel Type and Reactor Safety Performance  

SciTech Connect (OSTI)

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

R. Wigeland; J. Cahalan

2009-09-01T23:59:59.000Z

402

Westinghouse Small Modular Reactor nuclear steam supply system design  

SciTech Connect (OSTI)

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)

Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

2012-07-01T23:59:59.000Z

403

Removal of metal ions from aqueous solution  

DOE Patents [OSTI]

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

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

1990-11-13T23:59:59.000Z

404

Removal of metal ions from aqueous solution  

DOE Patents [OSTI]

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

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

1990-01-01T23:59:59.000Z

405

Gaseous reactor control system  

SciTech Connect (OSTI)

This paper describes a nuclear reactor control system for controlling the reactivity of the core of a nuclear reactor. It includes a control gas having a high neutron cross-section; a first tank containing a first supply of the control gas; a first conduit providing a first fluid passage extending into the core, the first conduit being operatively connected to communicate with the first tank; a first valve operatively connected to regulate the flow of the control gas between the first tank and the first conduit; a second conduit concentrically disposed around the first conduit such that a second fluid passage is defined between the outer surface of the first conduit and the inner surface of the second conduit; a second tank containing a second supply of the control gas, the second tank being operatively connected to communicate with the second fluid passage; a second supply valve operatively connected to regulate the flow of the control gas between the second tank and the second fluid passage.

Abdel-Khalik, S.

1991-09-03T23:59:59.000Z

406

Overview of the US stellarator reactor study  

SciTech Connect (OSTI)

This study, which uses a cost-minimization code that incorporates the ARIES costing and reactor component models with a I-D energy transport calculation, shows that a torsatron reactor could be competitive with a tokamak reactor.

Lyon, J.F. [Oak Ridge National Lab., TN (United States); Gulec, K. [Univ. of Tennessee, Knoxville, TN (United States); Miller, R.L. [Los Alamos National Lab., NM (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison, WI (United States)

1993-12-31T23:59:59.000Z

407

Process for particulate removal from coal liquids  

DOE Patents [OSTI]

Suspended solid particulates are removed from liquefied coal products by first subjecting such products to hydroclone action for removal in the underflow of the larger size particulates, and then subjecting the overflow from said hydroclone action, comprising the residual finer particulates, to an electrostatic field in an electrofilter wherein such finer particulates are deposited in the bed of beads of dielectric material on said filter. The beads are periodically cleaned by backwashing to remove the accumulated solids.

Rappe, Gerald C. (Macungie, PA)

1983-01-01T23:59:59.000Z

408

Method for removing contaminants from plastic resin  

DOE Patents [OSTI]

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

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

2008-12-09T23:59:59.000Z

409

Reactor Cost Analysis Brian James  

E-Print Network [OSTI]

Reactor Cost Analysis Brian James Directed Technologies, Inc. 6-7 November 2007 This presentation specification & optimization · Capital cost estimation · Projected hydrogen $/kg #12;Directed Technologies, Inc/WGS Membrane Reactor OTM/ Water-Splitting ANL With WGS #12;Directed Technologies, Inc. 6-7 November 2007 BILIWG

410

Method of making thermally removable polymeric encapsulants  

DOE Patents [OSTI]

A method of making a thermally-removable encapsulant by heating a mixture of at least one bis(maleimide) compound and at least one monomeric tris(furan) or tetrakis(furan) compound at temperatures from above room temperature to less than approximately 90.degree. C. to form a gel and cooling the gel to form the thermally-removable encapsulant. The encapsulant can be easily removed within approximately an hour by heating to temperatures greater than approximately 90.degree. C., preferably in a polar solvent. The encapsulant can be used in protecting electronic components that may require subsequent removal of the encapsulant for component repair, modification or quality control.

Small, James H. (Santa Fe, NM); Loy, Douglas A. (Albuquerque, NM); Wheeler, David R. (Albuquerque, NM); McElhanon, James R. (Albuquerque, NM); Saunders, Randall S. (late of Albuquerque, NM)

2001-01-01T23:59:59.000Z

411

Australia HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

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

412

Argentina HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

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

413

System for removing contaminants from plastic resin  

DOE Patents [OSTI]

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

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

2010-11-23T23:59:59.000Z

414

Sweden Plutonium Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

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

415

Solvent refined coal reactor quench system  

DOE Patents [OSTI]

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

Thorogood, Robert M. (Macungie, PA)

1983-01-01T23:59:59.000Z

416

Solvent refined coal reactor quench system  

DOE Patents [OSTI]

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

Thorogood, R.M.

1983-11-08T23:59:59.000Z

417

Plutonium Consumption Program, CANDU Reactor Project final report  

SciTech Connect (OSTI)

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

Not Available

1994-07-31T23:59:59.000Z

418

Fast reactors and nuclear nonproliferation  

SciTech Connect (OSTI)

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

Avrorin, E.N. [Russian Federal Nuclear Center - Zababakhin Institute of Applied Physics, Snezhinsk (Russian Federation); Rachkov, V.I.; Chebeskov, A.N. [State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering, Bondarenko Square, 1, Obninsk, Kaluga region, 249033 (Russian Federation)

2013-07-01T23:59:59.000Z

419

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

DOE Patents [OSTI]

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

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

1994-05-03T23:59:59.000Z

420

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

DOE Patents [OSTI]

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

McDermott, Daniel J. (Export, PA); Schrader, Kenneth J. (Penn Hills, PA); Schulz, Terry L. (Murrysville Boro, PA)

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.


421

Retrospective dosimetry analyses of reactor vessel cladding samples  

SciTech Connect (OSTI)

Reactor pressure vessel cladding samples for Ringhals Units 3 and 4 in Sweden were analyzed using retrospective reactor dosimetry techniques. The objective was to provide the best estimates of the neutron fluence for comparison with neutron transport calculations. A total of 51 stainless steel samples consisting of chips weighing approximately 100 to 200 mg were removed from selected locations around the pressure vessel and were sent to Pacific Northwest National Laboratory for analysis. The samples were fully characterized and analyzed for radioactive isotopes, with special interest in the presence of Nb-93m. The RPV cladding retrospective dosimetry results will be combined with a re-evaluation of the surveillance capsule dosimetry and with ex-vessel neutron dosimetry results to form a comprehensive 3D comparison of measurements to calculations performed with 3D deterministic transport code. (authors)

Greenwood, L. R.; Soderquist, C. Z. [Battelle Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Fero, A. H. [Westinghouse Electric Company, Cranberry Twp., PA 16066 (United States)

2011-07-01T23:59:59.000Z

422

Cooling molten salt reactors using “gas-lift”  

SciTech Connect (OSTI)

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

Zitek, Pavel, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz; Valenta, Vaclav, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz; Klimko, Marek, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz [University of West Bohemia in Pilsen, Univerzitní 8, 306 14 Pilsen (Czech Republic)

2014-08-06T23:59:59.000Z

423

Hydrothermal Processing of Macroalgal Feedstocks in Continuous-Flow Reactors  

SciTech Connect (OSTI)

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

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

2014-02-18T23:59:59.000Z

424

MOOSE simulating nuclear reactor CRUD buildup  

SciTech Connect (OSTI)

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

None

2014-02-06T23:59:59.000Z

425

THE MATERIALS OF FAST BREEDER REACTORS  

E-Print Network [OSTI]

metal fast breeder reactor (LMFBR) concern the behavior ofmetal fast breeder reactor (LMFBR). Despite the simplicityinduced by irradiation. LMFBR funding is the largest single

Olander, Donald R.

2013-01-01T23:59:59.000Z

426

MOOSE simulating nuclear reactor CRUD buildup  

ScienceCinema (OSTI)

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

None

2014-07-21T23:59:59.000Z

427

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION  

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

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1...

428

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network [OSTI]

gas reactors with the effective heat transfer of a molten salt coolant and the passive natural circulation safety systems of sodium fast reactors.

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

429

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network [OSTI]

pebble bed reactor,” Nuclear Engineering and Design, vol.the AVR reactor,” Nuclear Engineering and Design, vol. 121,Operating Experience,” Nuclear Engineering and Design, vol.

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

430

Automatic reactor power control for a pressurized water reactor  

SciTech Connect (OSTI)

An automatic reactor power control system is presented for a pressurized water reactor (PWR). The associated reactor control strategy is called mode K.' The new system implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial shape change, which allows automatic control of the axial power distribution. Thus, the mode K enables precise regulation of both the reactivity and the power distribution, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load-follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1,000-MW (electric) PWR. The simulation results illustrate that the mode K would be a practical reactor power control strategy for the increased automation of nuclear plants.

Jungin Choi (Kyungwon Univ. (Korea, Republic of)); Yungjoon Hah (Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)); Unchul Lee (Seoul National Univ. (Korea, Republic of))

1993-05-01T23:59:59.000Z

431

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

SciTech Connect (OSTI)

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

Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)] [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

2013-07-01T23:59:59.000Z

432

Nuclear reactor control apparatus  

DOE Patents [OSTI]

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

Sridhar, Bettadapur N. (Cupertino, CA)

1983-10-25T23:59:59.000Z

433

Nuclear reactor control  

DOE Patents [OSTI]

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

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

1982-01-01T23:59:59.000Z

434

Nuclear reactor control  

SciTech Connect (OSTI)

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

Cawley, W.E.; Warnick, R.F.

1982-03-30T23:59:59.000Z

435

Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One  

SciTech Connect (OSTI)

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials.

Glasstone, S.; Sesonske, A.

1994-12-31T23:59:59.000Z

436

University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor  

SciTech Connect (OSTI)

The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

Eric C. Woolstenhulme; Dana M. Hewit

2008-09-01T23:59:59.000Z

437

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect (OSTI)

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

Not Available

1993-11-01T23:59:59.000Z

438

Method And Apparatus For Arbitrarily Large Capacity Removable Media  

DOE Patents [OSTI]

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

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

2003-04-08T23:59:59.000Z

439

Development of Novel Water-Gas Shift Membrane Reactor  

SciTech Connect (OSTI)

This report summarizes the objectives, technical barrier, approach, and accomplishments for the development of a novel water-gas-shift (WGS) membrane reactor for hydrogen enhancement and CO reduction. We have synthesized novel CO{sub 2}-selective membranes with high CO{sub 2} permeabilities and high CO{sub 2}/H{sub 2} and CO{sub 2}/CO selectivities by incorporating amino groups in polymer networks. We have also developed a one-dimensional non-isothermal model for the countercurrent WGS membrane reactor. The modeling results have shown that H{sub 2} enhancement (>99.6% H{sub 2} for the steam reforming of methane and >54% H{sub 2} for the autothermal reforming of gasoline with air on a dry basis) via CO{sub 2} removal and CO reduction to 10 ppm or lower are achievable for synthesis gases. With this model, we have elucidated the effects of system parameters, including CO{sub 2}/H{sub 2} selectivity, CO{sub 2} permeability, sweep/feed flow rate ratio, feed temperature, sweep temperature, feed pressure, catalyst activity, and feed CO concentration, on the membrane reactor performance. Based on the modeling study using the membrane data obtained, we showed the feasibility of achieving H{sub 2} enhancement via CO{sub 2} removal, CO reduction to {le} 10 ppm, and high H{sub 2} recovery. Using the membrane synthesized, we have obtained <10 ppm CO in the H{sub 2} product in WGS membrane reactor experiments. From the experiments, we verified the model developed. In addition, we removed CO{sub 2} from a syngas containing 17% CO{sub 2} to about 30 ppm. The CO{sub 2} removal data agreed well with the model developed. The syngas with about 0.1% CO{sub 2} and 1% CO was processed to convert the carbon oxides to methane via methanation to obtain <5 ppm CO in the H{sub 2} product.

Ho, W. S. Winston

2004-12-29T23:59:59.000Z

440

Method of removing contaminants from plastic resins  

DOE Patents [OSTI]

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

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

2008-11-18T23:59:59.000Z

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

Method of removing contaminants from plastic resins  

DOE Patents [OSTI]

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

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

2007-08-07T23:59:59.000Z

442

Method for removing contaminants from plastic resin  

DOE Patents [OSTI]

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

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

2008-12-30T23:59:59.000Z

443

On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks  

SciTech Connect (OSTI)

This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

Samuel Bays; Ayodeji Alajo

2010-05-01T23:59:59.000Z

444

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

SciTech Connect (OSTI)

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

Prechtl, E.; Suessdorf, W.

2003-02-27T23:59:59.000Z

445

Immobilization of Cesium Traps from the BN-350 Fast Reactor (Aktau, Kazakhstan)  

SciTech Connect (OSTI)

During BN-350 reactor operations and also during the initial stages of decommissioning, cesium traps were used to decontaminate the reactor’s primary sodium coolant. Two different types of carbon-based trap were used – the MAVR series, low ash granulated graphite adsorber (LAG) contained in a carrier designed to be inserted into the reactor core during shutdown; and a series of ex-reactor trap accumulators(TAs) which used reticulated vitreous carbon (RVC) to reduce Cs-137 levels in the sodium after final reactor shutdown. In total four MAVRs and seven TAs were used at BN-350 to remove an estimated cumulative 755 TBq of cesium. The traps, which also contain residual sodium, need to be immobilized in an appropriate way to allow them to be consigned as waste packages for long term storage and, ultimately, disposal. The present paper reports on the current status of the implementation phase, with particular reference to the work done to date on the trap accumulators, which have the most similarity with the cesium traps used at other reactors.

J. A. Michelbacher; C. Knight; O. G. Romanenko; I. L. Tazhibaeva; I. L. Yakovlev; A. V. Rovneyko; V. I. Maev; D. Wells; A. Herrick

2011-03-01T23:59:59.000Z

446

Heat pipe cooled reactors for multi-kilowatt space power supplies  

SciTech Connect (OSTI)

Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFEs) to radiator heat pipes.

Ranken, W.A.; Houts, M.G.

1995-01-01T23:59:59.000Z

447

Membrane contactor/separator for an advanced ozone membrane reactor for treatment of recalcitrant organic pollutants in water  

SciTech Connect (OSTI)

An advanced ozone membrane reactor that synergistically combines membrane distributor for ozone gas, membrane contactor for pollutant adsorption and reaction, and membrane separator for clean water production is described. The membrane reactor represents an order of magnitude improvement over traditional semibatch reactor design and is capable of complete conversion of recalcitrant endocrine disrupting compounds (EDCs) in water at less than three minutes residence time. Coating the membrane contactor with alumina and hydrotalcite (Mg/Al=3) adsorbs and traps the organics in the reaction zone resulting in 30% increase of total organic carbon (TOC) removal. Large surface area coating that diffuses surface charges from adsorbed polar organic molecules is preferred as it reduces membrane polarization that is detrimental to separation. - Graphical abstract: Advanced ozone membrane reactor synergistically combines membrane distributor for ozone, membrane contactor for sorption and reaction and membrane separator for clean water production to achieve an order of magnitude enhancement in treatment performance compared to traditional ozone reactor. Highlights: Black-Right-Pointing-Pointer Novel reactor using membranes for ozone distributor, reaction contactor and water separator. Black-Right-Pointing-Pointer Designed to achieve an order of magnitude enhancement over traditional reactor. Black-Right-Pointing-Pointer Al{sub 2}O{sub 3} and hydrotalcite coatings capture and trap pollutants giving additional 30% TOC removal. Black-Right-Pointing-Pointer High surface area coating prevents polarization and improves membrane separation and life.

Chan, Wai Kit, E-mail: kekyeung@ust.hk [Department of Chemical and Biomolecular Engineering, Hong Kong University of Science and Technology, Clear Water Bay, Kowloon (Hong Kong); Joueet, Justine; Heng, Samuel; Yeung, King Lun [Department of Chemical and Biomolecular Engineering, Hong Kong University of Science and Technology, Clear Water Bay, Kowloon (Hong Kong); Schrotter, Jean-Christophe [Water Research Center of Veolia, Anjou Recherche, Chemin de la Digue, BP 76. 78603, Maisons Laffitte, Cedex (France)

2012-05-15T23:59:59.000Z

448

Reactor physics design of supercritical CO?-cooled fast reactors  

E-Print Network [OSTI]

Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO? (S-CO?) as a Brayton cycle working fluid in a direct ...

Pope, Michael A. (Michael Alexander)

2004-01-01T23:59:59.000Z

449

Reactor protection system design alternatives for sodium fast reactors  

E-Print Network [OSTI]

Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a ...

DeWitte, Jacob D. (Jacob Dominic)

2011-01-01T23:59:59.000Z

450

United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support  

SciTech Connect (OSTI)

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

Douglas Morrell

2011-03-01T23:59:59.000Z

451

Nuclear reactor downcomer flow deflector  

DOE Patents [OSTI]

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

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

2011-02-15T23:59:59.000Z

452

Horizontal Heat Exchanger Design and Analysis for Passive Heat Removal Systems  

SciTech Connect (OSTI)

This report describes a three-year project to investigate the major factors of horizontal heat exchanger performance in passive containment heat removal from a light water reactor following a design basis accident LOCA (Loss of Coolant Accident). The heat exchanger studied in this work may be used in advanced and innovative reactors, in which passive heat removal systems are adopted to improve safety and reliability The application of horizontal tube-bundle condensers to passive containment heat removal is new. In order to show the feasibility of horizontal heat exchangers for passive containment cooling, the following aspects were investigated: 1. the condensation heat transfer characteristics when the incoming fluid contains noncondensable gases 2. the effectiveness of condensate draining in the horizontal orientation 3. the conditions that may lead to unstable condenser operation or highly degraded performance 4. multi-tube behavior with the associated secondary-side effects This project consisted of two experimental investigations and analytical model development for incorporation into industry safety codes such as TRAC and RELAP. A physical understanding of the flow and heat transfer phenomena was obtained and reflected in the analysis models. Two gradute students (one funded by the program) and seven undergraduate students obtained research experience as a part of this program.

Vierow, Karen

2005-08-29T23:59:59.000Z

453

Flow reactor experiments on the selective non-catalytic removal of nitrogen oxides  

E-Print Network [OSTI]

?CO, and H, O are initially present in exhaust stream [57]. .. . . . 42 Fig. 21 Fig. 22 Reaction path diagram for RAPRENOx process [63]. .. . Reduction of nitric oxide as a function of temperature, concentration of oxygen, carbon monoxide, and water... the influence of carbon monoxide [89]. . . . . . . . . 58 Fig. 28 Effect of residence time on the NOxOUT process as a function of temperature, NO(initial)=125ppm, 0-ratio of 4 [90]. .. . . . . . . . . . . . . . . 60 Fig. 29 Ammonia slip as a function...

Gentemann, Alexander M.G.

2001-01-01T23:59:59.000Z

454

Removal of 1,082-Ton Reactor Among Richland Operations Office's 2014  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion | Department ofT ib l L dDepartmentnews-flashes OfficeTexas |4WebinarAccess toAccomplishments

455

Terracentric Nuclear Fission Reactor: Background, Basis, Feasibility, Structure, Evidence, and Geophysical Implications  

E-Print Network [OSTI]

The background, basis, feasibility, structure, evidence, and geophysical implications of a naturally occurring Terracentric nuclear fission georeactor are reviewed. For a nuclear fission reactor to exist at the center of the Earth, all of the following conditions must be met: (1) There must originally have been a substantial quantity of uranium within Earth's core; (2) There must be a natural mechanism for concentrating the uranium; (3) The isotopic composition of the uranium at the onset of fission must be appropriate to sustain a nuclear fission chain reaction; (4) The reactor must be able to breed a sufficient quantity of fissile nuclides to permit operation over the lifetime of Earth to the present; (5) There must be a natural mechanism for the removal of fission products; (6) There must be a natural mechanism for removing heat from the reactor; (7) There must be a natural mechanism to regulate reactor power level, and; (8) The location of the reactor or must be such as to provide containment and prevent meltdown. Herndon's georeactor alone is shown to meet those conditions. Georeactor existence evidence based upon helium measurements and upon antineutrino measurements is described. Geophysical implications discussed include georeactor origin of the geomagnetic field, geomagnetic reversals from intense solar outbursts and severe Earth trauma, as well as georeactor heat contributions to global dynamics.

J. Marvin Herndon

2013-12-31T23:59:59.000Z

456

2012 Annual Report Research Reactor Infrastructure Program  

SciTech Connect (OSTI)

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

Douglas Morrell

2012-11-01T23:59:59.000Z

457

Removable bearing arrangement for a wind turbine generator  

DOE Patents [OSTI]

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

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

2010-06-15T23:59:59.000Z

458

Method for changing removable bearing for a wind turbine generator  

DOE Patents [OSTI]

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

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

2008-04-22T23:59:59.000Z

459

Sulfur dioxide removal by enhanced electrostatics  

SciTech Connect (OSTI)

The economic removal of sulfur dioxide (SO{sub 2}) still represents a significant technical challenge which could determine the use of certain types of fossil fuels for energy production. This paper will present the preliminary results of an innovative research project utilizing a low-cost wet electrostatic precipitator to remove sulfur dioxide. There are many aspects for gas removal in an electrostatic precipitator which are not currently being used. This project utilizes electron attachment of free electrons onto gas molecules and ozone generation to remove sulfur dioxide which is a typical flue gas pollutant. This research was conducted on a bench-scale, wet electrostatic precipitator. A direct-current negative discharge corona is used to generate the ozone in-situ. This ozone will be used to oxidize SO{sub 2} to form sulfuric acid, which is very soluble in water. However, it is believed that the primary removal mechanism is electron attachment of the free electrons from the corona which force the SO{sub 2} to go to equilibrium with the water and be removed from the gas stream. Forcing the equilibrium has been shown to achieve removal efficiencies of up to 70%. The bench scale unit has been designed to operate wet or dry, positive and negative for comparison purposes. The applied dc voltage is variable from 0 to 100 kV, the flow rate is a nominal 7 m{sup 3}/hr and the collecting electrode area is 0.20 m{sup 2}. Tests are conducted on a simulated flue gas stream with SO{sub 2} ranging from 0 to 4,000 ppmv. This paper presents the results of tests conducted to determine the effect of operating conditions on removal efficiency. The removal efficiency was found to vary with gas residence time, water flow rate, inlet concentration, applied power, and the use of corona pulsing.

Larkin, K.; Tseng, C.; Keener, T.C.; Khang, S.J. [Univ. of Cincinnati, OH (United States)

1997-12-31T23:59:59.000Z

460

Unique features of space reactors  

SciTech Connect (OSTI)

Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K. 8 refs., 3 figs., 1 tab.

Buden, D.

1990-01-01T23:59:59.000Z

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to obtain the most current and comprehensive results.


461

Nuclear Reactors and Technology; (USA)  

SciTech Connect (OSTI)

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

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01T23:59:59.000Z

462

Spectral shift reactor control method  

SciTech Connect (OSTI)

The method is described of closely controlling the reactor water coolant temperature of an operating spectral-shift nuclear reactor, the reactor comprising a core formed of fuel assemblies through which the reactor water coolant flows; different types of elongated elements operable to be controllably moved into and out of the core; one type of the elongated elements comprising control rods formed of neutron absorbing material and operable to decrease reactivity through neutron absorption when inserted into the core; another of the types of elongated elements comprising displacer rods formed of material which has a low absorption for neutrons and which have overall neutron-absorbing and moderating characteristics essentially not exceeding those of hollow tubular Zircaloy members with a filling zirconium oxide or aluminum oxide, the displacer rods operating to displace an equivalent volume of water coolant fluid from the core when inserted therein to decrease reactivity and to increase reactivity when moved from the core.

Impink, A.J. Jr.

1987-08-18T23:59:59.000Z

463

University Reactor Matching Grants Program  

SciTech Connect (OSTI)

During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given.

John Valentine; Farzad Rahnema; Said Abdel-Khalik

2003-02-14T23:59:59.000Z

464

Interfacial effects in fast reactors  

E-Print Network [OSTI]

The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed ...

Saidi, Mohammad Said