Powered by Deep Web Technologies
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.


1

Savannah River Site Removes Dome, Opening Reactor for Recovery Act  

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

Savannah River Site Removes Dome, Opening Reactor for Recovery Act Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning American Recovery and Reinvestment Act workers achieved a significant milestone in the decommissioning of a Cold War reactor at the Savannah River Site this month after they safely removed its rusty, orange, 75-foot-tall dome. With the help of a 660-ton crane and lifting lugs, the workers pulled the 174,000-pound dome off the Heavy Water Components Test Reactor, capping more than 16 months of preparations. Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning More Documents & Publications Recovery Act Changes Hanford Skyline with Explosive Demolitions Recovery Act Workers Add Time Capsule Before Sealing Reactor for Hundreds

2

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

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

Recovery Act Funds Test Reactor Dome Removal in Historic D&D Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project February 1, 2011 - 12:00pm Addthis Media Contacts Jim Giusti, DOE (803) 952-7697 james-r.giusti@srs.gov Paivi Nettamo, SRNS (803) 646-6075 paivi.nettamo@srs.gov AIKEN, S.C. - The landscape of the Savannah River Site (SRS) is a little flatter and a little less colorful with the removal today of the 75-foot-tall rusty-orange dome from the Cold War-era test reactor. This $25-million reactor decommissioning and deactivation project is funded By the American Recovery and Reinvestment Act. Affectionately known by SRS employees as "Hector," the iconic Heavy Water Components Test Reactor (HWCTR) has stood in the Site's B Area since 1959

3

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

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

American Recovery and Reinvestment American Recovery and Reinvestment Act workers achieved a significant milestone in the decommissioning of a Cold War reactor at the Sa- vannah River Site this month after they safely re- moved its rusty, orange, 75-foot-tall dome. With the help of a 660-ton crane and lifting lugs, the work- ers pulled the 174,000-pound dome off the Heavy Water Components Test Reactor, capping more than 16 months of preparations. Workers will cut the dome into smaller pieces for disposal. Removal of the dome allows workers to access the 219,000-pound reactor vessel and two steam generators so they can remove and permanently dispose them onsite. Re- maining equipment will be moved to the cavity vacated by the vessel, and below-grade portions of the reactor will be

4

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

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

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

5

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

Office of Environmental Management (EM)

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

6

TMI-2 reactor vessel head removal  

SciTech Connect

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

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

WCH Removes Massive Test Reactor  

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

RICHLAND, WA -- Hanford's River Corridor contractor, Washington Closure Hanford, has met a significant cleanup challenge on the U.S. Department of Energy's (DOE) Hanford Site by removing a 1,082...

9

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

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

RICHLAND, WA – Hanford’s River Corridor contractor, Washington Closure Hanford, has met a significant cleanup challenge on the U.S. Department of Energy’s (DOE) Hanford Site by removing a 1,082-ton nuclear test reactor from the 300 Area.

10

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

SciTech Connect

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

Owen, M.B.

1997-04-01T23:59:59.000Z

11

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

12

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

SciTech Connect

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

P.C. Weaver

2010-07-09T23:59:59.000Z

13

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

14

Alternative cooling resource for removing the residual heat of reactor  

SciTech Connect

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

Park, H. C.; Lee, J. H.; Lee, D. S.; Jung, C. Y.; Choi, K. Y. [Korea Hydro and Nuclear Power Co., Ltd., 260 Naa-ri Yangnam-myeon Gyeongju-si, Gyeonasangbuk-do, 780-815 (Korea, Republic of)

2012-07-01T23:59:59.000Z

15

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

16

System for fuel rod removal from a reactor module  

DOE Patents (OSTI)

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

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

1988-07-28T23:59:59.000Z

17

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

18

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

19

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

20

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

E-Print Network (OSTI)

Nitrogen removal via nitrite in a sequencing batch reactor treating sanitary landfill leachate, for the automation of a bench-scale SBR treating leachate generated in old landfills. Attention was given 20­30% due to the low biodegradability of organic matter in the leach- ate from old landfills

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

Lava Dome | Open Energy Information  

Open Energy Info (EERE)

"coulees." Volcanic domes commonly occur within the craters or on the flanks of large composite volcanoes. The nearly circular Novarupta Dome that formed during the 1912 eruption...

22

Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor  

Science Journals Connector (OSTI)

The 500 MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam–Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1–0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than ?1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using ‘CRAFT’ software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by ? factor method. The reliability target for SGDHRS arrived from DHR reliability requirement and the ultimate number of demands per year (7/y) on SGDHRS is that the failure frequency should be ?1.4E-8/de. Since it is found from the analysis that the unreliability of SGDHRS with identical loops is 5.2E-6/de and dominated by leak rates of components like AHX, DHX and sodium dump and isolation valves, options with diversity measures in important components were studied. The failure probability of SGDHRS for a design consisting of 2 types of diverse loops (Diverse AHX, DHX and sodium dump and isolation valves) is 2.1E-8/de, which practically meets the reliability requirement.

A. John Arul; C. Senthil Kumar; S. Athmalingam; Om Pal Singh; K. Suryaprakasa Rao

2006-01-01T23:59:59.000Z

23

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

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

24

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

SciTech Connect

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

E.M. Harpenau

2010-12-15T23:59:59.000Z

25

Feasibility Investigation of the Decay Heat Removal Capability Using the Concept of a Thermosyphon in the Liquid Metal Reactor  

SciTech Connect

A new design concept for a decay heat removal system in a liquid metal reactor is proposed. The new design utilizes a thermosyphon to enhance the heat removal capacity and its heat transfer characteristics are analyzed against the current PSDRS (Passive Safety Decay heat Removal System) in the KALIMER (Korea Advanced Liquid Metal Reactor) design. The preliminary analysis results show that the new design with a thermosyphon yields substantial increase of 20{approx}40% in the decay heat removal capacity compared to the current design that do not have the thermosyphon. The new design reduces the temperature rise in the cooling air of the system and helps the surrounding structure in maintaining its mechanical integrity for long term operation at an accident. Also the analysis revealed the characteristics of the interactions among various heat transfer modes in the new design. (authors)

Yeon-Sik Kim; Yoon-Sub Sim; Eui-Kwang Kim [Korea Atomic Energy Research Institute, 150, Dukjin-Dong, Yusong-Gu, Taejon 305-353 (Korea, Republic of)

2002-07-01T23:59:59.000Z

26

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

SciTech Connect

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

Lata

1996-09-26T23:59:59.000Z

27

Ecology of the Microbial Community Removing Phosphate from Wastewater under Continuously Aerobic Conditions in a Sequencing Batch Reactor  

Science Journals Connector (OSTI)

...Community Removing Phosphate from Wastewater under Continuously Aerobic...chemical oxygen demand. One wastewater was synthetic, and the other...operation. Only after repeated recycling are poly(P)-accumulating...reactors (SBRs) fed synthetic wastewater with acetate as the sole carbon...

Johwan Ahn; Sarah Schroeder; Michael Beer; Simon McIlroy; Ronald C. Bayly; John W. May; George Vasiliadis; Robert J. Seviour

2007-02-09T23:59:59.000Z

28

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

SciTech Connect

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

Neville G. Pinto; Panagiotis G. Smirniotis

2006-03-31T23:59:59.000Z

29

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

30

Dome Tech | Open Energy Information  

Open Energy Info (EERE)

Dome Tech Dome Tech Jump to: navigation, search Name Dome-Tech Place Edison, New Jersey Zip 8837 Sector Services Product Edison-based provider of services in engineering, energy consulting, & project development with the intention of optimising building performance. Coordinates 40.556614°, -82.866255° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":40.556614,"lon":-82.866255,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

31

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....

32

Uncertainty evaluation of reliability of safety grade decay heat removal system of Indian prototype fast breeder reactor  

Science Journals Connector (OSTI)

Abstract Deterministic and probabilistic safety assessment of nuclear power reactor technology is very important in assuring that the design is robust and safety systems perform as per requirement. The parameters required as input data for such analysis have uncertainties associated with them. Their impact is to be assessed on the results obtained for such analyses and it affects the overall decision making process. Safety Grade Decay Heat Removal System (SGDHRS) is one of the safety systems in fast breeder reactors and itremoves decay heat after reactor shutdown. It is a critical safety system; hence failure frequency for SGDHR is targeted to be less than 1.0 × 10?7 per reactor year. By bringing diversity in some of the components of SGDHRS, such as sodium-to-sodium decay heat exchanger (DHX), sodium to air heat exchanger (AHX) and valves, one can achieve the targeted low failure frequency of SGDHRS. We perform uncertainty analysis of the reliability of such SGDHRS here. Uncertainty in failure rate (of components of SGDHRS) is assumed to follow the log-normal distribution with error factor of three. Monte Carlo method of sampling is used in MATLAB environment. Results are obtained in terms of mean, median and standard deviation values of failure frequency. Percentile and confidence interval analysis of mean values are also obtained. These provide 95 and 98 percentile and confidence interval values of 98%, 99% and 99.8%. It is found that error factor of failure frequency of SGDHRS is found to be less than 3 in all the cases except the one in which DHX, AHX and Valves are designed with diversity in design. It is to be noted here that error factor of all input parameters distribution is taken as 3.

H.L. Kumawat; Om Pal Singh; Prabhat Munshi

2013-01-01T23:59:59.000Z

33

Salt dome discoveries mounting in Mississippi  

SciTech Connect

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

34

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

SciTech Connect

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

35

Dome-Tech and Merck Teaming Profile  

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

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

36

Experimental investigations on decay heat removal in advanced nuclear reactors using single heater rod test facility: Air alone in the annular gap  

SciTech Connect

During a loss of coolant accident in nuclear reactors, radiation heat transfer accounts for a significant amount of the total heat transfer in the fuel bundle. In case of heavy water moderator nuclear reactors, the decay heat of a fuel bundle enclosed in the pressure tube and outer concentric calandria tube can be transferred to the moderator. Radiation heat transfer plays a significant role in removal of decay heat from the fuel rods to the moderator, which is available outside the calandria tube. A single heater rod test facility is designed and fabricated as a part of preliminary investigations. The objective is to anticipate the capability of moderator to remove decay heat, from the reactor core, generated after shut down. The present paper focuses mainly on the role of moderator in removal of decay heat, for situation with air alone in the annular gap of pressure tube and calandria tube. It is seen that the naturally aspirated air is capable of removing the heat generated in the system compared to the standstill air or stagnant water situations. It is also seen that the flowing moderator is capable of removing a greater fraction of heat generated by the heater rod compared to a stagnant pool of boiling moderator. (author)

Bopche, Santosh B.; Sridharan, Arunkumar [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400 076 (India)

2010-11-15T23:59:59.000Z

37

Reliability study of a special decay heat removal system of a gas-cooled fast reactor demonstrator  

Science Journals Connector (OSTI)

Abstract The European roadmap toward the development of generation IV concepts addresses the safety and reliability assessment of the special system designed for decay heat removal of a gas-cooled fast reactor demonstrator (GFRD). The envisaged system includes the combination of both active and passive means to accomplish the fundamental safety function. Failure probabilities are calculated on various system configurations, according to either pressurized or depressurized accident events under investigation, and integrated with probabilities of occurrence of corresponding hardware components and natural circulation performance assessment. The analysis suggests the improvement of measures against common cause failures (CCF), in terms of an appropriate diversification among the redundant systems, to reduce the system failure risk. Particular emphasis is placed upon passive system reliability assessment, being recognized to be still an open issue, and the approach based on the functional reliability is adopted to address the point. Results highlight natural circulation as a challenging factor for the decay heat removal safety function accomplishment by means of passive devices. With the models presented here, the simplifying assumptions and the limited scenarios considered according to the level of definition of the design, where many systems are not yet established, one can conclude that attention has to be paid to the functional aspects of the passive system, i.e. the ones not pertaining to the “hardware” of the system. In this article the results of the analysis are discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The design diversity of the components undergoing \\{CCFs\\} can be effective for the improvement and some accident management measures are also possible by making use of the long grace period in GFRD.

Luciano Burgazzi

2014-01-01T23:59:59.000Z

38

Resurgent Dome Complex | Open Energy Information  

Open Energy Info (EERE)

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

39

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

Office of Legacy Management (LM)

Tatum Salt Dome Test Site - MS 01 FUSRAP Considered Sites Site: Tatum Salt Dome Test Site (MS.01) Designated Name: Alternate Name: Location: Evaluation Year: Site Operations: Site...

40

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

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

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

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

Energy Department Sells Historic Teapot Dome Oilfield  

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

Today, the Energy Department finalized the sale of the historic Teapot Dome Oilfield located 35 miles north of Casper, Wyoming to Stranded Oil Resources Corporation, a subsidiary of Allegheny Capital Corporation.

42

Integrated continuous dissolution, refolding and tag removal of fusion proteins from inclusion bodies in a tubular reactor  

Science Journals Connector (OSTI)

Abstract An integrated continuous tubular reactor system was developed for processing an autoprotease expressed as inclusion bodies. The inclusion bodies were suspended and fed into the tubular reactor system for continuous dissolving, refolding and precipitation. During refolding, the dissolved autoprotease cleaves itself, separating the fusion tag from the target peptide. Subsequently, the cleaved fusion tag and any uncleaved autoprotease were precipitated out in the precipitation step. The processed exiting solution results in the purified soluble target peptide. Refolding and precipitation yields performed in the tubular reactor were similar to batch reactor and process was stable for at least 20 h. The authenticity of purified peptide was also verified by mass spectroscopy. Productivity (in mg/l/h and mg/h) calculated in the tubular process was twice and 1.5 times of the batch process, respectively. Although it is more complex to setup a tubular than a batch reactor, it offers faster mixing, higher productivity and better integration to other bioprocessing steps. With increasing interest of integrated continuous biomanufacturing, the use of tubular reactors in industrial settings offers clear advantages.

Siqi Pan; Monika Zelger; Alois Jungbauer; Rainer Hahn

2014-01-01T23:59:59.000Z

43

Tatum Dome Project, Lamar County, Mississippi  

SciTech Connect

The Tatum Dome site, near Hattiesburg, Mississippi, has been host to two nuclear detonations and two nonnuclear gas detonations. The top of the cavity formed by the detonations was 2660 feet below land surface and 1160 feet below the top of the salt dome. All detonations were totally contained within the salt dome in which they were fired. Activities at the site occurred between 1964 ad 1970. In 1978, NV was asked to provide proof that radiological contamination was not adversely affecting water quality in the aquifers over Tatum Dome. All except the Surficial Aquifer, which was known to be contaminated by tritium, were extensively tested to obtain hydrological data. Sufficient samples were taken to determine if contamination existed in the aquifer around the emplacement casing at surface ground zero (SGZ). Low tritium concentrations were found in the Local Aquifer, approximately 150 to 200 feet below land surface, during this project, but no radioactivity was found in any of the other aquifers above Tatum Dome. The Local Aquifer is the next aquifer below the Surficial Aquifer.

Fenske, P.R.; Humphrey, T.M. Jr.

1980-08-01T23:59:59.000Z

44

Geology of Damon Mound Salt Dome, Texas  

SciTech Connect

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

45

Ecology of the Microbial Community Removing Phosphate from Wastewater under Continuously Aerobic Conditions in a Sequencing Batch Reactor  

Science Journals Connector (OSTI)

...with synthetic wastewater eventually containing...the clarified plant effluent (Fig...Using synthetic wastewater. A typical profile...A carbon mass balance based on existing...phosphorus removal in wastewater treatment plants. Antonie Leeuwenhoek...

Johwan Ahn; Sarah Schroeder; Michael Beer; Simon McIlroy; Ronald C. Bayly; John W. May; George Vasiliadis; Robert J. Seviour

2007-02-09T23:59:59.000Z

46

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... Andrew Hluchanek, B. S. , Texas A&M University Directed by: Dr. Robert R. Unterberger Radar probing through salt was accomplished at 17 radar stations established in the United Salt Company mine at Hockley, Texas. The top of the salt dom is mapped...

Hluchanek, James Andrew

2012-06-07T23:59:59.000Z

47

Occurrence of gypsum in Gulf coast salt domes  

Science Journals Connector (OSTI)

...Occurrence of gypsum in Gulf coast salt domes Barton Donald Clinton...OF GYPSUM IN THE GULF COAST SALT DOMES. Sir: On accountof thepaucityof...concerningtheoccurrenceof gypsumandanhydriteon the salt domes. The followingremarksmay...5o-footsill of saltat 3,350feetat Palangana,possiblyalsoStrattonRidge...

Donald Clinton Barton

48

Ge atom distribution in buried dome islands  

SciTech Connect

Laser-assisted atom probe tomography microscopy is used to provide direct and quantitative compositional measurements of tri-dimensional Ge distribution in Ge dome islands buried by Si. Sub-nanometer spatial resolution 3D imaging shows that islands keep their facets after deposition of the Si cap, and that the island/substrate/Si cap interfaces are abrupt. The core of the domes contains 55% of Ge, while the island shell exhibits a constant composition of 15% of Ge. The {l_brace}113{r_brace} facets of the islands present a Ge enrichment up to 35%. The wetting layer composition is not homogeneous, varying from 9.5% to 30% of Ge.

Portavoce, A.; Berbezier, I.; Ronda, A.; Mangelinck, D. [CNRS, IM2NP, Case 142, 13397 Marseille Cedex 20 (France); Hoummada, K. [Aix-Marseille Universite, IM2NP, Case 142, 13397 Marseille Cedex 20 (France)

2012-04-16T23:59:59.000Z

49

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

SciTech Connect

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

50

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

Open Energy Info (EERE)

A detailed inventory was made of subsurface samples taken from deep geothermal test wells drilled in the resurgent Redondo dome in the Valles caldera of New Mexico. Attention...

51

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

52

Geology of Bravo Dome carbon dioxide gas field, New Mexico  

SciTech Connect

The Bravo Dome carbon dioxide gas field is located in Union and Harding Counties of northeast New Mexico. The Bravo Dome field covers approximately 800,000 acres, but areal boundaries of the field have not been fully defined. Production in 1989 was 113 bcf of gas from 272 wells. Cumulative production at the end of 1989 was 626 bcf. Estimated recoverable reserves are more than 10 tcf. The gas is 98-99% CO{sub 2}. Most CO{sub 2} produced from Bravo Dome is used for enhanced oil recovery in the Permian basin. The Bravo Dome is a faulted, southeast-plunging, basement-cored anticlinal nose. It is bordered on the east and south by large high-angle faults of Pennsylvanian and Wolfcampian (Early Permian) age. The principal reservoir in the Bravo Dome field is the Tubb sandstone (Leonardian-Permian) at depths of 1,900 to 2,950 ft. The Tubb consists of 0-400 ft of fine- to medium-grained, well-sorted, orange feldspathic sandstone. It rests unconformably on Precambrian basement on the highest parts of the Bravo Dome and is not offset by late Paleozoic faults that form the dome. The Cimmaron Anhydrite (Leonardian-Permian) conformably overlies the Tubb and is a vertical seal. The trap at Bravo Dome has structural and stratigraphic aspects. Drape of Tubb sandstone over the dome created structural closure on the northeast, southeast, and southwest flanks of the field. Trapping on the northwest flank of the field is associated with regional northwest thinning of the Tubb.

Broadhead, R.F. (New Mexico Bureau of Mines and Mineral Resources, Socorro (United States))

1991-03-01T23:59:59.000Z

53

F Reactor Inspection | Department of Energy  

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

Inspection F Reactor Inspection Addthis Description Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor last week before...

54

Final report on decommissioning of wells, boreholes, and tiltmeter sites, Gulf Coast Interior Salt Domes of Louisiana  

SciTech Connect

In the late 1970s, test holes were drilled in northern Louisiana in the vicinity of Vacherie and Rayburn`s Salt Domes as part of the Department of Energy`s (DOE) National Waste Terminal Storage (NWTS) (rename the Civilian Radioactive Waste Management (CRWM)) program. The purpose of the program was to evaluate the suitability of salt domes for long term storage or disposal of high-level nuclear waste. The Institute for Environmental Studies at Louisiana State University (IES/LSU) and Law Engineering Testing Company (LETCo) of Marietta, Georgia performed the initial field studies. In 1982, DOE awarded a contract to the Earth Technology Corporation (TETC) of Long Beach, California to continue the Gulf Coast Salt Dome studies. In 1986, DOE deferred salt domes from further consideration as repository sites. This report describes test well plugging and site abandonment activities performed by SWEC in accordance with Activity Plan (AP) 1--3, Well Plugging and Site Restoration of Work Sites in Louisiana. The objective of the work outlined in this AP was to return test sites to as near original condition as possible by plugging boreholes, removing equipment, regrading, and seeding. Appendices to this report contain forms required by State of Louisiana, used by SWEC to document decommissioning activities, and pertinent documentation related to lease/access agreements.

Not Available

1989-07-01T23:59:59.000Z

55

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

Open Energy Info (EERE)

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

56

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

57

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)

in specific humidity within their cold domes. As a result, increases in the latent heat fluxes observed in the cold domes were more dependent on the increases in wind speed. Stronger cold domes (cool, dry and strong winds) were associated with squall line MCSs...

Caesar, Kathy-Ann Lois

2012-06-07T23:59:59.000Z

58

Former Reactor Facilities Surveillance and Maintenance and  

E-Print Network (OSTI)

and Dark (2010) Majority of remaining radiation contained in reactor vessel (inside biological shield window damage Minor roof leaks to former office areas Stack and Grounds: Stack drain tank Safety of confinement dome Replaced exhaust fan Security system hardware/software upgrade BGRR: Covered vents above 2

Ohta, Shigemi

59

LANL demolishes first containment dome at disposal area  

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

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

60

A mechanical model of early salt dome growth  

E-Print Network (OSTI)

A MECHANICAL MODEL OF EARLY SALT DOME GROWTH A Thesis by FRANK ALBERT IRWIN Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE December 1988... Major Subject: Geology A MECHANICAL MODEL OF EARLY SALT DOME GROWTH A Thesis by FRANK ALBERT IRWIN Approved as to style and content by: aymond C. Fletcher (Chair of Committee) John H. Spang (Member) Wi tamR. B ant (Mem ) John H. Sp g (Head...

Irwin, Frank Albert

2012-06-07T23:59:59.000Z

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

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

SciTech Connect

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

62

Tatum Dome field study and analysis of monitoring data  

SciTech Connect

This report is a summary and interpretation of data collected at the Tatum Dome Site during April and May 1984. In addition to raw data, description of field activities and laboratory analytical results are presented. The report also includes all analytical results of shallow well samples from the site, and chemical, mineralogical and mechanical analyses of soil samples collected during April 1985. Analysis and interpretation of laboratory analyses are included.

Fordham, J.W.; Fenske, P.R.

1985-12-01T23:59:59.000Z

63

Thermal degradation of acrylonitrile–butadiene–styrene (ABS) containing flame retardants using a fluidized bed reactor: The effects of Ca-based additives on halogen removal  

Science Journals Connector (OSTI)

In this study, the thermal degradation of a waste fraction of acrylonitrile–butadiene–styrene containing brominated flame retardants was performed to reduce halogen content in the pyrolysis oil. Thermal degradation was completed using Ca-based additives (calcium oxide, calcium hydroxide and oyster shells) in a bench-scale pyrolysis plant equipped with a fluidized bed reactor and char separation system. Pyrolysis was carried out in a temperature range of 430–510 °C. In the absence of any additive, the oil yield amounted to about 77 wt.%. With the additives, the oil yield was markedly reduced to within a range of 45–64 wt.%. The principle compounds in the oils were toluene, ethylbenzene, styrene, cumene, ?-methylstyrene, phenol and heteroatom-containing compounds. When Ca(OH)2 was applied, total bromine and chlorine contents in the oil decreased to 0.05 and 0.04 wt.%, respectively. In addition, Ca(OH)2 reduced the antimony content in the oil to below 0.001 ppm. Most of the halogens and antimony in the feed material were present in the char obtained after pyrolysis.

Su-Hwa Jung; Seon-Jin Kim; Joo-Sik Kim

2012-01-01T23:59:59.000Z

64

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

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

65

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

SciTech Connect

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

66

Predicting optical and thermal characteristics of transparent single-glazed domed skylights  

SciTech Connect

Optical and thermal characteristics of domed skylights are important to solve the trade-off between daylighting and thermal design. However, there is a lack of daylighting and thermal design tools for domed skylights. Optical and thermal characteristics of transparent single-glazed hemispherical domed skylights under sun and sky light are evaluated based on an optical model for domed skylights. The optical model is based on tracing the beam and diffuse radiation transmission through the dome surface. A simple method is proposed to replace single-glazed hemispherical domed skylights by optically and thermally equivalent single-glazed planar skylights to accommodate limitations of energy computer programs. Under sunlight, single-glazed hemispherical domed skylights yield slightly lower equivalent solar transmittance and solar heat gain coefficient (SHGC) at near normal zenith angles than those of single-glazed planar skylights. However, single-glazed hemispherical domed skylights yield substantially higher equivalent solar transmittance and SHGC at high zenith angles and around the horizon. Under isotropic skylight, single-glazed hemispherical domed skylights yield slightly lower equivalent solar transmittance and SHGC than those of single-glazed planar skylights. Daily solar heat gains of single-glazed hemispherical domed skylights are higher than those of single-glazed horizontal planar skylights in both winter and summer. In summer, the solar heat gain of single-glazed hemispherical domed skylights can reach 3% to 9% higher than those of horizontal single-glazed planar skylights for latitudes varying between 0 and 55{degree} (north/south). In winter, however, the solar heat gains of single-glazed hemispherical domed skylights increase significantly with the increase of the site latitude and can reach 232% higher than those of horizontal single-glazed planar skylights, particularly for high latitude countries.

Laouadi, A.; Atif, M.R.

1999-07-01T23:59:59.000Z

67

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

68

Let’s Try That Again: Selling the Teapot Dome Oil Field  

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

The first time the Teapot Dome Oilfield was sold, it threw the Harding administration into scandal. Now -- 93 years later -- we're selling the field legally.

69

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

SciTech Connect

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

70

Nuclear Reactor (atomic reactor)  

Science Journals Connector (OSTI)

A nuclear reactor splits Uranium or Plutonium nuclei, and the...235 is fissionable but more than 99% of the naturally occurring Uranium is U238 that makes enrichment mandatory. In some reactors U238 and Thorium23...

2008-01-01T23:59:59.000Z

71

Radar investigation of the Cote Blanche salt dome  

E-Print Network (OSTI)

PADAR IiVVESTICATIOH CF THE COZE ELANCHR SALT DO&: A Thests EOHEET DOSAED SZEVAET Subm-'tt. ii' to the Crsduete Coilege of Ter: s AVi', i;nlu xsity in partfal fulfi' line?t of th ': quiremen fc z' tht degree o %P t S "t Clt. 'iCE iugust. l...HIC major Sub jest: Ceoohysfes RADAR INVESTIGATION OP THE COTE BLANCHE SALT DOME A Thesis by ROBERT DONALD STEWART Approved as to style and nor. tent by: (Chairman of Comml ee) ( Ir~c" (Head of De rtment ? Member) (Member) August 1974 ABSTRACT...

Stewart, Robert Donald

2012-06-07T23:59:59.000Z

72

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

73

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

Open Energy Info (EERE)

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

74

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

SciTech Connect

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

75

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

SciTech Connect

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

76

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

SciTech Connect

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

77

Vlf Electromagnetic Investigations Of The Crater And Central Dome Of Mount  

Open Energy Info (EERE)

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

78

Hanford Deep Dig Removes Contaminated Soil | Department of Energy  

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

Deep Dig Removes Contaminated Soil Deep Dig Removes Contaminated Soil Hanford Deep Dig Removes Contaminated Soil March 11, 2013 - 12:00pm Addthis An aerial view of Hanford’s D Area shows the D Reactor (lower left) and DR Reactor. Workers are digging 85 feet to groundwater at two sites there to remove chromium contamination. An aerial view of Hanford's D Area shows the D Reactor (lower left) and DR Reactor. Workers are digging 85 feet to groundwater at two sites there to remove chromium contamination. Workers remove soil contaminated with sodium dichromate to prevent the chemical from reaching the groundwater and eventually the Columbia River. Workers remove soil contaminated with sodium dichromate to prevent the chemical from reaching the groundwater and eventually the Columbia River.

79

Hanford Deep Dig Removes Contaminated Soil | Department of Energy  

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

Hanford Deep Dig Removes Contaminated Soil Hanford Deep Dig Removes Contaminated Soil Hanford Deep Dig Removes Contaminated Soil March 11, 2013 - 12:00pm Addthis An aerial view of Hanford’s D Area shows the D Reactor (lower left) and DR Reactor. Workers are digging 85 feet to groundwater at two sites there to remove chromium contamination. An aerial view of Hanford's D Area shows the D Reactor (lower left) and DR Reactor. Workers are digging 85 feet to groundwater at two sites there to remove chromium contamination. Workers remove soil contaminated with sodium dichromate to prevent the chemical from reaching the groundwater and eventually the Columbia River. Workers remove soil contaminated with sodium dichromate to prevent the chemical from reaching the groundwater and eventually the Columbia River.

80

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

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

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

82

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

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

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

1996-02-27T23:59:59.000Z

83

Time of growth of Opelika Dome, Henderson and Van Zandt counties, Texas  

E-Print Network (OSTI)

ABSTRACT The Gpelilca fielc! is located in lienderson nuM Van Zandt countiss? Texas. This location is ir. the western part c f the petroliferous F~t Texas Basin. The fielo has a dome-shapeu stxucture wl&ich is believed to hove been caused bi' a &esp...-seated salt ccome. At least 27 salt domes have been identifiec! in the Fast Texas Basin ared the;-. tructures of the known salt doves ars similar to the structure of the Opelika dome. The times and amounts cf, &cx&wth of ths & ore as well as the tires...

Peterson, Thomas Lowe

2012-06-07T23:59:59.000Z

84

Structural stability of andesite volcanoes and lava domes  

Science Journals Connector (OSTI)

...limit equilibrium analysis (LEA) methods...high degree of reliability. The principles...quantitative stability analyses developed. It...highly uncertain reliability. By mid-May...Finite element analysis of deformation...Structural Mechanics Reactor Technology, Chicago...

2000-01-01T23:59:59.000Z

85

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

Open Energy Info (EERE)

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

86

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

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

87

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

E-Print Network (OSTI)

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

Clark, Ryan J

2005-01-01T23:59:59.000Z

88

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

Open Energy Info (EERE)

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

89

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

E-Print Network (OSTI)

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... of MASTER OF SCIENCE December 1987 Major Subject: Oceanography VARIABILITY OF BOTTOM WATER DOMES AND GEOSTROPHIC CURRENTS IN THE EASTERN GULF OF ~ A Thesis by ERIK SAUL GOTTLIEB Approved as to style and content by: Davi A. rooks (Chairman...

Gottlieb, Erik S

2012-06-07T23:59:59.000Z

90

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

SciTech Connect

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

91

Reactor Thermal-Hydraulics  

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

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

92

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

93

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

SciTech Connect

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

94

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

SciTech Connect

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

95

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

SciTech Connect

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

96

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

97

Silica Scaling Removal Process  

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

Silica Scaling Removal Process Silica Scaling Removal Process Scientists at Los Alamos National Laboratory have developed a novel technology to remove both dissolved and colloidal...

98

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

99

F Reactor Inspection  

SciTech Connect

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

100

Nuclear divisional reactor  

SciTech Connect

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

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

Degradation of Dome Cutting Minerals in Hanford Waste - 13100  

SciTech Connect

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

102

Degradation of dome cutting minerals in Hanford waste  

SciTech Connect

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

103

Inferences On The Hydrothermal System Beneath The Resurgent Dome In Long  

Open Energy Info (EERE)

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

104

Transport reactor development status  

SciTech Connect

This project is part of METC`s Power Systems Development Facility (PSDF) located at Wilsonville, Alabama. The primary objective of the Advanced Gasifier module is to produce vitiated gases for intermediate-term testing of Particulate Control Devices (PCDs). The Transport reactor potentially allows particle size distribution, solids loading, and particulate characteristics in the off-gas stream to be varied in a number of ways. Particulates in the hot gases from the Transport reactor will be removed in the PCDs. Two PCDs will be initially installed in the module; one a ceramic candle filter, the other a granular bed filter. After testing of the initial PCDs they will be removed and replaced with PCDs supplied by other vendors. A secondary objective is to verify the performance of a Transport reactor for use in advanced Integrated Gasification Combined Cycle (IGCC), Integrated Gasification Fuel Cell (IG-FC), and Pressurized Combustion Combined Cycle (PCCC) power generation units. This paper discusses the development of the Transport reactor design from bench-scale testing through pilot-scale testing to design of the Process Development Unit (PDU-scale) facility at Wilsonville.

Rush, R.E.; Fankhanel, M.O.; Campbell, W.M.

1994-10-01T23:59:59.000Z

105

Natural Convection Shutdown Heat Removal Test Facility (NSTF)  

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

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

106

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

SciTech Connect

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

Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y. [Korea Atomic Energy Research Inst., 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2012-07-01T23:59:59.000Z

107

Tritium Removal from Carbon Plasma Facing Components  

SciTech Connect

Tritium removal is a major unsolved development task for next-step devices with carbon plasma-facing components. The 2-3 order of magnitude increase in duty cycle and associated tritium accumulation rate in a next-step tokamak will place unprecedented demands on tritium removal technology. The associated technical risk can be mitigated only if suitable removal techniques are demonstrated on tokamaks before the construction of a next-step device. This article reviews the history of codeposition, the tritium experience of TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) and the tritium removal rate required to support ITER's planned operational schedule. The merits and shortcomings of various tritium removal techniques are discussed with particular emphasis on oxidation and laser surface heating.

C.H. Skinner; J.P. Coad; G. Federici

2003-11-24T23:59:59.000Z

108

Brookhaven Lab Completes Decommissioning of Graphite Research Reactor:  

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

Brookhaven Lab Completes Decommissioning of Graphite Research Brookhaven Lab Completes Decommissioning of Graphite Research Reactor: Reactor core and associated structures successfully removed; waste shipped offsite for disposal Brookhaven Lab Completes Decommissioning of Graphite Research Reactor: Reactor core and associated structures successfully removed; waste shipped offsite for disposal September 1, 2012 - 12:00pm Addthis The Brookhaven Graphite Research Reactor’s bioshield, which contains the 700-ton reactor core, is shown prior to decommissioning. The Brookhaven Graphite Research Reactor's bioshield, which contains the 700-ton reactor core, is shown prior to decommissioning. Pictured here is the Brookhaven Graphite Research Reactor, where major decommissioning milestones were recently reached after the remaining radioactive materials from the facility’s bioshield were shipped to a licensed offsite disposal facility.

109

Analysis of the Reactor Cavity Cooling System for Very High Temperature Gas-cooled Reactors Using Computational Fluid Dynamics Tools  

E-Print Network (OSTI)

The design of passive heat removal systems is one of the main concerns for the modular Very High Temperature Gas-Cooled Reactors (VHTR) vessel cavity. The Reactor Cavity Cooling System (RCCS) is an important heat removal system in case of accidents...

Frisani, Angelo

2011-08-08T23:59:59.000Z

110

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

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

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

111

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

112

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

113

Workers Enter Cocooned F Reactor for Scheduled Inspection  

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

RICHLAND, Wash. – Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor last week before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008.

114

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

115

Beryllium-10 in the Taylor Dome ice core: Applications to Antarctic glaciology and paleoclimatology  

SciTech Connect

An ice core was drilled at Taylor dome, East Antarctica, reaching to bedrock at 554 meters. Oxygen-isotope measurements reveal climatic fluctuations through the last interglacial period. To facilitate comparison of the Taylor Dome paleoclimate record with geologic data and results from other deep ice cores, several glaciological issues need to be addressed. In particular, accumulation data are necessary as input for numerical ice-flow-models, for determining the flux of chemical constituents from measured concentrations, and for calculation of the offset in age between ice and trapped air in the core. The analysis of cosmogenic beryllium-10 provides a geochemical method for constraining the accumulation-rate history at Taylor Dome. High-resolution measurements were made in shallow firn cores and snow pits to determine the relationship among beryllium-10 concentrations, wet and dry deposition mechanisms, and snow-accumulation rates. Comparison between theoretical and measured variations in deposition over the last 75 years constrains the relationship between beryllium-10 deposition and global average production rates. The results indicate that variations in geomagnetically-modulated production-rate do not strongly influence beryllium-10 deposition at Taylor Dome. Although solar modulation of production rate is important for time scales of years to centuries, snow-accumulation rate is the dominant control on ice-core beryllium-10 concentrations for longer periods. Results show that the Taylor Dome core can be used to provide new constraints on regional climate over the last 130,000 years, complementing the terrestrial and marine geological record from the Dry Valley, Transantarctic Mountains and western Ross Sea.

Steig, E.J.

1996-12-31T23:59:59.000Z

116

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

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

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

117

NUCLEAR REACTORS.  

E-Print Network (OSTI)

??Nuclear reactors are devices containing fissionable material in sufficient quantity and so arranged as to be capable of maintaining a controlled, self-sustaining NUCLEAR FISSION chain… (more)

Belachew, Dessalegn

2010-01-01T23:59:59.000Z

118

Technologies for Boron Removal  

Science Journals Connector (OSTI)

Tests were performed to examine the removal of boron from aqueous solution either with polyvinyl alcohol (PVA) alone or by both PVA and other inorganic additives under room temperature. ... Added calcium hydroxide increased the co-removal of borate with PVA, and this offers a polishing treatment after borate removal by liming. ... As boron removal can be achieved by chemical precipitation and coagulation, it is logical to assume that the EC could remove boron from water and industrial effluent. ...

Yonglan Xu; Jia-Qian Jiang

2007-11-23T23:59:59.000Z

119

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

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

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

120

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)

oT ~ttee THd ORIGTN OF Tiki STRUCTURAL S&iR "'SION ABOV' ULF COAST SALT KRISS UITH PARTI'ULAR ~~JCS TO ' LAY CRgdK DGIQ JA' HI:. ' TON COK;TY~ TWAS Alfred Norman licDowsll A Thesis Submitted to the Graduate School of the Agricultural... Application of kodel Theory to uepression problem 13 Salt mis 'cele Models 14 Hodel Fault Patterns . 16 Aegio. ial Setting of Slay Greek oalt Dome. . . ~ Structural Interpretation of Faulting on Glar Greek Tertiary Growth History of 'la; Greek Some...

McDowell, Alfred Norman

2012-06-07T23:59:59.000Z

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

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

122

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

123

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

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

124

Seal integrity and feasibility of CO2 sequestration in the Teapot Dome EOR pilot: geomechanical site characterization  

Science Journals Connector (OSTI)

This paper reports a preliminary investigation of CO2 sequestration and seal integrity at Teapot Dome oil field, Wyoming, USA, with...2 leakage along reservoir-bounding faults. CO2 injection into reservoirs creat...

Laura Chiaramonte; Mark D. Zoback; Julio Friedmann; Vicki Stamp

2008-06-01T23:59:59.000Z

125

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

SciTech Connect

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

126

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

E-Print Network (OSTI)

SALT TECI'ONISM AND SEISMIC STRATIGRAPHY OF THE UPPER JURASSIC IN THE DESTIN DOME REGION, NORTHEASTERN GULF OF MEXICO A Thesis by GRANT MACRAE Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment... of the requirements for the degree of MASTER OF SCIENCE August 1990 Major Subject: Oceanography SALT TECI'ONISM AND SEISMIC STRATIGRAPHY OF THE UPPER JURASSIC IN THE DESTIN DOME REGION, NORTHEASTERN GULF OF MEXICO A Thesis by GRANT MACRAE Approved...

MacRae, Grant

2012-06-07T23:59:59.000Z

127

Use of dipmeter logs to refine structural mapping of Teapot Dome, Wyoming  

SciTech Connect

Teapot field, now Naval Petroleum Reserve 3, is an elongated, asymmetric structural dome with a north-northeast axial trend located on the southwest edge of the Powder River basin. Currently, more than 800 wells of various depths penetrate multiple reservoirs; over 30 dipmeter logs have been run during the past 34 years. Although structure contour maps of individual stratigraphic horizons have been drawn simply from conventional well data, more subtle features of deformation are interpreted from the use of dipmeter logs. Because dips are generally less than 15/sup 0/, careful computation of transverse and longitudinal dip directions was required for detailed structural analysis. All depth, dip, and azimuth data were entered into computer files, and a flow chart of steps for computer processing and structural interpretation was devised and followed. As designed by C.A. Bengston, SCAT (Statistical Curvature Analysis Technique) plots were drawn by computer. Interpretation of SCAT plots yielded quantitative descriptions of the asymmetry of Teapot dome, curvature of the anticlinal axial plane, vertical discontinuity of beds, location and orientation of normal faults, and curvature of beds in drag zones adjacent to faults. Structural definition was necessary to outline boundaries of reservoirs with tilted fluid contacts on the flanks of the dome and along fault planes. Location of such faults would be particularly important for the discovery of deeper pools, for instance, in the Tensleep Formation.

Beinkafner, K.J.

1986-08-01T23:59:59.000Z

128

TMI defueling project fuel debris removal system  

SciTech Connect

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

129

TMI defueling project fuel debris removal system  

SciTech Connect

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

130

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

131

naval reactors  

National Nuclear Security Administration (NNSA)

After operating for 34 years and training over 14,000 sailors, the Department of Energy S1C Prototype Reactor Site in Windsor, Connecticut, was returned to "green field"...

132

EM Employs Innovative Technology to Remove Radioactive Sludge | Department  

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

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

133

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

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

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

134

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

135

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

136

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

SciTech Connect

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

137

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

SciTech Connect

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

138

Research reactors - an overview  

SciTech Connect

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

West, C.D.

1997-03-01T23:59:59.000Z

139

Light Water Reactor Sustainability  

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

140

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

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

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

142

Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites |  

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

Evaluation of Removing Used Nuclear Fuel From Shutdown Evaluation of Removing Used Nuclear Fuel From Shutdown Sites Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites In January 2013, the Department of Energy issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America's Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses. Shutdown sites are defined as those commercial nuclear power reactor sites where the

143

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

144

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor  

SciTech Connect

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

145

Risk Removal | Department of Energy  

Energy Savers (EERE)

Risk Removal Risk Removal Workers safely remove old mercury tanks from the Y-12 National Security Complex. Workers safely remove old mercury tanks from the Y-12 National Security...

146

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

147

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab | Department  

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

Workers Clear Reactor Shields from Brookhaven Lab Workers Clear Reactor Shields from Brookhaven Lab Recovery Act Workers Clear Reactor Shields from Brookhaven Lab American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once used to absorb neutrons produced for research. The Brookhaven National Laboratory is using $39 million from the Recovery Act to decommission the Brookhaven Graphite Research Reactor, the world's first reactor built solely for peaceful research purposes. Recovery Act Workers Clear Reactor Shields from Brookhaven Lab More Documents & Publications Brookhaven Graphite Research Reactor Workshop 2011 ARRA Newsletters Idaho Crews Overcome Challenges to Safely Dispose 1-Million-Pound Hot Cell

148

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

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

Printer Friendly Printer Friendly Engineering Test Reactor (ETR) Vessel Relocated Engineering Test Reactor Vessel Pre-startup 1957 Click on image to enlarge. Image 1 of 5 Gantry jacks attached to ETR vessel. Initial lift starts. - Click on image to enlarge. Image 2 of 5 ETR vessel removed from substructure. Vessel lifted approximately 40 ft. - Click on image to enlarge. On Monday, September 24, 2007 the Engineering Test Reactor (ETR) vessel was removed from its location and delivered to the Idaho CERCLA Disposal Facility (ICDF). The long history of the ETR began for this water-cooled reactor with its start up in 1957, after taking only 2 years to build. According to "Proving the Principles," by Susan M. Stacy: When the Engineering Test Reactor started up at the Test Reactor Area in

149

Silica Scaling Removal Process  

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

Silica Scaling Removal Process Silica Scaling Removal Process Silica Scaling Removal Process Scientists at Los Alamos National Laboratory have developed a novel technology to remove both dissolved and colloidal silica using small gel particles. Available for thumbnail of Feynman Center (505) 665-9090 Email Silica Scaling Removal Process Applications: Cooling tower systems Water treatment systems Water evaporation systems Potential mining applications (produced water) Industry applications for which silica scaling must be prevented Benefits: Reduces scaling in cooling towers by up to 50% Increases the number of cycles of concentration substantially Reduces the amount of antiscaling chemical additives needed Decreases the amount of makeup water and subsequent discharged water (blowdown) Enables considerable cost savings derived from reductions in

150

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

SciTech Connect

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

151

Lead, zinc, and strontium in limestone cap rock from Tatum salt dome, Mississippi  

SciTech Connect

Limestone cap rock at Tatum salt dome, Mississippi, contains disseminated pyrite, sphalerite, and galena, and disseminated to massive amounts of strontianite (SrCO/sub 3/) and celestite (SrSO/sub 4/). Sulfide minerals are locally present in bitumen-rich areas of the upper, massive portion of the limestone cap rock, whereas strontium minerals are disseminated throughout this zone. However, sulfide and strontium minerals are most abundant in the lower banded portion of the limestone cap rock, which consists of alternating subhorizontal light and dark-colored bands. The dark bands are composed of calcite of variable grain size, sulfides, quartz, dolomite, albite, and up to 1% bitumen that apparently formed by the biodegradation of crude oil. Lighter bands are composed of variable amounts of coarsely crystalline, euhedral calcite, strontianite, and celestite resulting in strontium (Sr) contents of up to 30% locally. Banded limestone cap rock at Tatum dome formed at the top of the actively dissolving anhydrite zone by a combination of sulfate reduction and oxidation of liquid hydrocarbons by bacteria to cause the precipitation of calcite and sulfide minerals and the accumulation of insoluble residue from the anhydrite (quartz, albite, dolomite). Lead and zinc in the sulfide minerals could have been derived from the dissolving anhydrite, but the abundance of Sr minerals present requires an external source. Present-day oil field brines in central Mississippi contain up to 3000 ppm Sr, and basin brines of similar composition apparently contributed Sr to the cap-rock environment during formation.

Saunders, J.A.

1988-09-01T23:59:59.000Z

152

Identification of damage in dome-like structures using hybrid sensor measurements and artificial neural networks  

Science Journals Connector (OSTI)

A damage detection scheme using multi-type sensor-based hybrid sensing and artificial-neural-network- (ANN-) based information processing was developed for dome-like structures used in civil infrastructure. Accelerometers and strain sensors were used to provide a hybrid measurement with the purpose of acquiring rich information associated with structural damage. The optimal placement of multiple sensors was explored so as to capture the most appropriate and sensitive signal features (damage parameter vectors) for damage characterization. A back-propagation ANN was constructed with the inputs extracted from the hybrid measurement. To validate the capacity of the proposed damage identification scheme, finite element analysis was conducted to identify damage in a Schwedler dome structure as an example. The performance of ANNs, trained by three kinds of damage parameter vector extracted from signals captured by (i) a sole accelerometer, (ii) a sole strain sensor, and (iii) both kinds of sensor was compared, to observe that the one trained by hybrid sensor measurement outperformed the others. Error analysis for a series of parametric studies, in which noise at different levels was included in the training input, was further carried out, and robustness of the proposed damage identification scheme under noisy measurement was demonstrated.

Wei Lu; Jun Teng; Youlin Xu; Zhongqing Su

2013-01-01T23:59:59.000Z

153

Biochemical Removal of HAP Precursors from Coal  

SciTech Connect

Column biooxidation tests with Kentucky coal confirmed results of earlier shake flask tests showing significant removal from the coal of arsenic, selenium, cobalt, manganese, nickel and cadmium. Rates of pyrite biooxidation in Kentucky coal were only slightly more than half the rates found previously for Indiana and Pittsburgh coals. Removal of pyrite from Pittsburgh coal by ferric ion oxidation slows markedly as ferrous ions accumulate in solution, requiring maintenance of high redox potentials in processes designed for removal of pyrite and hazardous air pollutant (HAP) precursors by circulation of ferric solutions through coal. The pyrite oxidation rates obtained in these tests were used by Unifield Engineering to support the conceptual designs for alternative pyrite and HAP precursor bioleaching processes for the phase 2 pilot plant. Thermophilic microorganisms were tested to determine if mercury could be mobilized from coal under elevated growth temperatures. There was no evidence for mercury removal from coal under these conditions. However, the activity of the organisms may have liberated mercury physically. It is also possible that the organisms dissolved mercury and it readsorbed to the clay preferentially. Both of these possibilities are undergoing further testing. The Idaho National Engineering and Environmental Laboratory?s (INEEL) slurry column reactor was operated and several batches of feed coal, product coal, waste solids and leach solutions were submitted to LBL for HAP precursor analysis. Results to date indicate significant removal of mercury, arsenic and other HAP precursors in the combined physical-biological process.

Gregory J. Olson

1997-05-12T23:59:59.000Z

154

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

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

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

155

Removal of Chromium from Aqueous Solutions by Treatment with Carbon Aerogel Electrodes Using Response Surface Methodology  

Science Journals Connector (OSTI)

Removal of Chromium from Aqueous Solutions by Treatment with Carbon Aerogel Electrodes Using Response Surface Methodology ... In the past few years, various researchers have used different materials as electrodes in the electrochemical reactors and filters. ...

Parul Rana-Madaria; Mohan Nagarajan; Chitra Rajagopal; Bhagwan S. Garg

2005-07-23T23:59:59.000Z

156

Preparation and selection of Fe-Cu sorbent for COS removal in syngas  

Science Journals Connector (OSTI)

A series of iron-based sorbents prepared with iron trioxide hydrate, cupric oxide by a novel method was studied in a fixed-bed reactor for COS removal from syngas at moderate temperature. In addition, the sorbent...

Bowu Cheng; Zhaofei Cao; Yong Bai…

2010-12-01T23:59:59.000Z

157

Removal of RDX and HMX from an artificial groundwater by granular activated carbon  

E-Print Network (OSTI)

completely mixed batch reactor (CMBR) system was selected for all rate and isotherm experiments. A number of rate and isotherm experiments were conducted to measure performance in the removal of RDX and HMX using GAC depending on dissolved oxygen, natural...

Im, Jeong Ran

1999-01-01T23:59:59.000Z

158

Sub–inertial dynamics of density–driven flows in a continuously stratified fluid on a sloping bottom. II. Isolated eddies and radiating cold domes  

Science Journals Connector (OSTI)

...eddies and radiating cold domes Francis J. Poulin Gordon E. Swaters Applied Mathematics...and radiating cold domes By Francis J. Poulin and Gordon E. Swaters Applied Mathematics...Introduction This paper is a continuation of Poulin & Swaters (1999), hereafter referred...

1999-01-01T23:59:59.000Z

159

Light Water Reactors Technology Development - Nuclear Reactors  

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

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

160

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

SciTech Connect

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

A. B. Culp

2007-01-26T23:59:59.000Z

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

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

162

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

163

Novel Catalytic Membrane Reactors  

SciTech Connect

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

164

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

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

165

Improve reformer operation with trace sulfur removal  

SciTech Connect

Modern bimetallic reforming catalysts typically have feed specifications for sulfur of 0.5 to 1 wppm in the reformer naphtha carge. Sulfur in the raw naphtha is reduced to this level by naphtha hydrotreating. While most naphtha hydrotreating operations can usually obtain these levels without substantial problems. It is difficult to obtain levels much below 0.5 to 1 wppm with this process. Revamp of a constrained existing hydrotreater to reduce product sulfur slightly can be extremely costly typically entailing replacement or addition of a new reactor. At Engelhard the authors demonstrated that if the last traces of sulfur remaining from hydrotreating can be removed, the resulting ultra-low sulfur feed greatly improves the reformer operation and provides substantial economic benefit to the refiner. Removal of the remaining trace sulfur is accomplished in a simple manner with a special adsorbent bed, without adding complexity to the reforming operation.

McClung, R.G.; Novak, W.J.

1987-01-01T23:59:59.000Z

166

Photocatalytic reactor  

DOE Patents (OSTI)

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

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

1999-01-19T23:59:59.000Z

167

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

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

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

168

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

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

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

169

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

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

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

170

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

SciTech Connect

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

171

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

SciTech Connect

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

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

2013-09-10T23:59:59.000Z

172

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

SciTech Connect

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

Friedmann, S J; Stamp, V

2005-11-01T23:59:59.000Z

173

Intrapleural Fluid Infusion for MR-Guided High-Intensity Focused Ultrasound Ablation in the Liver Dome  

Science Journals Connector (OSTI)

Rationale and Objectives Magnetic resonance–guided high-intensity focused ultrasound (MR-HIFU) ablation of tumors in the liver dome is challenging because of the presence of air in the costophrenic angle. In this study, we used a porcine liver model and a clinical MR-HIFU system to assess the feasibility and safety of using intrapleural fluid infusion (IPI) to create an acoustic window for MR-HIFU ablation in the liver dome. Materials and Methods Healthy adult Dalland land pigs (n = 6) under general anesthesia were used with animal committee approval. Degassed saline (200–800 mL) was infused into the intrapleural space under ultrasound guidance. A clinical 1.5-T MR-HIFU system was used to perform sonications (4-mm treatment cells, 300–450 W, 20–30 seconds) in the liver dome under real-time MR thermometry. An intercostal firing technique was used to prevent rib heating in one experiment. Technical success was defined as a temperature increase (>10°C) in the target area. After termination, the animal was examined for thermal damage to liver, diaphragm, pleura, lung, or intercostal muscle. Results An acoustic window was established in all animals. A temperature increase in the target area was achieved in all animals (max. 47°C–67°C). MR thermometry showed no heating outside the target area. Intercostal firing effectively reduced rib heating (55°C vs. 42°C). Postmortem examination revealed no unwanted thermal damage. One complication occurred, in the first experiment, because of an ill-suited needle (displacement of the needle). Conclusions The results indicate that IPI may be used safely to assist MR-HIFU ablation of tumors in the liver dome. For reliable tissue coagulation, IPI must be combined with an intercostal sonication technique. Considering the proportion of patients with tumors in the liver dome, IPI widens the applicability of MR-HIFU ablation for liver tumors considerably.

Joost W. Wijlemans; Martijn de Greef; Gerald Schubert; Chrit T.W. Moonen; Maurice A.A.J. van den Bosch; Mario Ries

2014-01-01T23:59:59.000Z

174

Fluid flow in the resurgent dome of Long Valley Caldera: implications from thermal data and deep electrical sounding  

Science Journals Connector (OSTI)

Temperatures of 100°C are measured at 3 km depth in a well located on the resurgent dome in the center of Long Valley Caldera, California, despite an assumed >800°C magma chamber at 6–8 km depth. Local downflow of cold meteoric water as a process for cooling the resurgent dome is ruled out by a Peclét-number analysis of temperature logs. These analyses reveal zones with fluid circulation at the upper and lower boundaries of the Bishop Tuff, and an upflow zone in the metasedimentary rocks. Vertical Darcy velocities range from 10 to 70 cm a?1. A 21-km-long geoelectrical profile across the caldera provides resistivity values to the order of 100 to >103 ?m down to a depth of 6 km, as well as variations of self-potential. Interpretation of the electrical data with respect to hydrothermal fluid movement confirms that there is no downflow beneath the resurgent dome. To explain the unexpectedly low temperatures in the resurgent dome, we challenge the common view that the caldera as a whole is a regime of high temperatures and the resurgent dome is a local cold anomaly. Instead, we suggest that the caldera was cooled to normal thermal conditions by vigorous hydrothermal activity in the past, and that a present-day hot water flow system is responsible for local hot anomalies, such as Hot Creek and the area of the Casa Diablo geothermal power plant. The source of hot water has been associated with recent shallow intrusions into the West Moat. The focus of planning for future power plants should be to locate this present-day flow system instead of relying on heat from the old magma chamber.

Daniel F.C Pribnow; Claudia Schütze; Suzanne J Hurter; Christina Flechsig; John H Sass

2003-01-01T23:59:59.000Z

175

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

176

Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment |  

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

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

177

Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment |  

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

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

178

Variance Decomposition Sensitivity Analysis of a Passive Residual Heat Removal System Model  

E-Print Network (OSTI)

for improved reliability and safety. Sensitivity analysis can provide relevant insights on the responseVariance Decomposition Sensitivity Analysis of a Passive Residual Heat Removal System Model YU Yua Removal system (RHRs) in the High Temperature Gas-Cooled Reactor (HTGR). Keywords: Uncertainty

Paris-Sud XI, Université de

179

Neutrino mass hierarchy from nuclear reactor experiments  

Science Journals Connector (OSTI)

Ten years from now reactor neutrino experiments will attempt to determine which neutrino mass eigenstate is the most massive. In this paper 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 ?, a degeneracy prevents NO?A and T2K from determining either ? 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

2013-08-29T23:59:59.000Z

180

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

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

GEN-IV Reactors  

Science Journals Connector (OSTI)

Generation-IV reactors are a set of nuclear reactors currently being developed under international collaborations targeting ... economics, proliferation resistance, and physical protection of nuclear energy. Nuclear

Taek K. Kim

2013-01-01T23:59:59.000Z

182

The Netherlands Reactor Centre  

Science Journals Connector (OSTI)

... Two illustrated brochures in English have recently J. been issued by the Netherlands Reactor Centre ( ... Centre (Reactor Centrum Nederland). The first* gives a general survey of the ...

S. WEINTROUB

1964-04-04T23:59:59.000Z

183

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

184

Flexible Conversion Ratio Fast Reactor Systems Evaluation  

SciTech Connect

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

185

Condensate removal device  

DOE Patents (OSTI)

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

Maddox, James W. (Newport News, VA); Berger, David D. (Alexandria, VA)

1984-01-01T23:59:59.000Z

186

SRS Small Modular Reactors  

SciTech Connect

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

None

2012-04-27T23:59:59.000Z

187

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

188

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

SciTech Connect

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

189

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

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

190

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

191

DOE Removes Brookhaven Contractor  

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

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

192

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

193

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

194

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

195

Tatum Dome field study report and monitoring data analysis: A supplemental report  

SciTech Connect

This report is a supplement to the Water Resources Center, Desert Research Institute report, DOE/NV/10384-03, ``Tatum Dome Field Study Report and Monitoring Data Analysis,`` Water Resources Center Publication No. 45044. The field study was initiated during the Spring of 1984 because of persistent tritium concentrations in the Surficial Aquifer determined from observed annual water samples from the series of hydrologic monitoring holes (HMH). An anomalous increase in tritium concentrations in monthly water samples from some of the hydrologic monitoring holes was also observed during the Spring of 1984 by the State of Mississippi, Division of Radiological Health. This Spring increase in tritium concentrations may well have been present earlier, but was not recognized because monthly tritium concentration data were not collected prior to 1984. It is hypothesized that groundwater in the Surficial Aquifer is made up of two layers. The older and deeper water within the Surficial Aquifer contains tritium contamination. The shallower water, infiltrating from recent precipitation, is essentially tritium free. These waters do not naturally mix completely and are only significantly mixed in the hydrologic monitoring holes by the sampling procedure. The quantity of shallow infiltrating precipitation available for mixing varies inversely with the rate of evapotranspiration. Since dispersive mixing along the boundary between the two waters also occurs as a result of groundwater movement, the average concentration of tritium in the Surficial Aquifer is decreased by dilution as well as radioactive decay.

Fenske, P.R.

1989-09-01T23:59:59.000Z

196

Tatum Dome field study report and monitoring data analysis: A supplemental report  

SciTech Connect

This report is a supplement to the Water Resources Center, Desert Research Institute report, DOE/NV/10384-03, Tatum Dome Field Study Report and Monitoring Data Analysis,'' Water Resources Center Publication No. 45044. The field study was initiated during the Spring of 1984 because of persistent tritium concentrations in the Surficial Aquifer determined from observed annual water samples from the series of hydrologic monitoring holes (HMH). An anomalous increase in tritium concentrations in monthly water samples from some of the hydrologic monitoring holes was also observed during the Spring of 1984 by the State of Mississippi, Division of Radiological Health. This Spring increase in tritium concentrations may well have been present earlier, but was not recognized because monthly tritium concentration data were not collected prior to 1984. It is hypothesized that groundwater in the Surficial Aquifer is made up of two layers. The older and deeper water within the Surficial Aquifer contains tritium contamination. The shallower water, infiltrating from recent precipitation, is essentially tritium free. These waters do not naturally mix completely and are only significantly mixed in the hydrologic monitoring holes by the sampling procedure. The quantity of shallow infiltrating precipitation available for mixing varies inversely with the rate of evapotranspiration. Since dispersive mixing along the boundary between the two waters also occurs as a result of groundwater movement, the average concentration of tritium in the Surficial Aquifer is decreased by dilution as well as radioactive decay.

Fenske, P.R.

1989-09-01T23:59:59.000Z

197

Characterization of cement from a well at Teapot Dome Oil Field: Implications for geological sequestration  

Science Journals Connector (OSTI)

Wellbores represent the weakest link in terms of CO2 storage permanence. As a result, special attention to the numerous existing wells that perforate storage formations is needed. The pre-injection condition of the cement can influence the rate (and type) of alteration by the injected CO2 plume. The condition of the existing well cement depends on a variety of factors including wellbore/formation and wellbore/brine interactions as well as the composition and type of cement placed in the well (i.e. type of admixtures used, water/solids ratio, sulfate resistant mixes, etc.). In this paper, the details of recovering wellbore cement from an older well to determine pre-injection seal integrity are described. Petrographical and chemical analyses are presented for samples of cement that were retrieved from a 19-year-old well at Teapot Dome in Wyoming. Examination revealed that the retrieved cement had altered as a result of original slurry composition and with respect to the local downhole wellbore environment. Although samples were obtained from a single well, significant differences were observed in their alteration and condition. Sulfate attack resulted in abundant ettringite formation in a cement sample taken adjacent to the Wall Creek sandstone (3060 ft), while cement taken adjacent to the Tensleep formation (5478 ft) was decalcified and enriched in magnesium, owing to reaction of calcium hydroxide in the cement with the dolomitic formation.

George W. Scherer; Barbara Kutchko; Niels Thaulow; Andrew Duguid; Bryant Mook

2011-01-01T23:59:59.000Z

198

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

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

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

199

Coal hydrogenation and deashing in ebullated bed catalytic reactor  

DOE Patents (OSTI)

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

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

1983-01-01T23:59:59.000Z

200

Design options for a bunsen reactor.  

SciTech Connect

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

Moore, Robert Charles

2013-10-01T23:59:59.000Z

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

The integral fast reactor fuel cycle  

SciTech Connect

The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management.

Chang, Y.I. (Argonne National Lab., IL (United States))

1990-01-01T23:59:59.000Z

202

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

SciTech Connect

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

Wittenbrock, N. G.

1982-01-01T23:59:59.000Z

203

Ordered Vertex Removal Subgraph Problems  

E-Print Network (OSTI)

of the vertex removal and subgraph problems are shown to be P­complete. In addition, a natural lex­ icographicOrdered Vertex Removal and Subgraph Problems Ray Greenlaw Department of Computer Science University­8703196. #12; Vertex Removal and Graph Problems Ray Greenlaw Department of Computer Science FR­35

Greenlaw, Ray

204

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

205

The TMI defueling project fuel debris removal system  

SciTech Connect

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

206

Sodium removal process development for LMFBR fuel subassemblies  

SciTech Connect

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

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

1981-10-01T23:59:59.000Z

207

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

208

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

209

Elementary Reactor Physics  

Science Journals Connector (OSTI)

... THERE are few subjects which have developed at the rate at which reactor physics and ... physics and reactor theory have done. This, of course, is largely due to the circumstances in ...

J. F. HILL

1962-02-10T23:59:59.000Z

210

Colliding Beam Fusion Reactors  

Science Journals Connector (OSTI)

The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker–Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are ... the rec...

Norman Rostoker; Artan Qerushi; Michl Binderbauer

2003-06-01T23:59:59.000Z

211

Prospects for spheromak fusion reactors  

Science Journals Connector (OSTI)

The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on...

T. K. Fowler; D. D. Hua

1995-06-01T23:59:59.000Z

212

Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system  

SciTech Connect

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 RVACS can be characterized as a surface-based decay heat removal system. It 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 the core volume) and decay heat removal capability (proportional to the vessel surface area). 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 environmental 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 or annular regions formed 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 among 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 density and therefore the reactor power can be significantly increased, without losing the passive heat removal feature. This paper introduces the concept of using DRACS to enhance VHTR passive safety and economics. Three design options with different cooling pipe locations are discussed. Analysis results from a lumped volume based model and CFD simulations are presented. (authors)

Zhao, H.; Zhang, H.; Zou, L. [Idaho National Laboratory (United States); Sun, X. [Ohio State Univ. (United States)

2012-07-01T23:59:59.000Z

213

Advanced Test Reactor Tour  

SciTech Connect

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

214

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

215

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

216

Photoactivated metal removal  

SciTech Connect

The authors propose the use of photochromic dyes as light activated switches to bind and release metal ions. This process, which can be driven by solar energy, can be used in environmental and industrial processes to remove metals from organic and aqueous solutions. Because the metals can be released from the ligands when irradiated with visible light, regeneration of the ligands and concentration of the metals may be easier than with conventional ion exchange resins. Thus, the process has the potential to be less expensive than currently used metal extraction techniques. In this paper, the authors report on their studies of the metal binding of spirogyran dyes and the hydrolytic stability of these dyes. They have prepared a number of spirogyrans and measured their binding constants for calcium and magnesium. They discuss the relationship of the structure of the dyes to their binding strengths. These studies are necessary towards determining the viability of this technique.

Nimlos, M.R.; Filley, J.; Ibrahim, M.A.; Watt, A.S.; Blake, D.M.

1999-07-01T23:59:59.000Z

217

An Approach to Mapping of Shallow Petroleum Reservoirs Using Integrated Conventional 3D and Shallow P- and SH-Wave Seismic Reflection Methods at Teapot Dome Field in Casper, Wyoming.  

E-Print Network (OSTI)

??Using the famous Teapot Dome oil field in Casper, Wyoming, USA as a test case, we demonstrate how high-resolution compressional (P) and horizontally polarized shear… (more)

Okojie-Ayoro, Anita Onohuome 1981-

2007-01-01T23:59:59.000Z

218

The Effect of Flow Rate of Very Dilute Sulfuric Acid on Xylan, Lignin, and Total Mass Removal from Corn Stover  

E-Print Network (OSTI)

The Effect of Flow Rate of Very Dilute Sulfuric Acid on Xylan, Lignin, and Total Mass Removal from mass, xylan, and lignin and increases cellulose digestibility compared to batch operations at otherwise in corn stover at 180 °C. A flow rate of 10 mL/min in a 3.8-mL reactor enhanced xylan removal by about 25

California at Riverside, University of

219

Removal of BPA by enzyme polymerization using NF membranes  

Science Journals Connector (OSTI)

Abstract The application of laccase and peroxidase from horseradish (HRP) to facilitate the removal of bisphenol A (BPA) from aqueous solutions was investigated. Effect of pH and the enzyme dose was evaluated in order to determine the optimum conditions for the enzyme performance. The results indicate that BPA was quickly removed from aqueous solution since a BPA conversion over 95% was obtained in 180 min for both enzymes in optimal conditions; the higher the enzyme dose, the higher the removal percentage of BPA. It was also found that the optimum pH for the removal efficiency of BPA was around 7 for both enzymes. The use of a membrane-reactor integrated system with recycling of enzyme for BPA degradation is also presented. These results demonstrate the potential and limitations of using enzymatic BPA degradation, operated in a recycling mode coupled to a nanofiltration membrane. BPA removal efficiencies for several NF membranes were related to the BPA molecular weight, membrane pore sizes and membrane hydrophobicity. NF270 showed the best performance in membrane-assisted enzyme treatment: 89% removal of BPA for the two enzyme treatments and less than 35% flux decay were observed.

Ivonne Escalona; Joris de Grooth; Josep Font; Kitty Nijmeijer

2014-01-01T23:59:59.000Z

220

Ozone removal by HVAC filters  

Science Journals Connector (OSTI)

Residential and commercial HVAC filters that have been loaded with particles during operation in the field can remove ozone from intake or recirculated air. However, knowledge of the relative importance of HVAC filters as a removal mechanism for ozone in residential and commercial buildings is incomplete. We measured the ozone removal efficiencies of clean (unused) fiberglass, clean synthetic filters, and field-loaded residential and commercial filters in a controlled laboratory setting. For most filters, the ozone removal efficiency declined rapidly but converged to a non-zero (steady-state) value. This steady-state ozone removal efficiency varied from 0% to 9% for clean filters. The mean steady-state ozone removal efficiencies for loaded residential and commercial filters were 10% and 41%, respectively. Repeated exposure of filters to ozone following a 24-h period of no exposure led to a regeneration of ozone removal efficiency. Based on a theoretical scaling analysis of mechanisms that are involved in the ozone removal process, we speculate that the steady-state ozone removal efficiency is limited by reactant diffusion out of particles, and that regeneration is due to internal diffusion of reactive species to sites available to ozone for reaction. Finally, by applying our results to a screening model for typical residential and commercial buildings, HVAC filters were estimated to contribute 22% and 95%, respectively, of total ozone removal in HVAC systems.

P. Zhao; J.A. Siegel; R.L. Corsi

2007-01-01T23:59:59.000Z

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


221

The Argonaut Reactor - Reactors designed/built by Argonne National  

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

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

222

Mineralogic and isotopic constraints on the origin of strontium-rich cap rock, Tatum Dome, Mississippi, U.S.A.  

Science Journals Connector (OSTI)

The limestone portion of the salt dome cap rock at Tatum dome, Mississippi, is composed of an upper massive zone and a lower banded zone. The upper zone consists of equigranular fine-crystalline calcite with veinlets and disseminations of carbonaceous matter associated with minor amounts of detrital and authigenic quartz, sulfides and Sr minerals. The lower zone is composed of alternating light and dark calcite bands. The dark bands are composed of fine-grained and peloidal calcite, quartz, bitumen and disseminated sulfide minerals. The lighter bands consist of variable proportions of generally coarse-crystalline euhedral calcite, celestite and strontianite resulting in Sr contents of up to 30% locally. Solubility data for celestite, strontianite, calcite and anhydrite suggest that a decrease in temperature favors the replacement of Ca minerals by Sr minerals, which is consistent with the observed mineral textures and paragenesis. However, the source of cap rock Sr is difficult to determine. Anhydrite at Tatum dome contains 800 ppm Sr, but the abundance of Sr minerals in the limestone cap rock and the 87Sr86Sr ratios of limestone cap rock minerals require a Sr source other than local anhydrite. Sr released by the replacement of anhydrite by calcite is capable of producing a molar ratio of celestite to calcite of only ? 0.001, yet locally this ratio is ? 3. A likely source of additional Sr is oilfield brines, such as those in central Mississippi that contain up to 3000 ppm Sr along with significant Pb and Zn. Episodic introduction of brines into the cap rock in combination with the action of sulfate-reducing bacteria probably caused the sequential production of the dark bands. The lighter bands probably reflect introduction of a later Sr-enriched fluid or evolution of the original fluid in combination with changes in the chemical parameters controlling mineral precipitation. Calcite was replaced by Sr minerals during this later Sr-rich event.

James A. Saunders; James D. Prikryl; Harry H. Posey

1988-01-01T23:59:59.000Z

223

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

224

Bacterial Colonization of Pellet Softening Reactors Used during Drinking Water Treatment  

Science Journals Connector (OSTI)

...reactor biomass concentrations as high as 220 mg of ATP/m3 of reactor...were removed as a reusable product. High calcium and magnesium concentrations...such as scale deposits in water boilers, a higher demand for detergents in washing...

Frederik Hammes; Nico Boon; Marius Vital; Petra Ross; Aleksandra Magic-Knezev; Marco Dignum

2010-12-10T23:59:59.000Z

225

In-situ gamma-ray site characterization of the Tatum Salt Dome Test Site in Lamar County, Mississippi  

SciTech Connect

Field surveys of gamma-ray emitting nuclides and soil core sampling were conducted at 12 sites on the Tatum Salt Dome Test Site and surrounding control areas to determine exposure rates from surficial radioactivity. 137Cs was the only man-made radionuclide detected and was most abundant at three off-site locations on cultivated lawns. 137Cs inventories at all of the on-site survey locations were lower than expected, given the high annual precipitation in the area. The vertical distributions were more extended than those reported for undisturbed sites. Pressurized ion chamber measurements indicated no significant differences in exposure rates on and off the test site.

Faller, S.H. (Environmental Protection Agency, Las Vegas, NV (United States))

1992-06-01T23:59:59.000Z

226

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

227

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

228

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

229

Removing Arsenic from Drinking Water  

SciTech Connect

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

None

2011-01-01T23:59:59.000Z

230

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

SciTech Connect

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

231

Mercury and tritium removal from DOE waste oils  

SciTech Connect

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

232

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

233

Removal patterns of some alkyl phenyl ketones in a heterogeneous bacterial system  

E-Print Network (OSTI)

never lrave been wri t ten, I can only say thank you. TABLE OF CONTENTS pacCe ABSTRACT ACKNONLEDGEMENTS iLIST OF TABLES LIST OF FIGURES SECTION I Introduction and Literature Review SECTION II Apparatus and Materials Reactor Gas Chromatograph... Removal by Air Stripping Organic Component Analysis Suspended Solids pH of Reactor Fluids Diffused Air Flow Rate Temperature of Reactor Fluids SECTION IV Results Test A Test B Test C Test D Test E Test F Test G Test H Test. I Test R...

Salitros, James Joseph

2012-06-07T23:59:59.000Z

234

Generation -IV Reactor Concepts  

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

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

235

Power Reactor Progress  

Science Journals Connector (OSTI)

Argonne kicks off EBWR; Allis-Chalmers plans power reactor using both nuclear and conventional fuels ... NUCLEAR POWER took two giant steps last week. ... Just as the first nuclear power system in the U. S. designed and built solely for the generation of electric power went into full operation at Argonne, Allis-Chalmers came up with a new twist in power reactors—a controlled recirculation boiling reactor (CRBR) using both nuclear and conventional fuels (C&EN, Feb. 18, page 7). ...

1957-02-25T23:59:59.000Z

236

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

237

AEC Pushes Fusion Reactors  

Science Journals Connector (OSTI)

AEC Pushes Fusion Reactors ... Project Sherwood, as the study program is called, began in 1951-52 soon after the first successful thermonuclear explosion in the Pacific. ...

1955-10-10T23:59:59.000Z

238

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

239

Advanced Nuclear Research Reactor  

SciTech Connect

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

Lolich, J.V.

2004-10-06T23:59:59.000Z

240

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

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

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

DOE Patents (OSTI)

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

Walsh, John V. (Glendora, CA)

1987-12-15T23:59:59.000Z

242

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....

243

Portfolio for fast reactor collaboration  

SciTech Connect

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

244

RCS pressure under reduced inventory conditions following a loss of residual heat removal  

SciTech Connect

The thermal-hydraulic response of a closed-reactor coolant system to loss of residual heat removal (RHR) cooling is investigated. The processes examined include: core coolant boiling and steam generator reflux condensation, pressure increase on the primary side, heat transfer mechanisms on the steam generator primary and secondary sides, and effects of noncondensible gas on heat transfer processes.

Palmrose, D.E.; Hughes, E.D.; Johnsen, G.W.

1992-08-01T23:59:59.000Z

245

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

246

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

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

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

247

ADVANCES IN HEXAVALENT CHROMIUM REMOVAL AT HANFORD  

SciTech Connect

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

NESHEM DO; RIDDELLE J

2012-01-30T23:59:59.000Z

248

Handbook of Reactor Physics  

Science Journals Connector (OSTI)

... THIS handbook is one volume in a series sponsored by the United States Atomic Energy Commission with ... data and reference information in the field of reactors. The volume is devoted to reactor physics and radiation shielding, the latter subject occupying approximately a quarter of the book.

PETER W. MUMMERY

1956-08-25T23:59:59.000Z

249

Fast reactor safety  

Science Journals Connector (OSTI)

... SIR, - In his article on fast reactor safety (26 July, page 270) Norman Dombey claims to introduce to non-specialists ... , page 270) Norman Dombey claims to introduce to non-specialists some features of fast reactors that are not available outside the technical literature. The non-specialist would do well ...

R.D. SMITH

1979-08-23T23:59:59.000Z

250

Instrumentation of Nuclear Reactors  

Science Journals Connector (OSTI)

... s Lecture Theatre on January 8, a symposium of papers on the instrumentation of nuclear reactors was organized, at which about five hundred members and visitors attended, including guests from ... the Institution, took the chair and introduced Sir John Cockcroft, whose lecture on "Nuclear Reactors and their Applications" provided a general background for the three specialized papers which followed. ...

1953-03-07T23:59:59.000Z

251

Nuclear Research Reactors  

Science Journals Connector (OSTI)

... their countries for the advent of nuclear power. A few countries had built large research reactors for the production of isotopes and to study the behaviour of nuclear fuel, but ... production of isotopes and to study the behaviour of nuclear fuel, but the small training reactor had not been developed. Since then, research ...

T. E. ALLIBONE

1963-07-20T23:59:59.000Z

252

Improved catalyst loading reduces guard reactor fouling  

SciTech Connect

A new catalyst-loading strategy reduced the fouling tendency of the gas oil hydrotreater guard reactors at Syncrude Canada Ltd.'s heavy-crude upgrading facilities. Studies conducted on the guard reactors were designed to determine the thermal stability of the coker gas oil and to understand the properties of the fouling material. Small particles (described as fines) were present in the upper section of the removed catalyst bed. This part of the bed was then replaced in one of three ways. One way was to replace the catalyst with used, nonregenerated catalyst, and cover the catalyst with nonactive support balls, 10 and 13 mm in diameter. The second way was to fill the entire space with nonactive support balls, and the third way was to fill with regenerated oxidic catalyst combined with semiactive support balls (unsulfided).

Sanford, E.C.; Kirchen, R.P. (Syncrude Canada Ltd., Edmonton (CA))

1988-12-19T23:59:59.000Z

253

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab  

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

UPTON, N.Y. - American Recovery and Reinvestment Act UPTON, N.Y. - American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once used to absorb neutrons produced for research. The Brookhaven National Laboratory is using $39 million from the Recovery Act to decommission the Brookhaven Graphite Research Reactor, the world's first reactor built solely for peaceful research purposes. The decommissioning is slated for completion later this year and will end Office of Environmental Management legacy cleanup activities at the Lab. The neutron shields were located on the north and south sides of a 700-ton graphite pile. The three-inch-thick shields absorbed neutrons that escaped from the graphite pile. The shields also limited movement of the pile when the reactor was in opera-

254

A future for nuclear energy: pebble bed reactors  

Science Journals Connector (OSTI)

Pebble Bed Reactors could allow nuclear plants to support the goal of reducing global climate change in an energy hungry world. They are small, modular, inherently safe, use a demonstrated nuclear technology and can be competitive with fossil fuels. Pebble bed reactors are helium cooled reactors that use small tennis ball size fuel balls consisting of only 9 grams of uranium per pebble to provide a low power density reactor. The low power density and large graphite core provide inherent safety features such that the peak temperature reached even under the complete loss of coolant accident without any active emergency core cooling system is significantly below the temperature that the fuel melts. This feature should enhance public confidence in this nuclear technology. With advanced modularity principles, it is expected that this type of design and assembly could lower the cost of new nuclear plants removing a major impediment to deployment.

Andrew C. Kadak

2005-01-01T23:59:59.000Z

255

The Neutrino Mass Hierarchy from Nuclear Reactor Experiments  

E-Print Network (OSTI)

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

Emilio Ciuffoli; Jarah Evslin; Xinmin Zhang

2013-02-04T23:59:59.000Z

256

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.

257

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

258

Canadian university research reactors  

SciTech Connect

In Canada there are seven university research reactors: one medium-power (2-MW) swimming pool reactor at McMaster University and six low-power (20-kW) SLOWPOKE reactors at Dalhousie University, Ecole Polytechnique, the Royal Military College, the University of Toronto, the University of Saskatchewan, and the University of Alberta. This paper describes primarily the McMaster Nuclear Reactor (MNR), which operates on a wider scale than the SLOWPOKE reactors. The MNR has over a hundred user groups and is a very broad-based tool. The main applications are in the following areas: (1) neutron activation analysis (NAA); (2) isotope production; (3) neutron beam research; (4) nuclear engineering; (5) neutron radiography; and (6) nuclear physics.

Ernst, P.C.; Collins, M.F.

1989-11-01T23:59:59.000Z

259

Reactor & Nuclear Systems Publications | ORNL  

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

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

260

Effects of catalysts and additives on fluorocarbon removal with surface discharge plasma  

Science Journals Connector (OSTI)

The decomposition of fluorocarbons (1%) in Ar was investigated using a surface discharge-type plasma reactor. To enhance the effectiveness of plasma chemical processing, we investigated the effects of catalysts and additives. The removal rate increased when the plasma reactor was packed with TiO2 pellets as a catalyst. The catalytic effect seems to be derived from the direct activation of the TiO2 surface by the plasma discharge, because significant UV emission and temperature increase were not observed in the plasma reactor. The removal rate was also enhanced when water vapor, oxygen, or hydrogen was added to the reactant. In addition, the presence of TiO2 or the additional gases or both depressed byproduct formation. Analysis of the reaction products suggested that the catalysts and additives enhanced the process by preventing the recombination of decomposed fragments.

Atsushi Ogata; Hyun-Ha Kim; Shigeru Futamura; Satoshi Kushiyama; Koichi Mizuno

2004-01-01T23:59:59.000Z

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.


261

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

262

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

263

Microsoft Word - EC Sodium coolant removal.doc  

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

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

264

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

265

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

266

Reactor Safety Research Programs  

SciTech Connect

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

267

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

268

Recommendation 199: Recommendation to Remove Uncontaminated Areas...  

Office of Environmental Management (EM)

9: Recommendation to Remove Uncontaminated Areas of the Oak Ridge Reservation from the National Priorities List Recommendation 199: Recommendation to Remove Uncontaminated Areas of...

269

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

SciTech Connect

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

270

Reactor hot spot analysis  

SciTech Connect

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

Vilim, R.B.

1985-08-01T23:59:59.000Z

271

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

272

Removing Stains from Washable Fabrics.  

E-Print Network (OSTI)

of May 8, 1914, as amended, and June 30, 1914, in cooperation with the United States Department of Agriculture. Zerle L. Carpenter, Director, Texas Agricultural Extension Service, The Texas A&M University System. lOM-1l-88, New CLO ...I UUL. Z TA24S.7 8873 NO.1616 B.1616 / Texas Agricultural Extension Service LIBRARY FEB 0 1 1989 Texas A&M University Removing Stains from Washable Fabrics Ann Vanderpoorten 8eard* Most spots and stains can be removed by prompt...

Beard, Ann Vanderpoorten

1988-01-01T23:59:59.000Z

273

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

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

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

274

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

275

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

SciTech Connect

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

276

Multipollutant Removal with WOWClean® System  

E-Print Network (OSTI)

such as petcoke, coal, wood, diesel and natural gas. In addition to significant removal of CO2, test results demonstrate the capability to reduce 99.5% SOx (from levels as high as 2200 ppm), 90% reduction of NOx, and > 90% heavy metals. The paper will include...

Romero, M.

2010-01-01T23:59:59.000Z

277

The thermal regime in the resurgent dome of Long Valley Caldera, California: Inferences from precision temperature logs in deep wells  

Science Journals Connector (OSTI)

Long Valley Caldera in eastern California formed 0.76 Ma ago in a cataclysmic eruption that resulted in the deposition of 600 km3 of Bishop Tuff. The total current heat flow from the caldera floor is estimated to be ~ 290 MW, and a geothermal power plant in Casa Diablo on the flanks of the resurgent dome (RD) generates ~40 MWe. The RD in the center of the caldera was uplifted by ~ 80 cm between 1980 and 1999 and was explained by most models as a response to magma intrusion into the shallow crust. This unrest has led to extensive research on geothermal resources and volcanic hazards in the caldera. Here we present results from precise, high-resolution, temperature–depth profiles in five deep boreholes (327–1,158 m) on the RD to assess its thermal state, and more specifically 1) to provide bounds on the advective heat transport as a guide for future geothermal exploration, 2) to provide constraints on the occurrence of magma at shallow crustal depths, and 3) to provide a baseline for future transient thermal phenomena in response to large earthquakes, volcanic activity, or geothermal production. The temperature profiles display substantial non-linearity within each profile and variability between the different profiles. All profiles display significant temperature reversals with depth and temperature gradients Valley boreholes are at the approximate same elevation as the high-temperature unit in borehole M-1 in Casa Diablo indicating lateral or sub-lateral hydrothermal flow through the resurgent dome. Small differences in temperature between measurements in consecutive years in three of the wells suggest slow cooling of the shallow hydrothermal flow system. By matching theoretical curves to segments of the measured temperature profiles, we calculate horizontal groundwater velocities in the hydrothermal flow unit under the RD that range from 1.9 to 2.8 m/yr, which corresponds to a maximum power flowing through the RD of 3–4 MW. The relatively low temperatures and large isothermal segments at the bottom of the temperature profiles are inconsistent with the presence of magma at shallow crustal levels.

Shaul Hurwitz; Christopher D. Farrar; Colin F. Williams

2010-01-01T23:59:59.000Z

278

REACTOR GROUT THERMAL PROPERTIES  

SciTech Connect

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

279

Reactor Safety Planning for Prometheus Project, for Naval Reactors Information  

SciTech Connect

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

P. Delmolino

2005-05-06T23:59:59.000Z

280

Savannah River Site | Department of Energy  

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

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

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

B Reactor | Department of Energy  

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

Operational Management » History » Manhattan Project » Signature Operational Management » History » Manhattan Project » Signature Facilities » B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first large-scale plutonium production reactor. As at Oak Ridge, the need for labor turned Hanford into an atomic boomtown, with the population reaching 50,000 by summer 1944. Similar to the X-10 Graphite Reactor at Oak Ridge in terms of loading and unloading fuel, the B Reactor was built on a much larger scale and used water rather than air as a coolant. Whereas the X-10 had an initial design output of 1,000 kilowatts, the B Reactor was designed to operate at 250,000 kilowatts. Consisting of a 28- by 36-foot, 1,200-ton graphite cylinder lying on its side, the reactor was penetrated through its

282

Heat exchanger for reactor core and the like  

DOE Patents (OSTI)

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

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

1986-01-01T23:59:59.000Z

283

2011sr01.doc  

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

Tuesday, February 1, 2011 Tuesday, February 1, 2011 james-r.giusti@srs.gov Paivi Nettamo, SRNS, (803) 646-6075 paivi.nettamo@srs.gov Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project AIKEN, S.C. - The landscape of the Savannah River Site (SRS) is a little flatter and a little less colorful with the removal today of the 75-foot-tall rusty-orange dome from the Cold War-era test reactor. This $25-million reactor

284

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  

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

285

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

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

to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile November 7, 2005 - 12:38pm Addthis Will Be Redirected to Naval Reactors, Down-blended or Used for Space Programs WASHINGTON, DC - Secretary of Energy Samuel W. Bodman today announced that the Department of Energy's (DOE) National Nuclear Security Administration (NNSA) will remove up to 200 metric tons (MT) of Highly Enriched Uranium (HEU), in the coming decades, from further use as fissile material in U.S. nuclear weapons and prepare this material for other uses. Secretary Bodman made this announcement while addressing the 2005 Carnegie International Nonproliferation Conference in Washington, DC.

286

Stratigraphic and structural analysis of Shannon Sandstone, Teapot Dome field: implications for secondary and tertiary oil recovery  

SciTech Connect

The Department of Energy and Lawrence-Allison Associates have initiated three enhanced oil recovery (EOR) pilot projects in the last 4 years in the Shannon Sandstone at the Teapot Dome field. Performance of these pilot projects has generally been poor. As a result, a reevaluation of the geology for the entire field and for the pilot areas was conducted in an attempt to explain the pilots' performances. Based on core descriptions, conceptual reservoir flow patterns, and post-fireflood coring, only the bar-margin facies was found amenable to fluid displacement processes for oil recovery. This results in 25 million bbl of oil originally in place vs 180 million bbl of oil originally in place as a target for EOR. Stratigraphy alone does not explain the observed production patterns and simulation of the reservoir. Faulting and fracturing are extensive. Based on faults mapped in the in-situ pilot area, 12 to 16 northeast-trending normal faults per mile can be projected. Fracture orientations were obtained by mapping calcareous streaks in the upper and lower Shannon. These orientations confirm directions of premature fluid breakthroughs observed in the pilot projects and in the old East Teapot waterflood. Water resistivities and total dissolved solids measurements, water-cut maps, and daily oil production maps suggest that some faults are partial to total barriers to fluid flow across the field.

Chappelle, H.; Emsurak, G.; Obernyer, S.

1986-08-01T23:59:59.000Z

287

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

SciTech Connect

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

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

2012-01-23T23:59:59.000Z

288

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

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

289

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

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

290

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

291

Thermionic Reactor Design Studies  

SciTech Connect

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

292

Removing Barriers to Interdisciplinary Research  

E-Print Network (OSTI)

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

Naomi Jacobs; Martyn Amos

2010-12-19T23:59:59.000Z

293

Reactor and Material Supply | Y-12 National Security Complex  

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

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

294

Diagnostics for hybrid reactors  

SciTech Connect

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

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

2012-06-19T23:59:59.000Z

295

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

SciTech Connect

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

296

Structural materials for fusion reactors  

Science Journals Connector (OSTI)

Fusion Reactors will require specially engineered structural materials, which ... on safety considerations. The fundamental differences between fusion and other nuclear reactors arise due to the 14MeV neutronics ...

P. M. Raole; S. P. Deshpande

2009-04-01T23:59:59.000Z

297

Inferences on the hydrothermal system beneath the resurgent dome in Long Valley Caldera, east-central California, USA, from recent pumping tests and geochemical sampling  

Science Journals Connector (OSTI)

Quaternary volcanic unrest has provided heat for episodic hydrothermal circulation in the Long Valley caldera, including the present-day hydrothermal system, which has been active over the past 40 kyr. The most recent period of crustal unrest in this region of east-central California began around 1980 and has included periods of intense seismicity and ground deformation. Uplift totaling more than 0.7 m has been centered on the caldera’s resurgent dome, and is best modeled by a near-vertical ellipsoidal source centered at depths of 6–7 km. Modeling of both deformation and microgravity data now suggests that (1) there are two inflation sources beneath the caldera, a shallower source 7–10 km beneath the resurgent dome and a deeper source ?15 km beneath the caldera’s south moat and (2) the shallower source may contain components of magmatic brine and gas. The Long Valley Exploration Well (LVEW), completed in 1998 on the resurgent dome, penetrates to a depth of 3 km directly above this shallower source, but bottoms in a zone of 100°C fluid with zero vertical thermal gradient. Although these results preclude extrapolations of temperatures at depths below 3 km, other information obtained from flow tests and fluid sampling at this well indicates the presence of magmatic volatiles and fault-related permeability within the metamorphic basement rocks underlying the volcanic fill. In this paper, we present recently acquired data from LVEW and compare them with information from other drill holes and thermal springs in Long Valley to delineate the likely flow paths and fluid system properties under the resurgent dome. Additional information from mineralogical assemblages in core obtained from fracture zones in LVEW documents a previous period of more vigorous and energetic fluid circulation beneath the resurgent dome. Although this system apparently died off as a result of mineral deposition and cooling (and/or deepening) of magmatic heat sources, flow testing and tidal analyses of LVEW water level data show that relatively high permeability and strain sensitivity still exist in the steeply dipping principal fracture zone penetrated at a depth of 2.6 km. The hydraulic properties of this zone would allow a pressure change induced at distances of several kilometers below the well to be observable within a matter of days. This indicates that continuous fluid pressure monitoring in the well could provide direct evidence of future intrusions of magma or high-temperature fluids at depths of 5–7 km.

Christopher D. Farrar; Michael L. Sorey; Evelyn Roeloffs; Devin L. Galloway; James F. Howle; Ronald Jacobson

2003-01-01T23:59:59.000Z

298

Fast pyrolysis of a waste fraction of high impact polystyrene (HIPS) containing brominated flame retardants in a fluidized bed reactor: The effects of various Ca-based additives (CaO, Ca(OH)2 and oyster shells) on the removal of bromine  

Science Journals Connector (OSTI)

A waste fraction of high impact polystyrene (HIPS) containing brominated flame retardants and antimony trioxide (Sb2O3) as a synergist was pyrolyzed in a bench-scale system equipped with a fluidized bed and char separation system. Experiments were carried out to observe the effects of the reaction temperature and three additives (CaO, Ca(OH)2, oyster shells) on the removal of bromine. An analysis of the pyrolysis oils obtained showed the oils were mainly composed of toluene, ethyl-benzene, styrene, cumene, ?-methylstyrene, 1,3-diphenylpropane, 1,3-diphenylbutane and (1-bromoethyl)-benzene. When the Ca-based additives were used, the concentration of styrene was markedly increased; whereas, those of ethlybenzene and cumene were reduced. The total bromine content of pyrolysis oil produced without any additive at 459 °C was about 5 wt.%. When Ca(OH)2 and oyster shells were applied, the total bromine contents of the pyrolysis oils were decreased to 1.3 and 2.7 wt.%, respectively. The antimony content in the pyrolysis oil was relatively small due to the efficient operation of the char separation system.

Su-Hwa Jung; Seon-Jin Kim; Joo-Sik Kim

2012-01-01T23:59:59.000Z

299

Reactor Materials | Department of Energy  

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

Benefits Crosscutting Technology Development Reactor Materials Advanced Sensors and Instrumentation Proliferation and Terrorism Risk Assessment Advanced Methods for Manufacturing...

300

Reactor operation safety information document  

SciTech Connect

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

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

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

302

Thermal Reactor Safety  

SciTech Connect

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

303

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

304

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

SciTech Connect

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

305

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

306

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....

307

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....

308

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....

309

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...

310

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...

311

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...

312

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....

313

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....

314

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....

315

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....

316

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....

317

Ukraine 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 Ukraine HEU Removal Ukraine HEU Removal Location Ukraine United States 50 12' 24.8688" N,...

318

Japan HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Jobs Apply for Our Jobs Our Jobs Working at NNSA Blog Home content Four-Year Plan Japan HEU Removal Japan HEU Removal Location Japan United States 37 36' 59.5872" N, 140...

319

Remove Condensate with Minimal Air Loss  

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

This tip sheet outlines several condensate removal methods as part of maintaining compressed air system air quality.

320

Foreign Research Reactor Spent Nuclear Fuel Acceptance Program  

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

Global Threat Reduction Initiative: Global Threat Reduction Initiative: U.S. Nuclear Remove Program Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance 2007 DOE TEC Meeting Chuck Messick DOE/NNSA/SRS 2 Contents * Program Objective and Policy * Program implementation status * Shipment Information * Operational Logistics * Lessons Learned * Conclusion 3 U.S. Nuclear Remove Program Objective * To play a key role in the Global Threat Reduction Remove Program supporting permanent threat reduction by accepting program eligible material. * Works in conjunction with the Global Threat Reduction Convert Program to accept program eligible material as an incentive to core conversion providing a disposition path for HEU and LEU during the life of the Acceptance Program. 4 Reasons for the Policy

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

Laser-based coatings removal  

SciTech Connect

Over the years as building and equipment surfaces became contaminated with low levels of uranium or plutonium dust, coats of paint were applied to stabilize the contaminants in place. Most of the earlier paint used was lead-based paint. More recently, various non-lead-based paints, such as two-part epoxy, are used. For D & D (decontamination and decommissioning), it is desirable to remove the paints or other coatings rather than having to tear down and dispose of the entire building.

Freiwald, J.G.; Freiwald, D.

1995-12-01T23:59:59.000Z

322

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

323

Fusion reactor systems  

Science Journals Connector (OSTI)

In this review we consider deuterium-tritium (D-T) fusion reactors based on four different plasma-confinement and heating approaches: the tokamak, the theta-pinch, the magnetic-mirror, and the laser-pellet system. We begin with a discussion of the dynamics of reacting plasmas and basic considerations of reactor power balance. The essential plasma physical aspects of each system are summarized, and the main characteristics of the corresponding conceptual power plants are described. In tokamak reactors the plasma densities are about 1020 m-3, and the ? values (ratio of plasma pressure to confining magnetic pressure) are approximately 5%. Plasma burning times are of the order of 100-1000 sec. Large superconducting dc magnets furnish the toroidal magnetic field, and 2-m thick blankets and shields prevent heat deposition in the superconductor. Radially diffusing plasma is diverted away from the first wall by means of null singularities in the poloidal (or transverse) component of the confining magnetic field. The toroidal theta-pinch reactor has a much smaller minor diameter and a much larger major diameter, and operates on a 10-sec cycle with 0.1-sec burning pulses. It utilizes shock heating from high-voltage sources and adabatic-compression heating powered by low-voltage, pulsed cryogenic magnetic or inertial energy stores, outside the reactor core. The plasma has a density of about 1022 m-3 and ? values of nearly unity. In the power balance of the reactor, direct-conversion energy obtained by expansion of the burning high-? plasma against the containing magnetic field is an important factor. No divertor is necessary since neutral-gas flow cools and replaces the "spent" plasma between pulses. The open-ended mirror reactor uses both thermal conversion of neutron energy and direct conversion of end-loss plasma energy to dc electrical power. A fraction of this direct-convertor power is then fed back to the ioninjection system to sustain the reaction and maintain the plasma. The average ion energy is 600 keV, plasma diameter 6 m, and the plasma beta 85%. The power levels of the three magnetic-confinement devices are in the 500-2000 MWe range, with the exception of the mirror reactor, for which the output is approximately 200 MWe. In Laser-Pellet reactors, frozen D-T pellets are ignited in a cavity which absorbs the electromagnetic, charged particle, and neutron energy from the fusion reaction. The confinement is "inertial," since the fusion reaction occurs during the disassembly of the heated pellet. A pellet-cavity unit would produce about 200 MWt in pulses with a repetition rate of the order of 10 sec-1. Such units could be clustered to give power plants with outputs in the range of 1000 MWe.

F. L. Ribe

1975-01-01T23:59:59.000Z

324

Migration and retention of elements at the Oklo natural reactor  

SciTech Connect

The Oklo natural reactor, Gabon, permits study of fission-produced elemental behavior in a natural geologic environment. The uranium ore that sustained fission reactions formed about 2 billion years before present (BYBP), and the reactor was operative for about 5 x 10/sup 5/ yrs between about 1.95 to 2 BYBP. The many tons of fission products can, for the most part, be studied for their abundance and distribution today. Since reactor shutdown, many fissiogenic elements have not migrated from host pitchblende, and several others have migrated only a few tens of meters from the reactor ore. Only Xe and Kr have apparently been largely removed from the reactor zones. An element by element assessment of the Oklo rocks' ability to retain the fission products, and actinides and radiogenic Pb and Bi as well, leads to the conclusion that no widespread migration of the elements occurred. This suggests that rocks with more favorable geologic characteristics are indeed well suited for consideration for the storage of radioactive waste.

Brookins, D.G.

1982-01-01T23:59:59.000Z

325

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

SciTech Connect

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

326

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

327

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

328

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

329

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

330

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

331

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

332

Nitrogen removal from natural gas  

SciTech Connect

According to a 1991 Energy Information Administration estimate, U.S. reserves of natural gas are about 165 trillion cubic feet (TCF). To meet the long-term demand for natural gas, new gas fields from these reserves will have to be developed. Gas Research Institute studies reveal that 14% (or about 19 TCF) of known reserves in the United States are subquality due to high nitrogen content. Nitrogen-contaminated natural gas has a low Btu value and must be upgraded by removing the nitrogen. In response to the problem, the Department of Energy is seeking innovative, efficient nitrogen-removal methods. Membrane processes have been considered for natural gas denitrogenation. The challenge, not yet overcome, is to develop membranes with the required nitrogen/methane separation characteristics. Our calculations show that a methane-permeable membrane with a methane/nitrogen selectivity of 4 to 6 would make denitrogenation by a membrane process viable. The objective of Phase I of this project was to show that membranes with this target selectivity can be developed, and that the economics of the process based on these membranes would be competitive. Gas permeation measurements with membranes prepared from two rubbery polymers and a superglassy polymer showed that two of these materials had the target selectivity of 4 to 6 when operated at temperatures below - 20{degrees}C. An economic analysis showed that a process based on these membranes is competitive with other technologies for small streams containing less than 10% nitrogen. Hybrid designs combining membranes with other technologies are suitable for high-flow, higher-nitrogen-content streams.

NONE

1997-04-01T23:59:59.000Z

333

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

334

Nuclear Archeology for CANDU Power Reactors  

SciTech Connect

The goal of this work is the development of so-called 'nuclear archeology' techniques to predict the irradiation history of both fuel-related and non-fuel-related materials irradiated in the CANDU (CANada Deuterium Uranium) family of nuclear reactors. In this application to CANDU-type reactors, two different scenarios for the collection of the appropriate data for use in these procedures will be assumed: the first scenario is the removal of the pressure tubes, calandria tubes, or fuel cladding and destructive analysis of the activation products contained in these structural materials; the second scenario is the nondestructive analysis (NDA) of the same hardware items via high-resolution gamma ray scans. There are obvious advantages and disadvantages for each approach; however, the NDA approach is the central focus of this work because of its simplicity and lack of invasiveness. The use of these techniques along with a previously developed inverse capability is expected to allow for the prediction of average flux levels and irradiation time, and the total fluence for samples where the values of selected isotopes can be measured.

Broadhead, Bryan L [ORNL] [ORNL

2011-01-01T23:59:59.000Z

335

Thermionic Reactor Design Studies  

SciTech Connect

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

336

Actinide burning in the integral fast reactor  

SciTech Connect

During the past few years, Argonne National Laboratory has been developing the integral fast reactor (IFR), an advanced liquid-metal reactor concept. In the IFR, the inherent properties of liquid-metal cooling are combined with a new metallic fuel and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics, and waste management. A key feature of the IFR concept is its unique pyroprocessing. Pyroprocessing has the potential to radically improve long-term waste management strategies by exploiting the following attributes: 1. Minor actinides accompany plutonium product stream; therefore, actinide recycling occurs naturally. Actinides, the primary source of long-term radiological toxicity, are removed from the waste stream and returned to the reactor for in situ burning, generating useful energy. 2. High-level waste volume from pyroprocessing call be reduced substantially as compared with direct disposal of spent fuel. 3. Decay heat loading in the repository can be reduced by a large factor, especially for the long-term burden. 4. Low-level waste generation is minimal. 5. Troublesome fission products, such as [sup 99]Tc, [sup 129]I, and [sup 14]C, are contained and immobilized. Singly or in combination, the foregoing attributes provide important improvements in long-term waste management in terms of the ease in meeting technical performance requirements (perhaps even the feasibility of demonstrating that technical performance requirements can be met) and perhaps also in ultimate public acceptance. Actinide recycling, if successfully developed, could well help the current repository program by providing an opportunity to enhance capacity utilization and by deferring the need for future repositories. It also represents a viable technical backup option in the event unforeseen difficulties arise in the repository licensing process.

Chang, Y.I. (Argonne National Lab., IL (United States))

1993-01-01T23:59:59.000Z

337

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

SciTech Connect

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

338

Detailed compositional analysis of gas seepage at the National Carbon Storage Test Site, Teapot Dome, Wyoming, USA  

Science Journals Connector (OSTI)

A baseline determination of CO2 and CH4 fluxes and soil gas concentrations of CO2 and CH4 was made over the Teapot Dome oil field in the Naval Petroleum Reserve No. 3 (NPR-3) in Wyoming, USA. This was done in anticipation of experimentation with CO2 sequestration in the Pennsylvanian Tensleep Sandstone underlying the field at a depth of 1680 m. The baseline data were collected during the winter, 2004 in order to minimize near-surface biological activity in the soil profile. The baseline data were used to select anomalous locations that may be the result of seeping thermogenic gas, along with background locations. Five 10-m holes were drilled, 3 of which had anomalous gas microseepage, and 2 were characterized as “background.” These were equipped for nested gas sampling at depths of 10-, 5-, 3-, 2-, and 1-m depths. Methane concentrations as high as 170,000 ppmv (17%) were found, along with high concentrations of C2H6, C3H8, n-C4H10, and i-C4H10. Much smaller concentrations of C2H4 and C3H6 were observed indicating the beginning of hydrocarbon oxidation in the anomalous holes. The anomalous 10-m holes also had high concentrations of isotopically enriched CO2, indicating the oxidation of hydrocarbons. Concentrations of the gases decreased upward, as expected, indicating oxidation and transport into the atmosphere. The ancient source of the gases was confirmed by 14C determinations on CO2, with radiocarbon ages approaching 38 ka within 5 m of the surface. Modeling was used to analyze the distribution of hydrocarbons in the anomalous and background 10-m holes. Diffusion alone was not sufficient to account for the hydrocarbon concentration distributions, however the data could be fit with the addition of a consumptive reaction. First-order rate constants for methanotrophic oxidation were obtained by inverse modeling. High rates of oxidation were found, particularly near the surface in the anomalous 10-m holes, demonstrating the effectiveness of the process in the attenuation of CH4 microseepage. The results also demonstrate the importance of CH4 measurements in the planning of a monitoring and verification program for geological CO2 sequestration in sites with significant remaining hydrocarbons (i.e. spent oil reservoirs).

Ronald W. Klusman

2006-01-01T23:59:59.000Z

339

Nuclear Reactor Severe Accident Experiments  

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

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

340

Part 3: Removal Action | Department of Energy  

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

3: Removal Action 3: Removal Action Part 3: Removal Action Question: When may removal actions be initiated? Answer: Removal actions may be initiated when DOE determines that the action will prevent, minimize, stabilize, or eliminate a risk to health or the environment. The NCP specifies that the determination that a risk to health or the environment is appropriate for removal action should be based on: actual or potential exposure of humans, animals, or the food chain the presence of contained hazardous substances that pose a threat of release the threat of migration of the hazardous substances the threat of fire or explosion the availability of an appropriate Federal or State response capability [section 300.415(b)(2)]. In essence, where DOE identifies a threat of exposure to or migration of

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

Nuclear Reactor Materials and Fuels  

Science Journals Connector (OSTI)

Nuclear reactor materials and fuels can be classified into six categories: Nuclear fuel materials Nuclear clad materials Nuclear coolant materials Nuclear poison materials Nuclear moderator materials

Dr. James S. Tulenko

2012-01-01T23:59:59.000Z

342

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

343

Light Water Reactor Sustainability Newsletter  

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

344

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

SciTech Connect

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

345

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

346

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

347

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

348

Install Removable Insulation on Valves and Fittings  

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

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

349

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

350

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...

351

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...

352

Keeler-Pennwalt Wood Pole Removal  

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

natural environment. The entire remaining length of the Keeler-Pennwalt transmission line, from Keeler Substation to Structure 96, will be removed (approximately 9 miles)....

353

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

354

Dust removal in radio-frequency plasmas by a traveling potential modulation  

SciTech Connect

The dust contamination in plasma deposition processes plays a crucial role in the quality and the yield of the products. To improve the quality and the yield of plasma processing, a favorable way is to remove the dust particles actively from the plasma reactors.Our recent experiments in the striped electrode device show that a traveling plasma modulation allows for a systematic particle removal independent of the reactor size. Besides the rf powered electrode, the striped electrode device includes a segmented electrode that consists of 100 electrically insulated narrow stripes. A traveling potential profile is produced by the modulation of the voltage signals applied on the stripes. The dust particles are trapped in the potential wells and transported with the traveling of the potential profile.The particle-in-cell (PIC) simulation on the potential above the segmented electrode indicates that the traveling potential profile can be realized either by applying low-frequency (0.1-10 Hz) voltage signals with a fixed phase shift between adjacent stripes or high-frequency (10 kHz a circumflex AS 100 MHz) signals with the amplitudes modulated by a low-frequency envelope. The transportation of the dust particles is simulated with a two-dimensional molecular dynamics (MD) code with the potential profile obtained from the PIC simulation. The MD results reproduce the experimental observations successfully.This technology allows for an active removal of the contaminating particles in processing plasmas and it is independent of the reactor size. The removal velocity is controllable by adjusting the parameters for the modulation.

Li Yangfang; Jiang Ke; Thomas, Hubertus M.; Morfill, Gregor E. [Max-Planck-Institute for Extraterrestrial Physics, 85748 Garching (Germany)

2010-06-16T23:59:59.000Z

355

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Assessment of High Value Surveillance Materials Assessment of High Value Surveillance Materials Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely

356

DOE Drops Plan to Restart Reactor  

Science Journals Connector (OSTI)

...longer in flux. Hanford research reactor...decision to scrap the Hanford reactor, which...research. At public meetings, however...decision to scrap the Hanford reactor, which...research. At public meetings, however, FFTF...

Robert F. Service

2000-12-01T23:59:59.000Z

357

Operational Analysis of Multiregional Nuclear Reactor Kinetics  

Science Journals Connector (OSTI)

......Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAR H. S. HAIDAR...analytically for a multiregional nuclear reactor whose subregions are of arbitrary...Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAU H. S. HAIDAR......

NASSAR H. S. HAIDAR

1983-05-01T23:59:59.000Z

358

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

359

Novel Reactor Design for Solid Fuel Chemical Looping Combustion  

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

Novel Reactor Design for Solid Fuel Novel Reactor Design for Solid Fuel Chemical Looping Combustion Opportunity Research is active on the patent pending technology, titled "Apparatus and Method for Solid Fuel Chemical Looping Combustion." This technology is available for licensing and/or further collaborative research from the U.S. Department of Energy's National Energy Technology Laboratory. Overview The removal of CO2 from power plants is challenging because existing methods to separate CO2 from the gas mixture requires a significant fraction of the power plant output. Chemical-looping combustion (CLC) is a novel technology that utilizes a metal oxide oxygen carrier to transport oxygen to the fuel thereby avoiding direct contact between fuel and air. The use of CLC has the advantages of reducing the energy penalty while

360

Hydrothermal Processing of Macroalgal Feedstocks in Continuous-Flow Reactors  

SciTech Connect

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

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

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

362

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

363

Acid dyes removal using low cost adsorbents  

Science Journals Connector (OSTI)

Dyestuff production units and dyeing units have always had pressing need techniques that allow economical pre-treatment for colour in the effluent. The effectiveness of adsorption for dye removal from wastewaters has made it an ideal alternative to other expensive treatment options. Removal of acid green

A.H. Aydin; Y. Bulut; O. Yavuz

2004-01-01T23:59:59.000Z

364

Temperature effects on chemical reactor  

Science Journals Connector (OSTI)

In this paper we had to study some characteristics of the chemical reactors from which we can understand the reactor operation in different circumstances; from these and the most important factor that has a great effect on the reactor operation is the temperature it is a mathematical processing of a chemical problem that was already studied but it may be developed by introducing new strategies of control; in our case we deal with the analysis of a liquid?gas reactor which can make the flotation of the benzene to produce the ethylene; this type of reactors can be used in vast domains of the chemical industry especially in refinery plants where we find the oil separation and its extractions whether they are gases or liquids which become necessary for industrial technology especially in our century.

M. Azzouzi

2008-01-01T23:59:59.000Z

365

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

366

Nuclear reactors in the United States  

Science Journals Connector (OSTI)

Nuclear reactors in the United States ... A chart listing the operating and planned nuclear reactors in the United States. ... Nuclear / Radiochemistry ...

Hubert N. Alyea

1956-01-01T23:59:59.000Z

367

Advanced Reactor Research and Development Funding Opportunity...  

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

Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE)...

368

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION  

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

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

369

MOOSE simulating nuclear reactor CRUD buildup  

SciTech Connect

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

370

Advanced Reactor Technologies | Department of Energy  

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

Advanced Reactor Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies The Office of Advanced Reactor Technologies (ART) sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. The Office of Nuclear Energy (NE) will pursue these advancements through RD&D activities at the Department of Energy (DOE) national laboratories and U.S. universities, as well as through collaboration with industry and international partners. These activities will focus on advancing scientific

371

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

372

COST OF MERCURY REMOVAL IN IGCC PLANTS  

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

Cost of Mercury Removal Cost of Mercury Removal in an IGCC Plant Final Report September 2002 Prepared for: The United States Department of Energy National Energy Technology Laboratory By: Parsons Infrastructure and Technology Group Inc. Reading, Pennsylvania Pittsburgh, Pennsylvania DOE Product Manager: Gary J. Stiegel DOE Task Manager: James R. Longanbach Principal Investigators: Michael G. Klett Russell C. Maxwell Michael D. Rutkowski PARSONS The Cost of Mercury Removal in an IGCC Plant Final Report i September 2002 TABLE OF CONTENTS Section Title Page 1 Summary 1 2 Introduction 3 3 Background 4 3.1 Regulatory Initiatives 4 3.2 Mercury Removal for Conventional Coal-Fired Plants 4 3.3 Mercury Removal Experience in Gasification 5 3.4 Variability of Mercury Content in Coal 6 4 Design Considerations 7 4.1 Carbon Bed Location

373

Physics of nuclear reactor safety  

Science Journals Connector (OSTI)

Provides a concise review of the physical aspects of safety of nuclear fission reactors. It covers the developments of roughly the last decade. The introductory chapter contains an analysis of the changes in safety philosophy that are characteristic of the last decade and that have given rise to an increased importance of physical aspects because of the emphasis on passive or natural safety. The second chapter focuses on the basics of reactor safety, identifying the main risk sources and the main principles for a safe design. The third chapter concerns a systematic treatment of the physical processes that are fundamental for the properties of fission chain reacting processes and the control of those processes. Because of the rather specialized nature of the field of reactor physics, each paragraph contains a very concise description of the theory of the phenomenon under consideration, before presenting a review of the developments. Chapter 4 contains a short review of the thermal aspects of reactor safety, restricted to those aspects that are characteristic of the nuclear reactor field, because thermal hydraulics of fission reactors is not principally different from that of other physical systems. In chapter 5 the consequences of the physics treated in the preceding chapters for the dynamics and safety of actual reactors are reviewed. The systematics of the treatment is mainly based on a division of reactors into three categories according to the type of coolant, which to a large extent determines the safety properties of the reactors. The last chapter contains a physical analysis of the Chernobyl accident that occurred in 1986. The reason for an attempt to give a review of this accident, as complete as possible within the space limits set by the editors, is twofold: the Chernobyl accident is the most severe accident in history and physical properties of the reactor played a decisive role, thereby serving as an illustration of the material of the preceding chapters.

H van Dam

1992-01-01T23:59:59.000Z

374

Bioelectricity generation and microcystins removal in a blue-green algae powered microbial fuel cell  

Science Journals Connector (OSTI)

Bioelectricity production from blue-green algae was examined in a single chamber tubular microbial fuel cell (MFC). The blue-green algae powered MFC produced a maximum power density of 114 mW/m2 at a current density of 0.55 mA/m2. Coupled with the bioenergy generation, high removal efficiencies of chemical oxygen demand (COD) and nitrogen were also achieved in MFCs. Over 78.9% of total chemical oxygen demand (TCOD), 80.0% of soluble chemical oxygen demand (SCOD), 91.0% of total nitrogen (total-N) and 96.8% ammonium–nitrogen (NH3–N) were removed under closed circuit conditions in 12 days, which were much more effective than those under open circuit and anaerobic reactor conditions. Most importantly, the MFC showed great ability to remove microcystins released from blue-green algae. Over 90.7% of MC-RR and 91.1% of MC-LR were removed under closed circuit conditions (500 ?). This study showed that the MFC could provide a potential means for electricity production from blue-green algae coupling algae toxins removal.

Yong Yuan; Qing Chen; Shungui Zhou; Li Zhuang; Pei Hu

2011-01-01T23:59:59.000Z

375

13 - Generation IV reactor designs, operation and fuel cycle  

Science Journals Connector (OSTI)

Abstract: This chapter looks at Generation IV nuclear reactors, such as the very high-temperature reactor (VHTR), the supercritical water reactor (SCWR), the molten salt reactor (MSR), the sodium-cooled fast reactor (SFR), the lead-cooled fast reactor (LFR) and the gas-cooled fast reactor (GFR). Reactor designs and fuel cycles are also described.

N. Cerullo; G. Lomonaco

2012-01-01T23:59:59.000Z

376

System reliability: An example of nuclear reactor system analysis  

Science Journals Connector (OSTI)

This paper presents an overview of the three reliability investigations of a 900 \\{MWe\\} reactor residual heat removal system. Following a description of the system and its functions, the main procedures used in operational reliability analysis, based on the analysis of occurence records, are covered. The second part of the investigations covers predicted reliability, involving the use of the Markov method. It will be noted that the two types of analysis are in good agreement, the probability of system loss after two months' operation being of the order of 10?1. Additional investigation data are also given.

R. Coudray; J.M. Mattei

1984-01-01T23:59:59.000Z

377

University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor  

SciTech Connect

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

378

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

SciTech Connect

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

379

Heat pipe cooled reactors for multi-kilowatt space power supplies  

SciTech Connect

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

380

Thermal analysis and design of a passive reflux condenser for the simplified boiling water reactor  

SciTech Connect

At present, the advanced light water reactors (ALWRS) in the United States are being designed to remove reactor decay heat for a period of 72 h following a postulated loss-of-coolant accident (LOCA). The water in the pools external to the containment is evaporated or boiled off to remove the decay heat. It is presumed that the water in the pools can be replenished within 72 h through operator actions or outside assistance. Some countries in Europe require that the plant be designed to remove the reactor decay heat for a much longer duration than 72 h without external assistance. This paper presents an analysis and design of a passive heat exchanger called a reflux condenser (RC), which was considered for an ALWR-the 600-MW(electric) simplified boiling water reactor. The RC is required to condense the steam formed when the water in the pool in which the passive containment cooling system (PCCS) is immersed boils following a LOCA. The RCs are nuclear non-safety related. This paper presents steady-state performance of an RC at various outdoor air dry-bulb temperatures under still air conditions.

Bijlani, C.; Patti, F. (Burns Roe Inc., Oradell, NJ (United States)); Prasad, V. (SUNY, Stony Brook, NY (United States))

1993-01-01T23:59:59.000Z

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

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

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

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

382

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

383

Assessment of torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R[sub 0] = 6.6-8.8 m, on-axis magnetic field B[sup 0] = 4.8-7.5 T, B[sub max] (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

Lyon, J.F. (Oak Ridge National Lab., TN (United States)) [Oak Ridge National Lab., TN (United States); Painter, S.L. (Australian National Univ., Canberra, ACT (Australia)) [Australian National Univ., Canberra, ACT (Australia)

1992-12-01T23:59:59.000Z

384

Assessment of torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R{sub 0} = 6.6-8.8 m, on-axis magnetic field B{sup 0} = 4.8-7.5 T, B{sub max} (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

Lyon, J.F. [Oak Ridge National Lab., TN (United States); Painter, S.L. [Australian National Univ., Canberra, ACT (Australia)

1992-12-01T23:59:59.000Z

385

SDI: Solar Dome Instrument for Solar Irradiance Monitoring Tao Liu1, Ankur U. Kamthe1, Varick L. Erickson1, Carlos F. M. Coimbra2 and Alberto E. Cerpa1  

E-Print Network (OSTI)

SDI: Solar Dome Instrument for Solar Irradiance Monitoring Tao Liu1, Ankur U. Kamthe1, Varick L data for ground solar irradiance (direct normal and global irradiance) is a major obstacle for the de- velopment of adequate policies to promote and take advan- tage of existing solar technologies. Although

Cerpa, Alberto E.

386

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

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

387

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

388

In situ removal of contamination from soil  

DOE Patents (OSTI)

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

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

1997-10-14T23:59:59.000Z

389

2012 Annual Report Research Reactor Infrastructure Program  

SciTech Connect

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

Douglas Morrell

2012-11-01T23:59:59.000Z

390

Tritium diagnostics in a fusion reactor  

Science Journals Connector (OSTI)

Methods for controlling tritium in a fusion reactor are reviewed. The characteristic features of the...

A. I. Markin; N. I. Syromyatnikov; A. M. Belov

2010-05-01T23:59:59.000Z

391

A gas-cooled reactor surface power system  

Science Journals Connector (OSTI)

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed depending on the number of astronauts level of scientific activity and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

Ronald J. Lipinski; Steven A. Wright; Roger X. Lenard; Gary A. Harms

1999-01-01T23:59:59.000Z

392

Hungary HEU removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

removal | National Nuclear Security Administration removal | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > content > Four-Year Plan > Hungary HEU removal Hungary HEU removal Location Hungary United States 47° 11' 51.6336" N, 19° 41' 15" E See map: Google Maps Printer-friendly version Printer-friendly version Javascript is required to view this map.

393

Mexico HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Removal | National Nuclear Security Administration Removal | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment 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° 49' 55.3116" W See map: Google Maps Printer-friendly version Printer-friendly version Javascript is required to view this map.

394

Kazakhstan HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Kazakhstan HEU Removal | National Nuclear Security Administration Kazakhstan HEU Removal | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > content > Four-Year Plan > Kazakhstan HEU Removal Kazakhstan HEU Removal Location Kazakhstan United States 48° 59' 44.1492" N, 67° 3' 37.9692" E See map: Google Maps Printer-friendly version Printer-friendly version

395

Sweden Plutonium Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Sweden Plutonium Removal | National Nuclear Security Administration Sweden Plutonium Removal | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > content > Four-Year Plan > Sweden Plutonium Removal Sweden Plutonium Removal Location Sweden United States 62° 24' 4.4136" N, 15° 22' 51.096" E See map: Google Maps Printer-friendly version Printer-friendly version

396

Method of removing polychlorinated biphenyl from oil  

DOE Patents (OSTI)

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

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

1982-03-17T23:59:59.000Z

397

Part removal of 3D printed parts  

E-Print Network (OSTI)

An experimental study was performed to understand the correlation between printing parameters in the FDM 3D printing process, and the force required to remove a part from the build platform of a 3D printing using a patent ...

Peña Doll, Mateo

2014-01-01T23:59:59.000Z

398

Turkey HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Turkey HEU Removal | National Nuclear Security Administration Turkey HEU Removal | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > content > Four-Year 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 Printer-friendly version Printer-friendly version

399

Combustion synthesis continuous flow reactor  

DOE Patents (OSTI)

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

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

1998-01-01T23:59:59.000Z

400

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

1979-01-01T23:59:59.000Z

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


401

Unique features of space reactors  

SciTech Connect

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

402

Nuclear Reactors and Technology; (USA)  

SciTech Connect

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

403

Reactor physics project final report  

E-Print Network (OSTI)

This is the final report in an experimental and theoretical program to develop and apply single- and few-element methods for the determination of reactor lattice parameters. The period covered by the report is January 1, ...

Driscoll, Michael J.

1970-01-01T23:59:59.000Z

404

SULFURIC ACID REMOVAL PROCESS EVALUATION: LONG-TERM RESULTS  

SciTech Connect

The objective of this project is to demonstrate the use of alkaline reagents injected into the furnace of coal-fired boilers as a means of controlling sulfuric acid emissions. The project is being co-funded by the U.S. DOE National Energy Technology Laboratory, under Cooperative Agreement DE-FC26-99FT40718, along with EPRI, the American Electric Power Company (AEP), FirstEnergy Corp., the Tennessee Valley Authority, and Dravo Lime, Inc. Sulfuric acid controls are becoming of increasing interest to power generators with coal-fired units for a number of reasons. Sulfuric acid is a Toxic Release Inventory species and can cause a variety of plant operation problems such as air heater plugging and fouling, back-end corrosion, and plume opacity. These issues will likely be exacerbated with the retrofit of selective catalytic reduction (SCR) for NO{sub x} control on many coal-fired plants, as SCR catalysts are known to further oxidize a portion of the flue gas SO{sub 2} to SO{sub 3}. The project previously tested the effectiveness of furnace injection of four different calcium-and/or magnesium-based alkaline sorbents on full-scale utility boilers. These reagents were tested during four one- to two-week tests conducted on two FirstEnergy Bruce Mansfield Plant (BMP) units. One of the sorbents tested was a magnesium hydroxide byproduct slurry produced from a modified Thiosorbic{reg_sign} Lime wet flue gas desulfurization system. The other three sorbents are available commercially and include dolomite, pressure-hydrated dolomitic lime, and commercial magnesium hydroxide. The dolomite reagent was injected as a dry powder through out-of-service burners, while the other three reagents were injected as slurries through air-atomizing nozzles inserted through the front wall of the upper furnace, either across from the nose of the furnace or across from the pendant superheater tubes. After completing the four one- to two-week tests, the most promising sorbents were selected for longer-term (approximately 25-day) full-scale tests on two different units. The longer-term tests were conducted to confirm the effectiveness of the sorbents tested over extended operation on two different boilers, and to determine balance-of-plant impacts. The first long-term test was conducted on FirstEnergy's BMP, Unit 3, and the second test was conducted on AEP's Gavin Plant, Unit 1. The Gavin Plant testing provided an opportunity to evaluate the effects of sorbent injected into the furnace on SO{sub 3} formed across an operating SCR reactor. This report presents the results from those long-term tests. The tests determined the effectiveness of injecting commercially available magnesium hydroxide slurry (Gavin Plant) and byproduct magnesium hydroxide slurry (both Gavin Plant and BMP) for sulfuric acid control. The results show that injecting either slurry could achieve up to 70 to 75% overall sulfuric acid removal. At BMP, this overall removal was limited by the need to maintain acceptable electrostatic precipitator (ESP) particulate control performance. At Gavin Plant, the overall sulfuric acid removal was limited because the furnace injected sorbent was less effective at removing SO{sub 3} formed across the SCR system installed on the unit for NOX control than at removing SO{sub 3} formed in the furnace. The long-term tests also determined balance-of-plant impacts from slurry injection during the two tests. These include impacts on boiler back-end temperatures and pressure drops, SCR catalyst properties, ESP performance, removal of other flue gas species, and flue gas opacity. For the most part the balance-of-plant impacts were neutral to positive, although adverse effects on ESP performance became an issue during the BMP test.

Gary M. Blythe; Richard McMillan

2002-07-03T23:59:59.000Z

405

Safety and Techno-Economic Analysis of Solvent Selection for Supercritical Fischer-Tropsch Synthesis Reactors  

E-Print Network (OSTI)

of the fixed-bed reactor, among other disadvantages, is that the reaction is very exothermic, which is a concern in terms of safety hazards and also in terms of cost of heat removal. With the slurry reactor, a problem is that in the liquid media... at different lengths.4 After the reaction takes place, the amount of carbon monoxide consumed decreases and carbon dioxide is produced as a side product.9 The FTS reaction is an extremely exothermic process, which represents serious challenges...

Hamad, Natalie

2012-02-14T23:59:59.000Z

406

The integral fast reactor fuels reprocessing laboratory at Argonne National Laboratory, Illinois  

SciTech Connect

The processing of Integral Fast Reactor (IFR) metal fuel utilizes pyrochemical fuel reprocessing steps. These steps include separation of the fission products from uranium and plutonium by electrorefining in a fused salt, subsequent concentration of uranium and plutonium for reuse, removal, concentration, and packaging of the waste material. Approximately two years ago a facility became operational at Argonne National Laboratory-Illinois to establish the chemical feasibility of proposed reprocessing and consolidation processes. Sensitivity of the pyroprocessing melts to air oxidation necessitated operation in atmosphere-controlled enclosures. The Integral Fast Reactor Fuels Reprocessing Laboratory is described.

Wolson, R.D.; Tomczuk, Z.; Fischer, D.F.; Slawecki, M.A.; Miller, W.E.

1986-09-01T23:59:59.000Z

407

Applications of porous electrodes to metal-ion removal and the design of battery systems  

SciTech Connect

This dissertation treats the use of porous electrodes as electrochemical reactors for the removal of dilute metal ions. A methodology for the scale-up of porous electrodes used in battery applications is given. Removal of 4 ..mu..g Pb/cc in 1 M sulfuric acid was investigated in atmospheric and high-pressure, flow-through porous reactors. The atmospheric reactor used a reticulated vitreous carbon porous bed coated in situ with a mercury film. Best results show 98% removal of lead from the feed stream. Results are summarized in a dimensionless plot of Sherwood number vs Peclet number. High-pressure, porous-electrode experiments were performed to investigate the effect of pressure on the current efficiency. Pressures were varied up to 120 bar on electrode beds of copper or lead-coated spheres. The copper spheres showed high hydrogen evolution rates which inhibited lead deposition, even at high cathodic overpotentials. Use of lead spheres inhibited hydrogen evolution but often resulted in the formation of lead sulfate layers; these layers were difficult to reduce back to lead. Experimental data of one-dimensional porous battery electrodes are combined with a model for the current collector and cell connectors to predict ultimate specific energy and maximum specific power for complete battery systems. Discharge behavior of the plate as a whole is first presented as a function of depth of discharge. These results are combined with the voltage and weight penalties of the interconnecting bus and post, positive and negative active material, cell container, etc. to give specific results for the lithium-aluminum/iron sulfide high-temperature battery. Subject to variation is the number of positive electrodes, grid conductivity, minimum current-collector weight, and total delivered capacity. The battery can be optimized for maximum energy or power, or a compromise design may be selected.

Trost, G.G.

1983-09-01T23:59:59.000Z

408

Laser removal of sludge from steam generators  

DOE Patents (OSTI)

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

Nachbar, Henry D. (Ballston Lake, NY)

1990-01-01T23:59:59.000Z

409

Oil removal from water via adsorption  

E-Print Network (OSTI)

WILLIAM EDWARD JACOBS 1974 OIL REMOVAL FROM WATER VIA ADSORPTION A Thesis by WILLIAM EDWARD JACOBS Submitted to the Graduate College of Texas ASM University in partial fulfillment of the requirement for the degree of MASTER OF SCIENCE... December 1973 Major Subject: Civil Engineering OIL REMOVAL FROM WATER VIA ADSORPTION A Thesis by WILLIAM EDWARD JACOBS Approved as to style and content by: C airman of Committee ea o Department m er Member Memb December 1973 ABSTRACT Oil...

Jacobs, William Edward

2012-06-07T23:59:59.000Z

410

Alternate-fuel reactor studies  

SciTech Connect

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

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

1983-02-01T23:59:59.000Z

411

Solar solids reactor  

DOE Patents (OSTI)

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

Yudow, Bernard D. (Chicago, IL)

1987-01-01T23:59:59.000Z

412

Solar solids reactor  

DOE Patents (OSTI)

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

Yudow, B.D.

1986-02-24T23:59:59.000Z

413

Lead removal by using carbon nanotubes  

Science Journals Connector (OSTI)

Exposure to lead (Pb) can cause anemia, diseases of the liver and kidneys, brain damage and ultimately death. For these reasons, heavy metals must be removed as much as possible from water. The removal of Pb (II) ions from aqueous solution using carbon nanotubes (CNT) as the adsorbent was investigated. The effects of pH were studied at 25°C. Batch mode adsorption study has revealed that the removal of Pb (II) ions was maximum (85% removal) at pH 5 and achieved 83% removal at 40 mg/L of CNTs. The adsorption continuously increased in the pH range of 3-5, beyond which the adsorption could not be carried out due to the precipitation of metal. This study was also supported by characterisation of CNTs using FESEM. The characterisation suggested that at acidic condition (pH 5), the surfaces of CNTs are more aligned and well-integrated compared to CNTs at different pHs. Finally, it can be concluded that CNTs could be a potential adsorbent for the removal of Pb from wastewater.

A.A. Muataz; M. Fettouhi; A. Al-Mammum; N. Yahya

2009-01-01T23:59:59.000Z

414

Evaluation of Torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio.

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)

1994-03-01T23:59:59.000Z

415

When Do Commercial Reactors Permanently Shut Down?  

Reports and Publications (EIA)

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

2011-01-01T23:59:59.000Z

416

NO removal by reducing agents and additives in the selective non-catalytic reduction (SNCR) process  

Science Journals Connector (OSTI)

The effect of the additives on the selective non-catalytic reduction (SNCR) reaction has been determined in a three-stage laboratory scale reactor. The optimum reaction temperature is lowered and the reaction temperature window is widened with increasing concentrations of the gas additives (CO, CH4). The optimum reaction temperature is lowered and the maximum NO removal efficiency decreases with increasing the concentration of alcohol additives (CH3OH, C2H5OH). The addition of phenol lowers the optimum reaction temperature about 100–150 °C similar to that of the toluene addition. The volatile organic compounds (VOCs: C6H5OH, C7H8) can be utilized in the SNCR process to enhance NO reduction and removed at the same time. A previously proposed simple kinetic model can successfully apply the NO reduction by NH3 and the present additives.

Sang Wook Bae; Seon Ah Roh; Sang Done Kim

2006-01-01T23:59:59.000Z

417

300 GPM Solids Removal System A True Replacement for Back Flushable Powdered Filter Systems - 13607  

SciTech Connect

The EnergySolutions Solids Removal System (SRS) utilizes stainless steel cross-flow ultra-filtration (XUF) technology which allows it to reliably remove suspended solids greater than one (1) micron from liquid radwaste streams. The SRS is designed as a pre-treatment step for solids separation prior to processing through other technologies such as Ion Exchange Resin (IER) and/or Reverse Osmosis (RO), etc. Utilizing this pre-treatment approach ensures successful production of reactor grade water while 1) decreasing the amount of radioactive water being discharged to the environment; and 2) decreasing the amount of radioactive waste that must ultimately be disposed of due to the elimination of spent powdered filter media. (authors)

Ping, Mark R.; Lewis, Mark [EnergySolutions, Suite 100, Center Point II, 100 Center Point Circle, Columbia, SC 29210 (United States)] [EnergySolutions, Suite 100, Center Point II, 100 Center Point Circle, Columbia, SC 29210 (United States)

2013-07-01T23:59:59.000Z

418

Removal of \\{PAHs\\} with surfactant-enhanced soil washing: Influencing factors and removal effectiveness  

Science Journals Connector (OSTI)

PAH removal with surfactant enhanced washing was investigated through a series of laboratory tests to examine the effect of stirring speed, washing time, surfactant concentration, liquid/solid ratio, temperature, and on-and-off mode. The first four factors show significant influence on the PAH removal while the latter two do not. Total removal ratio and a new proposed parameter, solubilization percentage, are used to evaluate the effectiveness quantitatively.

Sheng Peng; Wei Wu; Jiajun Chen

2011-01-01T23:59:59.000Z

419

(Liquid metal reactor/fast breeder reactor research and development)  

SciTech Connect

The second meeting of the UJCC was held in Japan on June 6--8, 1990. The first day was devoted to presentations of the status of the US and Japanese Fast Breeder Reactor (FBR) programs and the status of specific areas of cooperative work. Briefly, the Japanese are following the FBR development program which has been in place since the 1970s. This program includes an FBR test reactor (JOYO), a pilot-scale reactor (MONJU), a demonstration-scale plant, and commercial-scale plants by about 2020. The US program has been redirected toward an actinide recycle mission using metal fuel and pyroprocessing of spent fuel to recovery both Pu and the higher actinides for return to the Liquid Metal Reactor (LMR). The second day was spent traveling from Tokyo to Tsuruga for a tour of the MONJU reactor. The tour was especially interesting. The third day was spent writing the minutes of the meeting and the return trip to Tokyo.

Homan, F.J.

1990-06-20T23:59:59.000Z

420

Sewage sludge ash to phosphorus fertiliser: Variables influencing heavy metal removal during thermochemical treatment  

SciTech Connect

The aim of this study was to improve the removal of heavy metals from sewage sludge ash by a thermochemical process. The resulting detoxified ash was intended for use as a raw material rich in phosphorus (P) for inorganic fertiliser production. The thermochemical treatment was performed in a rotary kiln where the evaporation of relevant heavy metals was enhanced by additives. The four variables investigated for process optimisation were treatment temperature, type of additive (KCl, MgCl{sub 2}) and its amount, as well as type of reactor (directly or indirectly heated rotary kiln). The removal rates of Cd, Cr, Cu, Ni, Pb, Zn and of Ca, P and Cl were investigated. The best overall removal efficiency for Cd, Cu, Pb and Zn could be found for the indirectly heated system. The type of additive was critical, since MgCl{sub 2} favours Zn- over Cu-removal, while KCl acts conversely. The use of MgCl{sub 2} caused less particle abrasion from the pellets in the kiln than KCl. In the case of the additive KCl, liquid KCl - temporarily formed in the pellets - acted as a barrier to heavy metal evaporation as long as treatment temperatures were not sufficiently high to enhance its reaction or evaporation.

Mattenberger, H.; Fraissler, G. [Austrian Bioenergy Centre GmbH, Inffeldgasse 21b, 8010 Graz (Austria); Brunner, T. [Austrian Bioenergy Centre GmbH, Inffeldgasse 21b, 8010 Graz (Austria); BIOS BIOENERGIESYSTEME GmbH, Inffeldgasse 21b, 8010 Graz (Austria)], E-mail: thomas.brunner@abc-energy.at; Herk, P. [ARP GmbH, Johann Sacklgasse 65-67, 8700 Leoben (Austria); Hermann, L. [Austrian Bioenergy Centre GmbH, Inffeldgasse 21b, 8010 Graz (Austria); ASH DEC Umwelt AG, Donaufelderstrasse 101/4/5, 1210 Wien (Austria); Obernberger, I. [Austrian Bioenergy Centre GmbH, Inffeldgasse 21b, 8010 Graz (Austria); BIOS BIOENERGIESYSTEME GmbH, Inffeldgasse 21b, 8010 Graz (Austria)

2008-12-15T23:59:59.000Z

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


421

Accident Performance of Light Water Reactor Cladding Materials  

SciTech Connect

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

Nelson, Andrew T. [Los Alamos National Laboratory

2012-07-24T23:59:59.000Z

422

Advanced Nuclear Reactors | Department of Energy  

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

Advanced Nuclear Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key structures like coolant pipes; pumps and tanks including their surrounding steel framing; and concrete containment and support structures. The Reactors Product Line within NEAMS is concerned with modeling the reactor vessel as well as those components of a complete power plant that

423

Advanced Reactor Technology Documents | Department of Energy  

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

Nuclear Reactor Technologies » Advanced Reactor Nuclear Reactor Technologies » Advanced Reactor Technologies » Advanced Reactor Technology Documents Advanced Reactor Technology Documents January 30, 2013 Advanced Reactor Concepts Technical Review Panel Report This report documents the establishment of a technical review process and the findings of the Advanced Reactor Concepts (ARC) Technical Review Panel (TRP).1 The intent of the process is to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. A goal of the process is to facilitate greater engagement between DOE and industry. The process involved establishing evaluation criteria, conducting a pilot review, soliciting concept inputs from industry entities, reviewing the concepts by TRP members and compiling the

424

Microsoft Word - power_reactors_briggs.doc  

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

Most common - Boiling Water and Pressurized Most common - Boiling Water and Pressurized Water Reactors About 80% of the world's nuclear reactors used for generating electricity are either boiling water reactors or pressurized water reactors. Of these, about 30% are boiling water reactors and 70% are pressurized water reactors. All power reactors currently in use in the United States are of these two types. Both types of reactors have been very successfully used for reliable, on-demand, emissions-free electricity generation for decades. How does a boiling water reactor work? Water flows from the bottom of the fuel to the top of the fuel, and as it moves past the fuel, it carries away the heat produced by the

425

Evaluation of Passive and Active Soot Filters for Removal of...  

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

Passive and Active Soot Filters for Removal of Particulate Emissions from Diesel Engines Evaluation of Passive and Active Soot Filters for Removal of Particulate Emissions from...

426

High Metal Removal Rate Process for Machining Difficult Materials...  

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

High Metal Removal Rate Process for Machining Difficult Materials High Metal Removal Rate Process for Machining Difficult Materials highmetalremovalprocessfactsheet.pdf More...

427

Colorado Nonhydrocarbon Gases Removed from Natural Gas (Million...  

Annual Energy Outlook 2012 (EIA)

Nonhydrocarbon Gases Removed from Natural Gas (Million Cubic Feet) Colorado Nonhydrocarbon Gases Removed from Natural Gas (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3...

428

Oak Ridge Removes Laboratory's Greatest Source of Groundwater...  

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

Removes Laboratory's Greatest Source of Groundwater Contamination Oak Ridge Removes Laboratory's Greatest Source of Groundwater Contamination May 1, 2012 - 12:00pm Addthis Workers...

429

Vehicle Technologies Office Merit Review 2014: Removing Barriers...  

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

Review 2014: Removing Barriers, Implementing Policies and Advancing Alternative Fuels Markets in New England Vehicle Technologies Office Merit Review 2014: Removing Barriers,...

430

Advanced Water Removal via Membrane Solvent Extraction | Department...  

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

Water Removal via Membrane Solvent Extraction Advanced Water Removal via Membrane Solvent Extraction advwaterremovalmse.pdf More Documents & Publications Advance Patent Waiver...

431

241-AZ-101 pump removal trough analysis  

SciTech Connect

As part of the current Hanford mission of environmental cleanup, various long length equipment must be removed from highly radioactive waste tanks. The removal of equipment will utilize portions of the Equipment Removal System for Project W320 (ERS-W320), specifically the 50 ton hydraulic trailer system. Because the ERS-W320 system was designed to accommodate much heavier equipment it is adequate to support the dead weight of the trough, carriage and related equipment for 241AZ101 pump removal project. However, the ERS-W320 components when combined with the trough and its` related components must also be analyzed for overturning due to wind loads. Two troughs were designed, one for the 20 in. diameter carriage and one for the 36 in. diameter carriage. A proposed 52 in. trough was not designed and, therefore is not included in this document. In order to fit in the ERS-W320 strongback the troughs were design with the same widths. Structurally, the only difference between the two troughs is that more material was removed from the stiffener plates on the 36 in trough. The reduction in stiffener plate material reduces the allowable load. Therefore, only the 36 in. trough was analyzed.

Coverdell, B.L.

1995-10-17T23:59:59.000Z

432

A Carbon Dioxide Gas Turbine Direct Cycle with Partial Condensation for Nuclear Reactors  

SciTech Connect

A carbon dioxide gas turbine power generation system with a partial condensation cycle has been proposed for thermal and fast nuclear reactors, in which compression is done partly in the liquid phase and partly in the gas phase. This cycle achieves higher cycle efficiency than a He direct cycle mainly due to reduced compressor work of the liquid phase and of the carbon dioxide real gas effect, especially in the vicinity of the critical point. If this cycle is applied to a thermal reactor, efficiency of this cycle is about 55% at a reactor outlet temperature of 900 deg. C and pressure of 12.5 MPa, which is higher by about 10% than a typical helium direct gas turbine cycle plant (PBMR) at 900 deg. C and 8.4 MPa; this cycle also provides comparable cycle efficiency at the moderate core outlet temperature of 600 deg. C with that of the helium cycle at 900 deg. C. If this cycle is applied to a fast reactor, it is anticipated to be an alternative to liquid metal cooled fast reactors that can provide slightly higher cycle efficiency at the same core outlet temperature; it would eliminate safety problems, simplify the heat transport system and simplify plant maintenance. A passive decay heat removal system is realized by connecting a liquid carbon dioxide storage tank with the reactor vessel and by supplying carbon dioxide gasified from the tank to the core in case of depressurization event. (authors)

Yasuyoshi Kato; Takeshi Nitawaki; Yoshio Yoshizawa [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo, 152-8550 (Japan)

2002-07-01T23:59:59.000Z

433

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

Energy Savers (EERE)

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

434

Global Optimization of Chemical Reactors and Kinetic Optimization  

E-Print Network (OSTI)

Model; 3-D; Monolith; Reactor; Optimization Introduction TheAngeles Global Optimization of Chemical Reactors and KineticGlobal Optimization of Chemical Reactors and Kinetic

ALHUSSEINI, ZAYNA ISHAQ

2013-01-01T23:59:59.000Z

435

Experimental Breeder Reactor-II Primary Tank System Wash Water Workshop  

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

Pre-Developmental Pre-Developmental INL EBR-II Wash Water Treatment Technologies (PBS # ADSHQTD0100 (0003199)) EBR-II Wash Water Workshop - The majority of the sodium has been removed, remaining material is mostly passivated. Similar closure projects have been successfully completed. Engineering needs to be developed to apply the OBA path. Page 1 of 2 Idaho National Laboratory Idaho Experimental Breeder Reactor-II Primary Tank System Wash Water Workshop Challenge In 1994 Congress ordered the shutdown of the Experimental Breeder Reactor-II (EBR-II) and a closure project was initiated. The facility was placed in cold shutdown, engineering began on sodium removal, the sodium was drained in 2001 and the residual sodium chemically passivated to render it less reactive in 2005. Since that time, approximately 700 kg of metallic sodium and 3500 kg of sodium bicarbonate remain in the facility. The

436

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

SciTech Connect

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

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

2000-02-01T23:59:59.000Z

437

UCLA program in reactor studies: The ARIES tokamak reactor study  

SciTech Connect

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

Not Available

1991-01-01T23:59:59.000Z

438

FROM CONCEPT TO REALITY, IN-SITU DECOMMISSIONING OF THE P AND R REACTORS AT THE SAVANNAH RIVER SITE  

SciTech Connect

SRS recently completed an approximately three year effort to decommission two SRS reactors: P-Reactor (Building 105-P) and R-Reactor (Building 105-R). Completed in December 2011, the concurrent decommissionings marked the completion of two relatively complex and difficult facility disposition projects at the SRS. Buildings 105-P and 105-R began operating as production reactors in the early 1950s with the mission of producing weapons material (e.g., tritium and plutonium-239). The 'P' Reactor and was shutdown in 1991 while the 'R' Reactor and was shutdown in 1964. In the intervening period between shutdown and deactivation & decommissioning (D&D), Buildings 105-P and 105-R saw limited use (e.g., storage of excess heavy water and depleted uranium oxide). For Building 105-P, deactivation was initiated in April 2007 and was essentially complete by June 2010. For Building 105-R, deactivation was initiated in August 2008 and was essentially complete by September 2010. For both buildings, the primary objective of deactivation was to remove/mitigate hazards associated with the remaining hazardous materials, and thus prepare the buildings for in-situ decommissioning. Deactivation removed the following hazardous materials to the extent practical: combustibles/flammables, residual heavy water, acids, friable asbestos (as needed to protect workers performing deactivation and decommissioning), miscellaneous chemicals, lead/brass components, Freon(reg sign), oils, mercury/PCB containing components, mold and some radiologically-contaminated equipment. In addition to the removal of hazardous materials, deactivation included the removal of hazardous energy, exterior metallic components (representing an immediate fall hazard), and historical artifacts along with the evaporation of water from the two Disassembly Basins. Finally, so as to facilitate occupancy during the subsequent in-situ decommissioning, deactivation implemented repairs to the buildings and provided temporary power.

Musall, J.; Blankenship, J.; Griffin, W.

2012-01-09T23:59:59.000Z

439

Rapid starting methanol reactor system  

DOE Patents (OSTI)

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

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

1984-01-01T23:59:59.000Z

440

Parametric Evaluation of Large-Scale High-Temperature Electrolysis Hydrogen Production Using Different Advanced Nuclear Reactor Heat Sources  

SciTech Connect

High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 °C to 950 °C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered.

Edwin A. Harvego; Michael G. McKellar; James E. O'Brien; J. Stephen Herring

2009-09-01T23:59:59.000Z

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


441

Method of making thermally removable polyurethanes  

DOE Patents (OSTI)

A method of making a thermally-removable polyurethane material by heating a mixture of a maleimide compound and a furan compound, and introducing alcohol and isocyanate functional groups, where the alcohol group and the isocyanate group reacts to form the urethane linkages and the furan compound and the maleimide compound react to form the thermally weak Diels-Alder adducts that are incorporated into the backbone of the urethane linkages during the formation of the polyurethane material at temperatures from above room temperature to less than approximately 90.degree. C. The polyurethane material can be easily removed within approximately an hour by heating to temperatures greater than approximately 90.degree. C. in a polar solvent. The polyurethane 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); McElhanon, James R. (Livermore, CA); Saunders, Randall S. (late of Albuquerque, NM); Durbin-Voss, Marvie Lou (Albuquerque, NM)

2002-01-01T23:59:59.000Z

442

Nitrate removal from drinking water -- Review  

SciTech Connect

Nitrate concentrations in surface water and especially in ground water have increased in Canada, the US, Europe, and other areas of the world. This trend has raised concern because nitrates cause methemoglobiinemia in infants. Several treatment processes including ion exchange, biological denitrification, chemical denitrification, reverse osmosis, electrodialysis, and catalytic denitrification can remove nitrates from water with varying degrees of efficiency, cost, and ease of operation. Available technical data, experience, and economics indicate that ion exchange and biological denitrification are more acceptable for nitrate removal than reverse osmosis. Ion exchange is more viable for ground water while biological denitrification is the preferred alternative for surface water. This paper reviews the developments in the field of nitrate removal processes.

Kapoor, A.; Viraraghavan, T. [Univ. of Regina, Saskatchewan (Canada)

1997-04-01T23:59:59.000Z

443

Management of Hanford Site non-defense production reactor spent nuclear fuel, Hanford Site, Richland, Washington  

SciTech Connect

The US Department of Energy (DOE) needs to provide radiologically, and industrially safe and cost-effective management of the non-defense production reactor spent nuclear fuel (SNF) at the Hanford Site. The proposed action would place the Hanford Site`s non-defense production reactor SNF in a radiologically- and industrially-safe, and passive storage condition pending final disposition. The proposed action would also reduce operational costs associated with storage of the non-defense production reactor SNF through consolidation of the SNF and through use of passive rather than active storage systems. Environmental, safety and health vulnerabilities associated with existing non-defense production reactor SNF storage facilities have been identified. DOE has determined that additional activities are required to consolidate non-defense production reactor SNF management activities at the Hanford Site, including cost-effective and safe interim storage, prior to final disposition, to enable deactivation of facilities where the SNF is now stored. Cost-effectiveness would be realized: through reduced operational costs associated with passive rather than active storage systems; removal of SNF from areas undergoing deactivation as part of the Hanford Site remediation effort; and eliminating the need to duplicate future transloading facilities at the 200 and 400 Areas. Radiologically- and industrially-safe storage would be enhanced through: (1) removal from aging facilities requiring substantial upgrades to continue safe storage; (2) utilization of passive rather than active storage systems for SNF; and (3) removal of SNF from some storage containers which have a limited remaining design life. No substantial increase in Hanford Site environmental impacts would be expected from the proposed action. Environmental impacts from postulated accident scenarios also were evaluated, and indicated that the risks associated with the proposed action would be small.

NONE

1997-03-01T23:59:59.000Z

444

N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning  

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

N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning Project Now Complete N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning Project Now Complete June 14, 2012 - 12:00pm Addthis Media Contacts Cameron Hardy Cameron.Hardy@rl.doe.gov 509-376-5365 Mark McKenna mmckenna@wch-rcc.com 509-372-9032 RICHLAND, WASH. - The U.S. Department of Energy's (DOE's) River Corridor contractor, Washington Closure Hanford, has completed placing N Reactor in interim safe storage, a process also known as "cocooning." N Reactor was the last of nine plutonium production reactors to be shut down at DOE's Hanford Site in southeastern Washington state. It was Hanford's longest-running reactor, operating from 1963 to 1987. "In the 1960's, N Reactor represented the future of energy in America.

445

Removal of 2-Aminophenol Using Novel Adsorbents  

Science Journals Connector (OSTI)

The positive values of entropy show the increased randomness at solid/solution interface with some structural changes in the adsorbate and adsorbent and the affinity of adsorbents toward 2AP. ... Upon doubling the adsorbent amount from 10 to 20 g/L, the amount of phenol adsorbed also increases by almost two-fold. ... It is quite evident that, after 6 h of equilibrium, 27% of the total 2-aminophenol is removed by 10 g/L of the adsorbent slag, while 20 g/L of slag removed 37% of 2-aminophenol and 30 g/L of adsorbent adsorbs 42% under identical experimental conditions. ...

Vinod K. Gupta; Dinesh Mohan; Suhas; Kunwar P. Singh

2006-01-04T23:59:59.000Z

446

Process for removing metals from water  

DOE Patents (OSTI)

A process for removing metals from water including the steps of prefiltering solids from the water, adjusting the pH to between about 2 and 3, reducing the amount of dissolved oxygen in the water, increasing the pH to between about 6 and 8, adding water-soluble sulfide to precipitate insoluble sulfide- and hydroxide-forming metals, adding a containing floc, and postfiltering the resultant solution. The postfiltered solution may optionally be eluted through an ion exchange resin to remove residual metal ions. 2 tabs.

Napier, J.M.; Hancher, C.M.; Hackett, G.D.

1987-06-29T23:59:59.000Z

447

Fracture characterization and fluid flow simulation with geomechanical constraints for a CO2–EOR and sequestration project Teapot Dome Oil Field, Wyoming, USA  

Science Journals Connector (OSTI)

Mature oil and gas reservoirs are attractive targets for geological sequestration of CO2 because of their potential storage capacities and the possible cost offsets from enhanced oil recovery (EOR). In this work, we analyze the fracture system of the Tensleep Formation to develop a geomechanically-constrained 3D reservoir fluid flow simulation at Teapot Dome Oil Field, WY, USA. Teapot Dome is the site of a proposed CO2-EOR and sequestration pilot project. The objective of this work is to model the migration of the injected CO2 in the fracture reservoir, as well as to obtain limits on the rates and volumes of CO2 that can be injected, without compromising seal integrity. Furthermore we want to establish the framework to design injection experiments that will provide insight into the fracture network of the reservoir, in particular of fracture permeability and connectivity. Teapot Dome is an elongated asymmetrical, basement-cored anticline with a north-northeast axis. The Tensleep Fm. in this area is highly fractured, and consists of an intercalation of eolian-dune sandstones and inter-dune deposits. The dune sandstones are permeable and porous intervals with different levels of cementation that affects their porosity, permeability, and fracture intensity. The inter-dune deposits consist of thin sabkha carbonates, minor evaporates, and thin but widespread extensive beds of very low-permeability dolomicrites. The average permeability is 30 mD, ranging from 10–100 mD. The average reservoir thickness is 50 ft. The caprock for the Tensleep Fm. consists of the Opeche Shale member, and the anhydrite of the Minnekhata member. The reservoir has strong aquifer drive. In the area under study, the Tensleep Fm. has its structural crest at 1675 m. It presents a 2-way closure trap against a NE-SW fault to the north and possibly the main thrust to the west. The CO2-EOR and sequestration project will consist of the injection of 1 million cubic feet of supercritical CO2 for six weeks. A previous geomechanical analysis suggested that the trapping faults do not appear to be at risk of reactivation and it was estimated that caprock integrity is not a risk by the buoyancy pressure of the maximum CO2 column height that the formation can hold. However, in the present study we established the presence of critically stressed minor faults and fractures in the reservoir and caprock, which if reactivated, could not only enhance the permeability of the reservoir, but potentially compromise the top seal capacity. The results of the preliminary fluid flow simulations indicate that the injected CO2 will rapidly rise to the top layers, above the main producing interval, and will accumulate in the fractures, where almost none will get into the matrix.

Laura Chiaramonte; Mark Zoback; Julio Friedmann; Vicki Stamp; Chris Zahm

2011-01-01T23:59:59.000Z

448

Graphite Reactor | ornl.gov  

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

Graphite Reactor Graphite Reactor 'In the early, desperate days of World War II, the United States launched the top-secret, top-priority Manhattan Project...' In the early, desperate days of U.S. involvement in World War II, American scientists began to fear that the German discovery of uranium fission in 1939 might enable the Nazis to develop a super bomb. Afraid of losing this crucial race, the United States launched the top-secret, top-priority Manhattan Project. The plan was to create two atomic weapons-one fueled by plutonium, the other by enriched uranium. Hanford, Washington, was selected as the site for plutonium production, but before large reactors could be built there, a pilot plant was necessary to prove the feasibility of scaling up from laboratory experiments. A secluded, rural area near Clinton, Tennessee, was

449

Business Opportunities for Small Reactors  

SciTech Connect

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

Minato, Akio; Nishimura, Satoshi [Central Research Institute of Electric Power Industry - CRIEPI, 2-11-1 Iwado-Kita, Komae, Tokyo 201-8511 (Japan); Brown, Neil W. [Lawrence Livermore National Laboratory - LLNL, PO Box 808, Livermore, CA 94551 (United States)

2007-07-01T23:59:59.000Z

450

Compact-Toroid Fusion Reactor (CTOR) based on the Field-Reversed Theta Pinch  

SciTech Connect

Scoping studies of a translating Compact Torus Reactor (CTOR) have been made on the basis of a dynamic plasma model and an overall systems approach. This CTOR embodiment uses a Field-Reversed Theta Pinch as a plasma source. The field-reversed plasmoid would be formed and compressionally heated to ignition prior to injection into and translation through a linear burn chamber, thereby removing the high-technology plamoid source from the hostile reactor environment. Stabilization of the field-reversed plasmoid would be provided by a passive conducting shell located outside the high-temperature blanket but within the low-field superconducting magnets and associated radition shielding. On the basis of this batch-burn but thermally steady-state approach, a reactor concept emerges with a length below approx. 40 m that generates 300 to 400 MWe of net electrical power with a recirculating power fraction less than 0.15.

Hagenson, R.L.; Krakowski, R.A.

1981-01-01T23:59:59.000Z

451

Actinide Burning in CANDU Reactors  

SciTech Connect

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

Hyland, B.; Dyck, G.R. [Atomic Energy of Canada Limited, Chalk River, Ontario, K0J 1J0 (Canada)

2007-07-01T23:59:59.000Z

452

Preliminary requirements for a Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR)  

SciTech Connect

A Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR) design is being developed at MIT to provide the first demonstration and test of a salt-cooled reactor using high-temperature fuel. The first step is to define the requirements. The top level requirements are (1) provide the confidence that a larger demonstration reactor is warranted and (2) develop the necessary data for a larger-scale reactor. Because requirements will drive the design of the FHTR, a significant effort is being undertaken to define requirements and understand the tradeoffs that will be required for a practical design. The preliminary requirements include specifications for design parameters and necessary tests of major reactor systems. Testing requirements include demonstration of components, systems, and procedures for refueling, instrumentation, salt temperature control to avoid coolant freezing, salt chemistry and volume control, tritium monitoring and control, and in-service inspection. Safety tests include thermal hydraulics, neutronics - including intrinsic core shutdown mechanisms such as Doppler feedback - and decay heat removal systems. Materials and coolant testing includes fuels (including mechanical wear and fatigue) and system corrosion behavior. Preliminary analysis indicates a thermal power output below 30 MW, an initial core using pebble-bed or prismatic-block fuel, peak outlet temperatures of at least 700 deg. C, and use of FLi{sup 7}Be ({sup 7}LiF-BeF{sub 2}) coolant. The option to change-out the reactor core, fuel type, and major components is being investigated. While the FHTR will be used for materials testing, its primary mission is as a reactor system performance test to enable the design and licensing of a FHR demonstration power reactor. (authors)

Massie, M.; Forsberg, C.; Forget, B. [Dept. of Nuclear Science and Engineering, Massachusetts Inst. of Technology, Cambridge, MA 02139 (United States); Hu, L. W. [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

2012-07-01T23:59:59.000Z

453

BNL | Our History: Reactors as Research Tools  

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

> See also: Accelerators > See also: Accelerators Brookhaven History: Using Reactors as Research Tools BGRR Brookhaven Graphite Research Reactor The Brookhaven Graphite Research Reactor (BGRR) was the Laboratory's first big machine and the first peace-time reactor built in the United States following World War II. The reactor's primary mission was to produce neutrons for scientific experimentation and to refine reactor technology. At the time, the BGRR could accommodate more simultaneous experiments than any other reactor. Scientists and engineers from every corner of the U.S. came to use the reactor, which was not only a source of neutrons for experiments, but also an excellent training facility. Researchers used the BGRR's neutrons as tools for studying atomic nuclei and the structure of solids, and to investigate many physical, chemical and

454

New fast-reactor approach. [LMFBR  

SciTech Connect

The design parameters for a 1000 MW LMFBR type reactor are presented. The design requires the multiple primary coolant pumps and heat exchangers to be located around the core within the reactor vessel.

Folkrod, J.R.; Kann, W.J.; Klocksieben, R.H.

1983-01-01T23:59:59.000Z

455

Reactor accelerator coupling experiments: a feasability study  

E-Print Network (OSTI)

The Reactor Accelerator Coupling Experiments (RACE) are a set of neutron source driven subcritical experiments under temperature feedback conditions. These experiments will involve coupling an accelerator driven neutron source to a TRIGA reactor...

Woddi Venkat Krishna, Taraknath

2006-08-16T23:59:59.000Z

456

Reactivity control assembly for nuclear reactor  

DOE Patents (OSTI)

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

Bollinger, Lawrence R. (Schenectady, NY)

1984-01-01T23:59:59.000Z

457

Inherent safety concepts in nuclear power reactors  

Science Journals Connector (OSTI)

Different inherent safety concepts being considered in fast and thermal reactors are presented after outlining the basic goals of nuclear reactor safety, the ‘defence in depth’ philosophy to achieve these goal...

O M Pal Singh; R Shankar Singh

1989-06-01T23:59:59.000Z

458

Choice of coils for a fusion reactor  

Science Journals Connector (OSTI)

...configurations. The most ambitious is the International Thermonuclear Experimental Reactor, a large tokamak planned for construction...configuration has features in common with the International Thermonuclear Experimental Reactor experiment. Mathematical Model We...

Romeo Alexander; Paul R. Garabedian

2007-01-01T23:59:59.000Z

459

Computational Analysis of Fluid Flow in Pebble Bed Modular Reactor  

E-Print Network (OSTI)

High Temperature Gas-cooled Reactor (HTGR) is a Generation IV reactor under consideration by Department of Energy and in the nuclear industry. There are two categories of HTGRs, namely, Pebble Bed Modular Reactor (PBMR) and Prismatic reactor. Pebble...

Gandhir, Akshay

2012-10-19T23:59:59.000Z

460

The development of structural materials for fusion reactors  

Science Journals Connector (OSTI)

...severely exposed parts of future fusion reactors and pose key problems...successful implementation of fusion reactors as an efficient source...conditions in the International Thermonuclear Experimental Reactor (ITER...environmental attractiveness of fusion reactors. In this paper...

1999-01-01T23:59:59.000Z

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.


461

Utilization of Refractory Metals and Alloys in Fusion Reactor Structures  

Science Journals Connector (OSTI)

In design of fusion reactors, structural material selection is very crucial to improve reactor’s performance. Different types of materials have been proposed for use in fusion reactor structures. Among these mate...

Mustafa Übeyli; ?enay Yalç?n

2006-12-01T23:59:59.000Z

462

Process removes Sr from nuclear wastes  

Science Journals Connector (OSTI)

Process removes Sr from nuclear wastes ... Scientists at Argonne National Laboratory have devised a chemical process for extracting and recovering strontium-90 from liquid nuclear wastes. ... Argonne chemist E. Philip Horwitz, head of the team, says it could be a significant aid in managing such radioactive wastes. ...

WARD WORTHY

1990-09-10T23:59:59.000Z

463

Removing dissolved inorganic contaminants from water  

SciTech Connect

This article describes the physicochemical treatment processes typically used to remove the more common inorganic contaminants from water and wastewater. These are precipitation, coprecipitation, adsorption, ion exchange, membrane separations by reverse osmosis and electrodialysis, and combinations of these processes. The general criteria for process selection are discussed, and the processes and their typical applications are described.

Clifford, D.; Subramonian, S.; Sorg, T.J.

1986-11-01T23:59:59.000Z

464

Method for Removing Precipitates in Biofuel  

At ORNL the application of ultrasonic energy, or sonication, has been shown to successfully remove or prevent the formation of 50–90% of the precipitates in biofuels. Precipitates can plug filters as biodiesel is transported from one location to another, and often cannot be detected by visual inspection....

2010-12-08T23:59:59.000Z

465

Automatic Red Eye Removal for Digital Photography  

E-Print Network (OSTI)

Chapter 1 Automatic Red Eye Removal for Digital Photography FRANCESCA GASPARINI DISCo, Dipartimento The red eye effect is a well known problem in photography. It is often seen in amateur shots taken with a built-in flash, but the problem is also well known to professional photographers. Red eye is the red

Schettini, Raimondo

466

Plastic bottles > Remove lids (not recyclable)  

E-Print Network (OSTI)

Plastic bottles Please: > Remove lids (not recyclable) > Empty bottles > Rinse milk bottles, & other bottles if possible > Squash bottles www.st-andrews.ac.uk/estates/environment All types of plastic bottle accepted Clear, opaque and coloured bottles Labels can remain on X No plastic bags X No plastics

Brierley, Andrew

467

ASBESTOS PIPE-INSULATION REMOVAL ROBOT SYSTEM  

SciTech Connect

This final topical report details the development, experimentation and field-testing activities for a robotic asbestos pipe-insulation removal robot system developed for use within the DOE's weapon complex as part of their ER and WM program, as well as in industrial abatement. The engineering development, regulatory compliance, cost-benefit and field-trial experiences gathered through this program are summarized.

Unknown

2000-09-15T23:59:59.000Z

468

Removable partial overdentures for the irradiated patient  

SciTech Connect

Patients who have received radiotherapy to the head and neck area must avoid dental extractions and seek simplicity in treatment and home care follow-up. For partially edentulous patients, removable partial overdenture therapy can fulfill these goals while maintaining the high level of function and aesthetics desired by patients.11 references.

Rosenberg, S.W. (New York Univ. School of Dentistry, NY (USA))

1990-10-01T23:59:59.000Z

469

Pentek metal coating removal system: Baseline report  

SciTech Connect

The Pentek coating removal technology was tested and is being evaluated at Florida International University (FIU) as a baseline technology. In conjunction with FIU`s evaluation of efficiency and cost, this report covers evaluation conducted for safety and health issues. It is a commercially available technology and has been used for various projects at locations throughout the country. The Pentek coating removal system consisted of the ROTO-PEEN Scaler, CORNER-CUTTER{reg_sign}, and VAC-PAC{reg_sign}. They are designed to remove coatings from steel, concrete, brick, and wood. The Scaler uses 3M Roto Peen tungsten carbide cutters while the CORNER-CUTTER{reg_sign} uses solid needles for descaling activities. These hand tools are used with the VAC-PAC{reg_sign} vacuum system to capture dust and debris as removal of the coating takes place. The safety and health evaluation during the testing demonstration focused on two main areas of exposure: dust and noise. Dust exposure minimal, but noise exposure was significant. Further testing for each exposure is recommended because of the environment where the testing demonstration took place. It is feasible that the dust and noise levels will be higher in an enclosed operating environment of different construction. In addition, other areas of concern found were arm-hand vibration, whole-body, ergonomics, heat stress, tripping hazards, electrical hazards, machine guarding, and lockout/tagout.

NONE

1997-07-31T23:59:59.000Z

470

Method of preparation of removable syntactic foam  

DOE Patents (OSTI)

Easily removable, environmentally safe, low-density, syntactic foams are disclosed which are prepared by mixing insoluble microballoons with a solution of water and/or alcohol-soluble polymer to produce a pourable slurry, optionally vacuum filtering the slurry in varying degrees to remove unwanted solvent and solute polymer, and drying to remove residual solvent. The properties of the foams can be controlled by the concentration and physical properties of the polymer, and by the size and properties of the microballoons. The suggested solute polymers are non-toxic and soluble in environmentally safe solvents such as water or low-molecular weight alcohols. The syntactic foams produced by this process are particularly useful in those applications where ease of removability is beneficial, and could find use in packaging recoverable electronic components, in drilling and mining applications, in building trades, in art works, in the entertainment industry for special effects, in manufacturing as temporary fixtures, in agriculture as temporary supports and containers and for delivery of fertilizer, in medicine as casts and splints, as temporary thermal barriers, as temporary protective covers for fragile objects, as filters for particulate matter, which matter may be easily recovered upon exposure to a solvent, as in-situ valves (for one-time use) which go from maximum to minimum impedance when solvent flows through, and for the automatic opening or closing of spring-loaded, mechanical switches upon exposure to a solvent, among other applications.

Arnold, Jr., Charles (Albuquerque, NM); Derzon, Dora K. (Albuquerque, NM); Nelson, Jill S. (Albuquerque, NM); Rand, Peter B. (Albuquerque, NM)

1995-01-01T23:59:59.000Z

471

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

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

Colley, Robert W. (Richland, WA)

1984-01-01T23:59:59.000Z

472

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

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

Colley, R.W.

1982-06-29T23:59:59.000Z

473

Light Water Reactor Sustainability (LWRS) Program  

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

Light Water Reactor Sustainability (LWRS) Program Login Instructions go here. User ID: Password: Log In Forgot your password?...