Powered by Deep Web Technologies
Note: This page contains sample records for the topic "transport reactor integrated" 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

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

2

Supertruck - Improving Transportation Efficiency through Integrated...  

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

Improving Transportation Efficiency through Integrated Vehicle, Engine and Powertrain Research Supertruck - Improving Transportation Efficiency through Integrated Vehicle, Engine...

3

Integrated Transportation System Design Optimization  

E-Print Network (OSTI)

Integrated Transportation System Design Optimization by Christine Taylor B.S. Cornell University by . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Professor Jaime Peraire Chairman, Department Graduate Committee #12;2 #12;Integrated Transportation System Abstract Traditionally, the design of a transportation system has focused on either the vehicle design

4

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

5

Transportation and Stationary Power Integration: Workshop Proceedings  

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

Proceedings for the Transportation and Stationary Power Integration Workshop held on October 27, 2008 in Phoenix, Arizona

6

Advanced High-Temperature, High-Pressure Transport Reactor Gasification  

SciTech Connect

The U.S. Department of Energy (DOE) National Energy Technology Laboratory Office of Coal and Environmental Systems has as its mission to develop advanced gasification-based technologies for affordable, efficient, zero-emission power generation. These advanced power systems, which are expected to produce near-zero pollutants, are an integral part of DOE's Vision 21 Program. DOE has also been developing advanced gasification systems that lower the capital and operating costs of producing syngas for chemical production. A transport reactor has shown potential to be a low-cost syngas producer compared to other gasification systems since its high-throughput-per-unit cross-sectional area reduces capital costs. This work directly supports the Power Systems Development Facility utilizing the KBR transport reactor located at the Southern Company Services Wilsonville, Alabama, site. Over 2800 hours of operation on 11 different coals ranging from bituminous to lignite along with a petroleum coke has been completed to date in the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center (EERC). The EERC has established an extensive database on the operation of these various fuels in both air-blown and oxygen-blown modes utilizing a pilot-scale transport reactor gasifier. This database has been useful in determining the effectiveness of design changes on an advanced transport reactor gasifier and for determining the performance of various feedstocks in a transport reactor. The effects of different fuel types on both gasifier performance and the operation of the hot-gas filter system have been determined. It has been demonstrated that corrected fuel gas heating values ranging from 90 to 130 Btu/scf have been achieved in air-blown mode, while heating values up to 230 Btu/scf on a dry basis have been achieved in oxygen-blown mode. Carbon conversions up to 95% have also been obtained and are highly dependent on the oxygen-coal ratio. Higher-reactivity (low-rank) coals appear to perform better in a transport reactor than the less reactive bituminous coals. Factors that affect TRDU product gas quality appear to be coal type, temperature, and air/coal ratios. Testing with a higher-ash, high-moisture, low-rank coal from the Red Hills Mine of the Mississippi Lignite Mining Company has recently been completed. Testing with the lignite coal generated a fuel gas with acceptable heating value and a high carbon conversion, although some drying of the high-moisture lignite was required before coal-feeding problems were resolved. No ash deposition or bed material agglomeration issues were encountered with this fuel. In order to better understand the coal devolatilization and cracking chemistry occurring in the riser of the transport reactor, gas and solid sampling directly from the riser and the filter outlet has been accomplished. This was done using a baseline Powder River Basin subbituminous coal from the Peabody Energy North Antelope Rochelle Mine near Gillette, Wyoming.

Michael Swanson; Daniel Laudal

2008-03-31T23:59:59.000Z

7

Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes  

SciTech Connect

This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540°C and the helium coolant was delivered at 7 MPa at 625–925°C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the hightemperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

Lee O. Nelson

2011-04-01T23:59:59.000Z

8

Device Scale Model Development for Transport Reactor  

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

Gary J. stiegel Gary J. stiegel Gasification Technology Manager National Energy Technology Laboratory 626 Cochrans Mill Road P.O. Box 10940 Pittsburgh, PA 15236 412-386-4499 gary.stiegel@netl.doe.gov Chris Guenther Computational Science Division National Energy Technology Laboratory 3610 Collins Ferry Road P. O. Box 880 Morgantown, WV 26507 304-285-4483 chris.guenther@netl.doe.gov 8/2006 Gasification Technologies Device Scale MoDel DevelopMent for tranSport reactor Background Coal gasification is an efficient and environmentally acceptable technology that can utilize the vast coal reserves in the United States to produce clean affordable power and reduce dependence on foreign oil. Coal and other carbon containing materials can be gasified to produce a synthesis gas. This syngas can be fed to a

9

The Integral Fast Reactor (IFR) - Reactors designed/built by Argonne  

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

Integral Fast Reactor Integral Fast Reactor 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

10

Integrated transportation system design optimization  

E-Print Network (OSTI)

Traditionally, the design of a transportation system has focused on either the vehicle design or the network flow, assuming the other as given. However, to define a system level architecture for a transportation system, ...

Taylor, Christine P. (Christine Pia), 1979-

2007-01-01T23:59:59.000Z

11

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

12

Integration for Seamless Transport | Open Energy Information  

Open Energy Info (EERE)

the reasons for the relative failure of integrated transport polices with particular reference to experience in the UK. LEDSGP green logo.png This tool is included in the...

13

NEW OPTIMIZATION-BASED APPROACH TO CHEMICAL REACTOR SYNTHESIS TOWARDS THE FULL INTEGRATION OF REACTOR  

E-Print Network (OSTI)

NEW OPTIMIZATION-BASED APPROACH TO CHEMICAL REACTOR SYNTHESIS ­ TOWARDS THE FULL INTEGRATION solutions. However, it does not provide optimal reactor design from both economical and environmental and methods for reactor design. It also explores the possibilities for actuation improvement for the optimal

Van den Hof, Paul

14

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

15

Hot-Gas Filter Testing with a Transport Reactor Gasifier  

SciTech Connect

Today, coal supplies over 55% of the electricity consumed in the United States and will continue to do so well into the next century. One of the technologies being developed for advanced electric power generation is an integrated gasification combined cycle (IGCC) system that converts coal to a combustible gas, cleans the gas of pollutants, and combusts the gas in a gas turbine to generate electricity. The hot exhaust from the gas turbine is used to produce steam to generate more electricity from a steam turbine cycle. The utilization of advanced hot-gas particulate and sulfur control technologies together with the combined power generation cycles make IGCC one of the cleanest and most efficient ways available to generate electric power from coal. One of the strategic objectives for U.S. Department of Energy (DOE) IGCC research and development program is to develop and demonstrate advanced gasifiers and second-generation IGCC systems. Another objective is to develop advanced hot-gas cleanup and trace contaminant control technologies. One of the more recent gasification concepts to be investigated is that of the transport reactor gasifier, which functions as a circulating fluid-bed gasifier while operating in the pneumatic transport regime of solid particle flow. This gasifier concept provides excellent solid-gas contacting of relatively small particles to promote high gasification rates and also provides the highest coal throughput per unit cross-sectional area of any other gasifier, thereby reducing capital cost of the gasification island.

Swanson, M.L.; Hajicek, D.R.

2002-09-18T23:59:59.000Z

16

A next-generation reactor concept: The Integral Fast Reactor (IFR)  

SciTech Connect

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

Chang, Y.I.

1992-01-01T23:59:59.000Z

17

A next-generation reactor concept: The Integral Fast Reactor (IFR)  

SciTech Connect

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

Chang, Y.I.

1992-07-01T23:59:59.000Z

18

Neutron transport analysis for nuclear reactor design  

DOE Patents (OSTI)

Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values.

Vujic, Jasmina L. (Lisle, IL)

1993-01-01T23:59:59.000Z

19

Light Water Reactor Sustainability Program - Integrated Program Plan |  

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

Light Water Reactor Sustainability Program - Integrated Program Light Water Reactor Sustainability Program - Integrated Program Plan Light Water Reactor Sustainability Program - Integrated Program Plan The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U. S. Department of Energy (DOE), performed in close collaboration and cooperation with related industry R&D programs. The LWRS Program provides technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants, utilizing the unique capabilities of the national laboratory system. Sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer than-initially-licensed lifetime. It has two facets

20

Advanced Reactors Thermal Energy Transport for Process Industries  

SciTech Connect

The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.

P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith

2014-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport reactor integrated" 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

The Integral Fast Reactor: A practical approach to waste management  

SciTech Connect

This report discusses development of the method for pyroprocessing of spent fuel from the Integral Fast Reactor (or Advanced Liquid Metal Reactor). The technology demonstration phase, in which recycle will be demonstrated with irradiated fuel from the EBR-II reactor has been reached. Methods for recovering actinides from spent LWR fuel are at an earlier stage of development but appear to be technically feasible at this time, and a large-scale demonstration of this process has begun. The utilization of fully compatible processes for recycling valuable spent fuel materials promises to provide substantial economic incentives for future applications of the pyroprocessing technology.

Laidler, J.J.

1993-12-31T23:59:59.000Z

22

Integrated intelligent systems in advanced reactor control rooms  

SciTech Connect

An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs., 5 figs.

Beckmeyer, R.R.

1989-01-01T23:59:59.000Z

23

Integral Fast Reactor Program annual progress report, FY 1991  

SciTech Connect

This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

Not Available

1992-06-01T23:59:59.000Z

24

Integral Fast Reactor Program annual progress report, FY 1991  

SciTech Connect

This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R D.

Not Available

1992-06-01T23:59:59.000Z

25

Numerical simulations of ion transport membrane oxy-fuel reactors for CO? capture applications  

E-Print Network (OSTI)

Numerical simulations were performed to investigate the key features of oxygen permeation and hydrocarbon conversion in ion transport membrane (ITM) reactors. ITM reactors have been suggested as a novel technology to enable ...

Hong, Jongsup

2013-01-01T23:59:59.000Z

26

Actinide recycle potential in the integral fast reactor  

SciTech Connect

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. In the IFR pyroprocessing, minor actinides accompany plutonium product stream, and therefore, actinide recycle occurs naturally. The fast neutron spectrum of the IFR makes it an ideal actinide burner, as well. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and potential implications on long-term waste management.

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

1993-12-31T23:59:59.000Z

27

Mass tracking and material accounting in the Integral Fast Reactor (IFR)  

SciTech Connect

The Integral Fast Reactor (IFR) is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory (ANL). There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure the compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities at ANL-West, utilizing Experimental Breeder Reactor 2 and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-Tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations. The components of the MTG System include: (1) an Oracle database manager with a Fortran interface, (2) a set of MTG Tasks'' which collect, manipulate and report data, (3) a set of MTG Terminal Sessions'' which provide some interactive control of the Tasks, and (4) a set of servers which manage the Tasks and which provide the communications link between the MTG System and Operator Control Stations, which control process equipment and monitoring devices within the FCF.

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01T23:59:59.000Z

28

Mass tracking and material accounting in the integral fast reactor (IFR)  

SciTech Connect

This paper reports on the Integral Fast Reactor (IFR) which is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory. There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure with compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities at ANL-West, utilizing Experimental Breeder Reactor II and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations.

Orechwa, Y.; Adams, C.H.; White, A.M. (Argonne National Lab., IL (United States). Reactor Analysis and Safety Div.)

1991-01-01T23:59:59.000Z

29

Progress and status of the integral fast reactor (IFR) development program  

SciTech Connect

This paper discusses the Integral Fast Reactor (IFR) development program, in which the entire reactor system - reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are also presented.

Chang, Y.I. (Argonne National Lab., Argonne, IL (US))

1992-01-01T23:59:59.000Z

30

Light Water Reactor Sustainability Program: Integrated Program Plan |  

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

Integrated Program Plan Integrated Program Plan Light Water Reactor Sustainability Program: Integrated Program Plan Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas- emitting electric power generation in the United States. Domestic demand for electrical energy is expected to grow by more than 30% from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license, for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power

31

(COMEDIE program review and fission product transport in MHTGR reactor)  

SciTech Connect

The subcontract between Martin Marietta Energy Systems, Inc., and the CEA provides for the refurbishment of the high pressure COMEDIE test loop in the SILOE reactor and a series of experiments to characterize fission product lift-off from MHTGR heat exchanger surfaces under several depressurization accident scenarios. The data will contribute to the validation of models and codes used to predict fission product transport in the MHTGR. In the meeting at CEA headquarters in Paris the program schedule and preparation for the DCAA and Quality Assurance audits were discussed. Long-range interest in expanded participation in the gas-cooled reactor technology Umbrella Agreement was also expressed by the CEA. At the CENG, in Grenoble, technical details on the loop design, fabrication components, development of test procedures, and preparation for the DOE quality assurance (QA) audit in May were discussed. After significant delays in CY 1989 it appears that good progress is being made in CY 1990 and the first major test will be initiated by December. An extensive list of agreements and commitments was generated to facilitate the coordination and planning of future work. 2 figs., 2 tabs.

Stansfield, O.M.

1990-03-15T23:59:59.000Z

32

TRANSPORTATION ENERGY FORECASTS FOR THE 2007 INTEGRATED ENERGY  

E-Print Network (OSTI)

for the information in this report; nor does any party represent that the uses of this information will not infringe of transportation fuel and crude oil import requirements to establish the quantitative baseline to support its fuels, integration of energy use and land use planning, and transportation fuel infrastructure

33

Transportation and Stationary Power Integration Workshop | Department...  

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

and other groups met to discuss the topic of integrating stationary fuel cell combined heat and power (CHP) systems and hydrogen production infrastructure for vehicles. The...

34

56113.Transport and integration 13.1 Transport  

E-Print Network (OSTI)

be classified according to the hazard involved: · Mirror segments. The size of each segment allows transport by laser beams. This grid defines the X-Y-Z location of each node of the structure. Each node can several light cranes with a maximum payload of 3 tons. thereby allowing for redundancy. · The structural

Liske, Jochen

35

Oxygen transport membrane system and method for transferring heat to catalytic/process reactors  

DOE Patents (OSTI)

A method and apparatus for producing heat used in a synthesis gas production is provided. The disclosed method and apparatus include a plurality of tubular oxygen transport membrane elements adapted to separate oxygen from an oxygen containing stream contacting the retentate side of the membrane elements. The permeated oxygen is combusted with a hydrogen containing synthesis gas stream contacting the permeate side of the tubular oxygen transport membrane elements thereby generating a reaction product stream and radiant heat. The present method and apparatus also includes at least one catalytic reactor containing a catalyst to promote the stream reforming reaction wherein the catalytic reactor is surrounded by the plurality of tubular oxygen transport membrane elements. The view factor between the catalytic reactor and the plurality of tubular oxygen transport membrane elements radiating heat to the catalytic reactor is greater than or equal to 0.5.

Kelly, Sean M; Kromer, Brian R; Litwin, Michael M; Rosen, Lee J; Christie, Gervase Maxwell; Wilson, Jamie R; Kosowski, Lawrence W; Robinson, Charles

2014-01-07T23:59:59.000Z

36

Modeling and Optimization of Membrane Reactors for Carbon Capture in Integrated Gasification Combined Cycle Units  

Science Journals Connector (OSTI)

Modeling and Optimization of Membrane Reactors for Carbon Capture in Integrated Gasification Combined Cycle Units ... This paper investigates the alternative of precombustion capture of carbon dioxide from integrated gasification combined cycle (IGCC) plants using membrane reactors equipped with H2-selective zeolite membranes for the water gas shift reaction. ...

Fernando V. Lima; Prodromos Daoutidis; Michael Tsapatsis; John J. Marano

2012-03-08T23:59:59.000Z

37

KINETICS OF HOT-GAS DESULFURIZATION SORBENTS FOR TRANSPORT REACTORS  

SciTech Connect

Hot-gas desulfurization for the integrated gasification combined cycle (IGCC) process has been investigated by many researchers to remove effectively hydrogen sulfide with various metal oxide sorbents at elevated temperatures. Various metal oxide sorbents are formulated with metal oxides such as Fe, Co, Zn, and Ti. Initial reaction kinetics of formulated sorbents with hydrogen sulfide is studied in the presence of various amounts of moisture and hydrogen at various reaction temperatures. The objectives of this research are to study initial reaction kinetics for a sorbent-hydrogen sulfide heterogeneous reaction system, to investigate effects of concentrations of hydrogen sulfide, hydrogen, and moisture on dynamic absorption of H{sub 2}S into sorbents, and to evaluate effects of temperature and sorbent amounts on dynamic absorption of H{sub 2}S into sorbents. Experimental data on initial reaction kinetics of hydrogen sulfide with metal oxide sorbents were obtained with a 0.83-cm{sup 3} differential reactor. In this report, the reactivity of AHI-5 was examined. This sorbent was obtained from the Research Triangle Institute (RTI). The sorbent in the form of 70 {micro}m particles are reacted with 9000-18000 ppm hydrogen sulfide at 350-500 C. The range of space time of reaction gas mixtures is 0.071-0.088 s. The range of reaction duration is 4-10800 s.

K.C. Kwon

2001-01-01T23:59:59.000Z

38

Reduced Order Model Compensator Control of Species Transport in a CVD Reactor  

E-Print Network (OSTI)

Reduced Order Model Compensator Control of Species Transport in a CVD Reactor G.M. Kepler, H for computation of feedback controls and compensators in a high pressure chemical vapor deposition (HPCVD) reactor, through open­loop optimization [6, 21, 34]. However, because of process variability and the in­ creasing

39

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

40

NREL: Transportation Research - Electric Vehicle Grid Integration  

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

Grid Integration Illustration of a house with a roof-top photovoltaic system. A wind turbine and utility towers appear in the background. A car, parked in the garage, is...

Note: This page contains sample records for the topic "transport reactor integrated" 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

VISION 2050: An Integrated National Transportation System  

E-Print Network (OSTI)

of the Federal Aviation Administration (FAA) Research, Engineering and Development Advisory Committee (REDAC advisory committees, including the FAA REDAC, NASA ASTAC, and the U.S. Department of Transportation (DOT February 2001 Copies available on these Web sites: http://scitech.dot.gov http://research.faa.gov/aar/redac

Handy, Susan L.

42

Air Quality and Intelligent Transportation Systems: Understanding Integrated Innovation, Deployment and Adaptation of Public Technologies  

E-Print Network (OSTI)

Air Quality and Intelligent Transportation Systems: Understanding Integrated Innovation, Deployment;2 Air Quality and Intelligent Transportation Systems: Understanding Integrated Innovation, Deployment between state and local transportation investments and metropolitan air quality goals. In this context

de Weck, Olivier L.

43

Advanced High-Temperature, High-Pressure Transport Reactor Gasification  

SciTech Connect

The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was 50 hours of gasification on a petroleum coke from the Hunt Oil Refinery and an additional 73 hours of operation on a high-ash coal from India. Data from these tests indicate that while acceptable fuel gas heating value was achieved with these fuels, the transport gasifier performs better on the lower-rank feedstocks because of their higher char reactivity. Comparable carbon conversions have been achieved at similar oxygen/coal ratios for both air-blown and oxygen-blown operation for each fuel; however, carbon conversion was lower for the less reactive feedstocks. While separation of fines from the feed coals is not needed with this technology, some testing has suggested that feedstocks with higher levels of fines have resulted in reduced carbon conversion, presumably due to the inability of the finer carbon particles to be captured by the cyclones. These data show that these low-rank feedstocks provided similar fuel gas heating values; however, even among the high-reactivity low-rank coals, the carbon conversion did appear to be lower for the fuels (brown coal in particular) that contained a significant amount of fines. The fuel gas under oxygen-blown operation has been higher in hydrogen and carbon dioxide concentration since the higher steam injection rate promotes the water-gas shift reaction to produce more CO{sub 2} and H{sub 2} at the expense of the CO and water vapor. However, the high water and CO{sub 2} partial pressures have also significantly reduced the reaction of (Abstract truncated)

Michael L. Swanson

2005-08-30T23:59:59.000Z

44

Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems  

SciTech Connect

Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

2011-04-06T23:59:59.000Z

45

Integrated System for Retrieval, Transportation and Consolidated Storage of Used Nuclear Fuel in the US - 13312  

SciTech Connect

The current inventory of used nuclear fuel assemblies (UNFAs) from commercial reactor operations in the United States totals approximately 65,000 metric tons or approximately 232,000 UNFAs primarily stored at the 104 operational reactors in the US and a small number of decommissioned reactors. This inventory is growing at a rate of roughly 2,000 to 2,400 metric tons each year, (Approx. 7,000 UNFAs) as a result of ongoing commercial reactor operations. Assuming an average of 10 metric tons per storage/transportation casks, this inventory of commercial UNFAs represents about 6,500 casks with an additional of about 220 casks every year. In January 2010, the Blue Ribbon Commission (BRC) [1] was directed to conduct a comprehensive review of policies for managing the back end of the nuclear fuel cycle and recommend a new plan. The BRC issued their final recommendations in January 2012. One of the main recommendations is for the United States to proceed promptly to develop one or more consolidated storage facilities (CSF) as part of an integrated, comprehensive plan for safely managing the back end of the nuclear fuel cycle. Based on its extensive experience in storage and transportation cask design, analysis, licensing, fabrication, and operations including transportation logistics, Transnuclear, Inc. (TN), an AREVA Subsidiary within the Logistics Business Unit, is engineering an integrated system that will address the complete process of commercial UNFA management. The system will deal with UNFAs in their current storage mode in various configurations, the preparation including handling and additional packaging where required and transportation of UNFAs to a CSF site, and subsequent storage, operation and maintenance at the CSF with eventual transportation to a future repository or recycling site. It is essential to proceed by steps to ensure that the system will be the most efficient and serve at best its purpose by defining: the problem to be resolved, the criteria to evaluate the solutions, and the alternative solutions. The complexity of the project is increasing with time (more fuel assemblies, new storage systems, deteriorating logistics infrastructure at some sites, etc.) but with the uncertainty on the final disposal path, flexibility and simplicity will be critical. (authors)

Bracey, William; Bondre, Jayant; Shelton, Catherine [Transnuclear, Inc., 7135 Minstrel Way Suite 300, Columbia MD 21045 (United States)] [Transnuclear, Inc., 7135 Minstrel Way Suite 300, Columbia MD 21045 (United States); Edmonds, Robert [AREVA Federal Services, 7207 IBM Drive, Charlotte NC 28262 (United States)] [AREVA Federal Services, 7207 IBM Drive, Charlotte NC 28262 (United States)

2013-07-01T23:59:59.000Z

46

Integrated transport and renewable energy systems B. V. Mathiesen*  

E-Print Network (OSTI)

, as electricity and heating. In this paper, a coherent effort to integrate transport into energy planning2 emissions, electricity and heating have traditionally been in focus. As more and more countries have been successful within electricity and heating where political focus has produced actions

47

Dynamic analysis and application of fuel elements pneumatic transportation in a pebble bed reactor  

Science Journals Connector (OSTI)

Abstract Almost 10,000 spherical fuel elements are transported pneumatically one by one in the pipeline outside the core of a pebble bed reactor every day. Any failure in the transportation will lead to the shutdown of the reactor, even safety accidents. In order to ensure a stable and reliable transportation, it's of great importance to analyze the motion and force condition of the fuel element. In this paper, we focus on the dynamic analysis of the pneumatic transportation of the fuel element and derive kinetic equations. Then we introduce the design of the transportation pipeline. On this basis we calculate some important data such as the velocity of the fuel element, the force between the fuel element and the pipeline and the efficiency of the pneumatic transportation. Then we analyze these results and provide some suggestions for the design of the pipeline. The experiment was carried out on an experimental platform. The velocities of the fuel elements were measured. The experimental results were consistent with and validated the theoretical analysis. The research may offer the basis for the design of the transportation pipeline and the optimization of the fuel elements transportation in a pebble bed reactor.

Hongbing Liu; Dong Du; Zandong Han; Yirong Zou; Jiluan Pan

2014-01-01T23:59:59.000Z

48

Parallel heat transport in integrable and chaotic magnetic fields  

SciTech Connect

The study of transport in magnetized plasmas is a problem of fundamental interest in controlled fusion, space plasmas, and astrophysics research. Three issues make this problem particularly challenging: (i) The extreme anisotropy between the parallel (i.e., along the magnetic field), {chi}{sub ||} , and the perpendicular, {chi}{sub Up-Tack }, conductivities ({chi}{sub ||} /{chi}{sub Up-Tack} may exceed 10{sup 10} in fusion plasmas); (ii) Nonlocal parallel transport in the limit of small collisionality; and (iii) Magnetic field lines chaos which in general complicates (and may preclude) the construction of magnetic field line coordinates. Motivated by these issues, we present a Lagrangian Green's function method to solve the local and non-local parallel transport equation applicable to integrable and chaotic magnetic fields in arbitrary geometry. The method avoids by construction the numerical pollution issues of grid-based algorithms. The potential of the approach is demonstrated with nontrivial applications to integrable (magnetic island), weakly chaotic (Devil's staircase), and fully chaotic magnetic field configurations. For the latter, numerical solutions of the parallel heat transport equation show that the effective radial transport, with local and non-local parallel closures, is non-diffusive, thus casting doubts on the applicability of quasilinear diffusion descriptions. General conditions for the existence of non-diffusive, multivalued flux-gradient relations in the temperature evolution are derived.

Castillo-Negrete, D. del; Chacon, L. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831-8071 (United States)

2012-05-15T23:59:59.000Z

49

Parallel heat transport in integrable and chaotic magnetic fields  

SciTech Connect

The study of transport in magnetized plasmas is a problem of fundamental interest in controlled fusion, space plasmas, and astrophysics research. Three issues make this problem particularly chal- lenging: (i) The extreme anisotropy between the parallel (i.e., along the magnetic field), , and the perpendicular, , conductivities ( / may exceed 1010 in fusion plasmas); (ii) Magnetic field lines chaos which in general complicates (and may preclude) the construction of magnetic field line coordinates; and (iii) Nonlocal parallel transport in the limit of small collisionality. Motivated by these issues, we present a Lagrangian Green s function method to solve the local and non-local parallel transport equation applicable to integrable and chaotic magnetic fields in arbitrary geom- etry. The method avoids by construction the numerical pollution issues of grid-based algorithms. The potential of the approach is demonstrated with nontrivial applications to integrable (magnetic island chain), weakly chaotic (devil s staircase), and fully chaotic magnetic field configurations. For the latter, numerical solutions of the parallel heat transport equation show that the effective radial transport, with local and non-local closures, is non-diffusive, thus casting doubts on the appropriateness of the applicability of quasilinear diffusion descriptions. General conditions for the existence of non-diffusive, multivalued flux-gradient relations in the temperature evolution are derived.

Del-Castillo-Negrete, Diego B [ORNL; Chacon, Luis [ORNL

2012-01-01T23:59:59.000Z

50

Analysis of kinetic models of the methanol-to-gasoline (MTG) process in an integral reactor  

Science Journals Connector (OSTI)

From experimental results obtained in a wide range of operating conditions (temperature and contact time) in an isothermal fixed bed integral reactor, the validity both of the kinetic models proposed in the literature as well as their modifications, for the methanol-to-gasoline (MTG) process at zero time on-stream, has been studied. The kinetic parameters for the various models have been calculated by solving the equation of mass conservation in the reactor for the lumps of the kinetic models. The usefulness of the model of Schipper and Krambeck for simulating the operation in the isothermal fixed bed integral reactor has been proven in the 573–648 K range.

Ana G. Gayubo; Pedro L. Benito; Andrés T. Aguayo; Itziar Aguirre; Javier Bilbao

1996-01-01T23:59:59.000Z

51

Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents  

SciTech Connect

This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises.

Ellison, P.G.; Monson, P.R. (Westinghouse Savannah River Co., Aiken, SC (United States)); Mitchell, H.A. (Concord Associates, Inc., Knoxville, TN (United States))

1990-01-01T23:59:59.000Z

52

Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents  

SciTech Connect

This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises.

Ellison, P.G.; Monson, P.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Mitchell, H.A. [Concord Associates, Inc., Knoxville, TN (United States)

1990-12-31T23:59:59.000Z

53

Transportation and Stationary Power Integration Workshop Agenda, October 27, 2008, Phoenix, Arizonia  

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

Agenda for the Transportation and Stationary Power Integration Workshop held on October 27, 2008 in Phoenix, AZ

54

Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications  

SciTech Connect

This report is a summary of analyses performed by the NGNP project to determine whether it is technically and economically feasible to integrate high temperature gas cooled reactor (HTGR) technology into industrial processes. To avoid an overly optimistic environmental and economic baseline for comparing nuclear integrated and conventional processes, a conservative approach was used for the assumptions and calculations.

Lee Nelson

2011-09-01T23:59:59.000Z

55

An integrated approach for the verification of fresh mixed oxide fuel (MOX) assemblies at light water reactor MOX recycle reactors  

SciTech Connect

This paper presents an integrated approach for the verification of mixed oxide (MOX) fuel assemblies prior to their being loaded into the reactor. There is a coupling of the verification approach that starts at the fuel fabrication plant and stops with the transfer of the assemblies into the thermal reactor. The key measurement points are at the output of the fuel fabrication plant, the receipt at the reactor site, and the storage in the water pool as fresh fuel. The IAEA currently has the capability to measure the MOX fuel assemblies at the output of the fuel fabrication plants using a passive neutron coincidence counting systems of the passive neutron collar (PNCL) type. Also. at the MOX reactor pool, the underwater coincidence counter (UWCC) has been developed to measure the MOX assemblies in the water. The UWCC measurement requires that the fuel assembly be lifted about two meters up in the storage rack to avoid interference from the fuel that is stored in the rack. This paper presents a new method to verify the MOX fuel assemblies that are in the storage rack without the necessity of moving the fuel. The detector system is called the Underwater MOX Verification System (UMVS). The integration and relationship of the three measurements systems is described.

Menlove, Howard O [Los Alamos National Laboratory; Lee, Sang - Yoon [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

56

INTEGRATION OF HIGH TEMPERATURE GAS REACTORS WITH IN SITU OIL SHALE RETORTING  

SciTech Connect

This paper evaluates the integration of a high-temperature gas-cooled reactor (HTGR) to an in situ oil shale retort operation producing 7950 m3/D (50,000 bbl/day). The large amount of heat required to pyrolyze the oil shale and produce oil would typically be provided by combustion of fossil fuels, but can also be delivered by an HTGR. Two cases were considered: a base case which includes no nuclear integration, and an HTGR-integrated case.

Eric P. Robertson; Michael G. McKellar; Lee O. Nelson

2011-05-01T23:59:59.000Z

57

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

58

Light Water Reactor Sustainability Program Integrated Program Plan  

SciTech Connect

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

2013-04-01T23:59:59.000Z

59

Light Water Reactor Sustainability Program Integrated Program Plan  

SciTech Connect

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

McCarthy, Kathryn A [INL; Busby, Jeremy [ORNL; Hallbert, Bruce [INL; Bragg-Sitton, Shannon [INL; Smith, Curtis [INL; Barnard, Cathy [INL

2014-04-01T23:59:59.000Z

60

Light Water Reactor Sustainability Program Integrated Program Plan  

SciTech Connect

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

2012-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport reactor integrated" 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

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development  

SciTech Connect

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

Till, C.E.; Chang, Y.I.

1993-01-01T23:59:59.000Z

62

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development  

SciTech Connect

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

Till, C.E.; Chang, Y.I.

1993-03-01T23:59:59.000Z

63

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development  

SciTech Connect

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs.

Till, C.E.; Chang, Y.I.

1991-01-01T23:59:59.000Z

64

Actinide recycle potential in the IFR (Integral Fast Reactor)  

SciTech Connect

Rising concern about the greenhouse effect reinforces the need to reexamine the question of a next-generation reactor concept that can contribute significantly toward substitution for fossil-based energy generation. Even with only the nuclear capacity on-line today, world-wide reasonably assured uranium resources would last for only about 50 years. If nuclear is to make a significant contribution, breeding is a fundamental requirement. Uranium resources can then be extended by two orders of magnitude, making nuclear essentially a renewable energy source. The key technical elements of the IFR concept are metallic fuel and fuel cycle technology based on pyroprocessing. Pyroprocessing is radically different from the conventional PUREX reprocessing developed for the LWR oxide fuel. Chemical feasibility of pyroprocessing has been demonstrated. The next major step in the IFR development program will be the full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. IFR fuel cycle closure based on pyroprocessing can also have a dramatic impact on the waste management options, and in particular on the actinide recycling. 6 figs.

Chang, Y.I.

1989-01-01T23:59:59.000Z

65

Thermodynamic and transport properties of thoria–urania fuel of Advanced Heavy Water Reactor  

Science Journals Connector (OSTI)

High temperature thermochemistry of thoria–urania fuel for Advanced Heavy Water Reactor was investigated. Oxygen potential development within the matrix and distribution behaviors of the fission products (fps) in different phases were worked out with the help of thermodynamic and transport properties of the fps as well as fission generated oxygen and the detailed balance of the elements. Some of the necessary data for different properties were generated in this laboratory while others were taken from literatures. Noting the behavior of poor transports of gases and volatile species in the thoria rich fuel (thoria–3 mol% urania), the evaluation shows that the fuel will generally bear higher oxygen potential right from early stage of burnup, and Mo will play vital role to buffer the potential through the formation of its oxygen rich chemical states. The problems related to the poor transport and larger retention of fission gases (Xe) and volatiles (I, Te, Cs) are discussed.

M. Basu (Ali); R. Mishra; S.R. Bharadwaj; D. Das

2010-01-01T23:59:59.000Z

66

Dynamic Complexity Study of Nuclear Reactor and Process Heat Application Integration  

SciTech Connect

Abstract This paper describes the key obstacles and challenges facing the integration of nuclear reactors with process heat applications as they relate to dynamic issues. The paper also presents capabilities of current modeling and analysis tools available to investigate these issues. A pragmatic approach to an analysis is developed with the ultimate objective of improving the viability of nuclear energy as a heat source for process industries. The extension of nuclear energy to process heat industries would improve energy security and aid in reduction of carbon emissions by reducing demands for foreign derived fossil fuels. The paper begins with an overview of nuclear reactors and process application for potential use in an integrated system. Reactors are evaluated against specific characteristics that determine their compatibility with process applications such as heat outlet temperature. The reactor system categories include light water, heavy water, small to medium, near term high-temperature, and far term high temperature reactors. Low temperature process systems include desalination, district heating, and tar sands and shale oil recovery. High temperature processes that support hydrogen production include steam reforming, steam cracking, hydrogen production by electrolysis, and far-term applications such as the sulfur iodine chemical process and high-temperature electrolysis. A simple static matching between complementary systems is performed; however, to gain a true appreciation for system integration complexity, time dependent dynamic analysis is required. The paper identifies critical issues arising from dynamic complexity associated with integration of systems. Operational issues include scheduling conflicts and resource allocation for heat and electricity. Additionally, economic and safety considerations that could impact the successful integration of these systems are considered. Economic issues include the cost differential arising due to an integrated system and the economic allocation of electricity and heat resources. Safety issues include changes in regulatory constraints imposed on the facilities. Modeling and analysis tools, such as System Dynamics for time dependent operational and economic issues and RELAP5 3D for chemical transient affects, are evaluated. The results of this study advance the body of knowledge toward integration of nuclear reactors and process heat applications.

J'Tia Patrice Taylor; David E. Shropshire

2009-09-01T23:59:59.000Z

67

Transportation requirements for the disposition of excess weapon plutonium by burning in fission reactors  

SciTech Connect

Both the US and Russia are planning to dispose of about 50 Mg of excess weapon plutonium over a 25-year period. One option is to transfer the plutonium to Advanced Light Water (power) Reactors (ALWRs) for use as fuel. Subsequent disposal would then be considered commercial spent fuel. This disposition option, like others, involves the transportation of plutonium in various material forms as it proceeds through various points in the recovery operation. This paper examines both the disposition option and the issues surrounding the transportation of 50 Mg of excess plutonium within the US under current regulatory and infrastructure constraints. Transportation issues include criticality control, shielding, and containment of the contents. Allowable limits on each of these issues are specified by the applicable (or selected) regulation. The composition and form of the radioactive materials to be transported will determine, in part, the applicable portions of the regulations as well as the packaging design. The regulations and the packaging design, along with safeguard and security issues, will determine the quantity of plutonium or fuel assemblies per package as well as the number of packages per shipment and the type of highway carrier. For the disposition of 50 Mg of weapon plutonium using ALWRs in a 25-year campaign, the annual shipment rates are determined for the various types of carriers.

Hovingh, J.; Walter, C.E.

1996-01-01T23:59:59.000Z

68

Transportation risk assessment of radioactive wastes generated by the N-Reactor stabilization program at the Hanford Site, Washington  

SciTech Connect

The potential radiological and nonradiological risks associated with specific radioactive waste shipping campaigns at the Hanford Site are estimated. The shipping campaigns analyzed are associated with the transportation of wastes from the N-Reactor site at the 200-W Area, both within the Hanford Reservation, for disposal. The analysis is based on waste that would be generated from the N-Reactor stabilization program.

Wheeler, T.

1994-12-01T23:59:59.000Z

69

Transportation  

Science Journals Connector (OSTI)

The romantic rides in Sandburg’s “eagle-car” changed society. On the one hand, motor vehicle transportation is an integral thread of society’s fabric. On the other hand, excess mobility fractures old neighborh...

David Hafemeister

2014-01-01T23:59:59.000Z

70

A Review of Proposed Upgrades to the High Flux Isotope Reactor and Potential Impacts to Reactor Vessel Integrity  

SciTech Connect

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was scheduled in October 2000 to implement design upgrades that include the enlargement of the HB-2 and HB-4 beam tubes. Higher dose rates and higher radiation embrittlement rates were predicted for the two beam-tube nozzles and surrounding vessel areas. ORNL had performed calculations for the upgraded design to show that vessel integrity would be maintained at acceptable levels. Pacific Northwest National Laboratory (PNNL) was requested by the U.S. Department of Energy Headquarters (DOE/HQ) to perform an independent peer review of the ORNL evaluations. PNNL concluded that the calculated probabilities of failure for the HFIR vessel during hydrostatic tests and for operational conditions as estimated by ORNL are an acceptable basis for selecting pressures and test intervals for hydrostatic tests and for justifying continued operation of the vessel. While there were some uncertainties in the embrittlement predictions, the ongoing efforts at ORNL to measure fluence levels at critical locations of the vessel wall and to test materials from surveillance capsules should be effective in dealing with embrittlement uncertainties. It was recommended that ORNL continue to update their fracture mechanics calculations to reflect methods and data from ongoing research for commercial nuclear power plants. Such programs should provide improved data for vessel fracture mechanics calculations.

Simonen, Fredric A.

2001-05-31T23:59:59.000Z

71

Electric Car Sharing as an Integrated Part of Public Transport: Customers’ Needs and Experience  

Science Journals Connector (OSTI)

The project BeMobility/Berlin elektroMobil aims to investigate the benefits and draw backs of electric vehicles as part of the public transport system. Therefore, about 40 electric vehicles (EV) were integrated i...

Steffi Kramer; Christian Hoffmann…

2014-01-01T23:59:59.000Z

72

Overview of Options to Integrate Stationary Power Generation from Fuel Cells with Hydrogen Demand for the Transportation Sector  

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

Overview of Options to Integrate Stationary Overview of Options to Integrate Stationary Power Generation from Fuel Cells with Hydrogen Demand for the Transportation Sector Overview of Options to Integrate Stationary Overview of Options to Integrate Stationary Power Generation from Fuel Cells with Power Generation from Fuel Cells with Hydrogen Demand for the Transportation Hydrogen Demand for the Transportation Sector Sector Fred Joseck U.S. DOE Hydrogen Program Transportation and Stationary Power Integration Workshop (TSPI) Transportation and Stationary Power Transportation and Stationary Power Integration Workshop (TSPI) Integration Workshop (TSPI) Phoenix, Arizona October 27, 2008 2 Why Integration? * Move away from conventional thinking...fuel and power generation/supply separate * Make dramatic change, use economies of scale,

73

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

74

Enhanced Hydrogen Production Integrated with CO2 Separation in a Single-Stage Reactor  

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

EnhancEd hydrogEn Production EnhancEd hydrogEn Production intEgratEd with co 2 SEParation in a SinglE-StagE rEactor Description One alternative for the United States to establish independence from foreign energy sources is to utilize the nation's abundant domestic reserves of coal. Gasification provides a route to produce liquid fuels, chemical feedstocks, and hydrogen from coal. Coal continues to be viewed as the fuel source for the 21st century. Products from coal gasification, however, contain other gases, particularly carbon dioxide, as well as other contaminants that must be removed to produce the pure stream of hydrogen needed to operate fuel cells and other devices. This project seeks to demonstrate a technology to efficiently produce a pure hydrogen stream from

75

Proliferation resistance of the fuel cycle for the Integral Fast Reactor  

SciTech Connect

Argonne National Laboratory has developed an electrorefining pyrochemical process for recovery and recycle of metal fuel discharged from the Integral Fast Reactor (FR). This inherently low decontamination process has an overall decontamination factor of only about 100 for the plutonium metal product. As a result, all of the fuel cycle operations must be conducted in heavily shielded cells containing a high-purity argon atmosphere. The FR fuel cycle possesses high resistance to clandestine diversion or overt, state- supported removal of plutonium for nuclear weapons production because of two main factors: the highly radioactive product, which is also contaminated with heat- and neutron-producing isotopes of plutonium and other actinide elements, and the difficulty of removing material from the FR facility through the limited number of cell transfer locks without detection.

Burris, L.

1993-09-01T23:59:59.000Z

76

Fuel cycle facility control system for the Integral Fast Reactor Program  

SciTech Connect

As part of the Integral Fast Reactor (IFR) Fuel Demonstration, a new distributed control system designed, implemented and installed. The Fuel processes are a combination of chemical and machining processes operated remotely. To meet this special requirement, the new control system provides complete sequential logic control motion and positioning control and continuous PID loop control. Also, a centralized computer system provides near-real time nuclear material tracking, product quality control data archiving and a centralized reporting function. The control system was configured to use programmable logic controllers, small logic controllers, personal computers with touch screens, engineering work stations and interconnecting networks. By following a structured software development method the operator interface was standardized. The system has been installed and is presently being tested for operations.

Benedict, R.W.; Tate, D.A.

1993-09-01T23:59:59.000Z

77

Fission product transport and behavior during two postulated loss-of-flow transients in the Advanced Test Reactor  

SciTech Connect

The fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradations) in the Advanced Test Reactor (ATR) has been analyzed. These transients are designated ATR transients LCP 15 (high pressure) and LPP9 (low pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be studied. A probabilistic risk analysis was performed that indicated that the probability of occurrence for these two transients is on the order of 10[sup [minus]5] and 10[sup [minus]7] per reactor year for LCP15 and LPP9, respectively. The fission product behavior analysis included calculations of the gaseous and highly volatile fission product (xenon, krypton, cesium, iodine, and tellurium) inventories in the fuel before accident initiation, release of the fission products from the fuel into the reactor vessel during core melt, the probable chemical forms, and transport of the fission products from the core through the reactor vessel and existing piping to the confinement. In addition to a base-case analysis of fission product behavior, a series of analyses was performed to determine the sensitivity of fission product release to several parameters including steam flow rate, (structural) aluminum oxidation, and initial aerosol size. The base-case analyses indicate that the volatile fission products (excluding the noble gases) will be transported as condensed species on zinc aerosols.

Adams, J.P.; Carboneau, M.L.; Hagrman, D.L. (Idaho National Engineering Lab. EG and G Idaho, Idaho Falls, ID (United States))

1993-07-01T23:59:59.000Z

78

Proton Transport in Triflic Acid Hydrates Studied via Path Integral Car?Parrinello Molecular Dynamics  

Science Journals Connector (OSTI)

Proton Transport in Triflic Acid Hydrates Studied via Path Integral Car?Parrinello Molecular Dynamics ... The mono-, di-, and tetrahydrates of trifluoromethanesulfonic acid, which contain characteristic H3O+, H5O2+, and H9O4+ structures, provide model systems for understanding proton transport in materials with high perfluorosulfonic acid density such as perfluorosulfonic acid membranes commonly employed in hydrogen fuel cells. ... Hydrogen is described as a promising future fuel if the fuel cell technol. ...

Robin L. Hayes; Stephen J. Paddison; Mark E. Tuckerman

2009-12-07T23:59:59.000Z

79

Integrable Aspects of Universal Quantum Transport in Chaotic Cavities  

E-Print Network (OSTI)

The Painlev\\'e transcendents discovered at the turn of the XX century by pure mathematical reasoning, have later made their surprising appearance -- much in the way of Wigner's "miracle of appropriateness" -- in various problems of theoretical physics. The notable examples include the two-dimensional Ising model, one-dimensional impenetrable Bose gas, corner and polynuclear growth models, one dimensional directed polymers, string theory, two dimensional quantum gravity, and spectral distributions of random matrices. In the present contribution, ideas of integrability are utilized to advocate emergence of an one-dimensional Toda Lattice and the fifth Painlev\\'e transcendent in the paradigmatic problem of conductance fluctuations in quantum chaotic cavities coupled to the external world via ballistic point contacts. Specifically, the cumulants of the Landauer conductance of a cavity with broken time-reversal symmetry are proven to be furnished by the coefficients of a Taylor-expanded Painlev\\'e V function. Further, the relevance of the fifth Painlev\\'e transcendent for a closely related problem of sample-to-sample fluctuations of the noise power is discussed. Finally, it is demonstrated that inclusion of tunneling effects inherent in realistic point contacts does not destroy the integrability: in this case, conductance fluctuations are shown to be governed by a two-dimensional Toda Lattice.

Eugene Kanzieper

2014-10-02T23:59:59.000Z

80

Integration of renewable energy into the transport and electricity sectors through V2G  

E-Print Network (OSTI)

Keywords: V2G Vehicle to grid Energy system analysis Sustainable energy systems Electric vehicle EV for electricity, transport and heat, includes hourly fluctuations in human needs and the environment (wind energy systems allows integration of much higher levels of wind electricity without excess electric

Firestone, Jeremy

Note: This page contains sample records for the topic "transport reactor integrated" 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

Coupling of kinetic Monte Carlo simulations of surface reactions to transport in a fluid for heterogeneous catalytic reactor modeling  

SciTech Connect

We have developed a method to couple kinetic Monte Carlo simulations of surface reactions at a molecular scale to transport equations at a macroscopic scale. This method is applicable to steady state reactors. We use a finite difference upwinding scheme and a gap-tooth scheme to efficiently use a limited amount of kinetic Monte Carlo simulations. In general the stochastic kinetic Monte Carlo results do not obey mass conservation so that unphysical accumulation of mass could occur in the reactor. We have developed a method to perform mass balance corrections that is based on a stoichiometry matrix and a least-squares problem that is reduced to a non-singular set of linear equations that is applicable to any surface catalyzed reaction. The implementation of these methods is validated by comparing numerical results of a reactor simulation with a unimolecular reaction to an analytical solution. Furthermore, the method is applied to two reaction mechanisms. The first is the ZGB model for CO oxidation in which inevitable poisoning of the catalyst limits the performance of the reactor. The second is a model for the oxidation of NO on a Pt(111) surface, which becomes active due to lateral interaction at high coverages of oxygen. This reaction model is based on ab initio density functional theory calculations from literature.

Schaefer, C.; Jansen, A. P. J. [Eindhoven University of Technology, P.O. Box 513, 5600 MB Eindhoven (Netherlands)

2013-02-07T23:59:59.000Z

82

Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR (pressurized-water-reactor) plants  

SciTech Connect

Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs.

Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

1988-01-01T23:59:59.000Z

83

ENHANCED HYDROGEN PRODUCTION INTEGRATED WITH CO2 SEPARATION IN A SINGLE-STAGE REACTOR  

SciTech Connect

The water gas shift reaction (WGSR) plays a major role in increasing the hydrogen production from fossil fuels. However, the enhanced hydrogen production is limited by thermodynamic constrains posed by equilibrium limitations of WGSR. This project aims at using a mesoporous, tailored, highly reactive calcium based sorbent system for incessantly removing the CO{sub 2} product which drives the equilibrium limited WGSR forward. In addition, a pure sequestration ready CO{sub 2} stream is produced simultaneously. A detailed project vision with the description of integration of this concept with an existing coal gasification process for hydrogen production is presented. Conceptual reactor designs for investigating the simultaneous water gas shift and the CaO carbonation reactions are presented. In addition, the options for conducting in-situ sorbent regeneration under vacuum or steam are also reported. Preliminary, water gas shift reactions using high temperature shift catalyst and without any sorbent confirmed the equilibrium limitation beyond 600 C demonstrating a carbon monoxide conversion of about 80%. From detailed thermodynamic analyses performed for fuel gas streams from typical gasifiers the optimal operating temperature range to prevent CaO hydration and to effect its carbonation is between 575-830 C.

Himanshu Gupta; Mahesh Iyer; Bartev Sakadjian; Liang-Shih Fan

2005-03-10T23:59:59.000Z

84

The Mass Tracking System for the Integral Fast Reactor fuel cycle  

SciTech Connect

As part of the Fuel Cycle Facility (FCF) of Argonne National Laboratory`s Integral Fast Reactor (IFR) demonstration, a computer-based Mass-Tracking (MTG) System has been developed. The MTG System collects, stores, retrieves and processes data on all operations which directly affect the flow of process material through FCF and supports such activities as process modeling, compliance with operating limits (e.g., criticality safety), material control and accountability and operational information services. Its architecture is client/server, with input and output connections to operator`s equipment-control stations on the floor of FCF as well as to terminal sessions. Its heterogeneous database includes a relational-database manager as well as both binary and ASCII data files. The design of the database, and the software that supports it, is based on a model of discrete accountable items distributed in space and time and constitutes a complete historical record of the material processed in FCF. Although still under development, much of the MTG System has been qualified and is in production use.

Adams, C.H.; Beitel, J.C.; Birgersson, G.; Bucher, R.G.; Carrico, C.B.; Daly, T.A. [Argonne National Lab., IL (United States); Keyes, R.W. [Argonne National Lab., Idaho Falls, ID (United States)

1994-07-01T23:59:59.000Z

85

6 - Other nuclear energy applications: Hydrogen for transport desalination ships space research reactors for radioisotopes  

Science Journals Connector (OSTI)

Publisher Summary This chapter describes several nuclear energy applications. Hydrogen itself is likely to be an important future fuel; like electricity, it is an energy carrier. Nuclear energy can be used to make hydrogen electrolytically; and in the future, high-temperature reactors are likely to be used for thermochemical production. Desalination is energy-intensive. Nuclear energy is already being used for desalination, and nuclear energy has the potential for much greater use. Nuclear power has also revolutionized the navy; it is particularly suitable for vessels that need to be at sea for long periods without refueling, or for powerful submarine propulsion. After a gap of several years, there is a revival of interest in the use of nuclear fission power for space missions as well. Many of the world's nuclear reactors are used for research and training, materials testing, or the production of radioisotopes for medicine and industry. Research reactors are much smaller than power reactors or those propelling ships, and many are on university campuses. Research reactors are simpler than power reactors and operate at lower temperatures.

Ian Hore-Lacy

2007-01-01T23:59:59.000Z

86

Evaluating the Interstate Highway Transportation System in West Africa: Recommendations for an Integrated Highway Network  

E-Print Network (OSTI)

is wasted when the projects do not succeed. An integrated interstate highway system could transform the West African region by providing a transportation network 3-4 that links all of the West African nations together. The project would be the biggest...% reported traveling on highways in other parts of the world. Respondents indicated traveling in the following areas: Taiwan, China, Mexico, Brazil, Argentina, and Suriname. Three respondents indicated general geographic areas: two indicated Asia and one...

Nyang, Lamin Bumi

2010-12-17T23:59:59.000Z

87

INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT  

SciTech Connect

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

2008-09-01T23:59:59.000Z

88

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2  

SciTech Connect

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

Not Available

1992-04-01T23:59:59.000Z

89

An integrated performance model for high temperature gas cooled reactor coated particle fuel  

E-Print Network (OSTI)

The performance of coated fuel particles is essential for the development and deployment of High Temperature Gas Reactor (HTGR) systems for future power generation. Fuel performance modeling is indispensable for understanding ...

Wang, Jing, 1976-

2004-01-01T23:59:59.000Z

90

Thermal hydraulic performance analysis of a small integral pressurized water reactor core  

E-Print Network (OSTI)

A thermal hydraulic analysis of the International Reactor Innovative and Secure (IRIS) core has been performed. Thermal margins for steady state and a selection of Loss Of Flow Accidents have been assessed using three ...

Blair, Stuart R. (Stuart Ryan), 1972-

2003-01-01T23:59:59.000Z

91

Experimental characterization of an Ion Transport Membrane (ITM) reactor for methane oxyfuel combustion  

E-Print Network (OSTI)

Ion Transport Membranes (ITM) which conduct both electrons and oxygen ions have been investigated experimentally for oxygen separation and fuel (mostly methane) conversion purposes over the last three decades. The fuel ...

Apo, Daniel Jolomi

2012-01-01T23:59:59.000Z

92

Integration of numerical analysis tools for automated numerical optimization of a transportation package design  

SciTech Connect

The use of state-of-the-art numerical analysis tools to determine the optimal design of a radioactive material (RAM) transportation container is investigated. The design of a RAM package`s components involves a complex coupling of structural, thermal, and radioactive shielding analyses. The final design must adhere to very strict design constraints. The current technique used by cask designers is uncoupled and involves designing each component separately with respect to its driving constraint. With the use of numerical optimization schemes, the complex couplings can be considered directly, and the performance of the integrated package can be maximized with respect to the analysis conditions. This can lead to more efficient package designs. Thermal and structural accident conditions are analyzed in the shape optimization of a simplified cask design. In this paper, details of the integration of numerical analysis tools, development of a process model, nonsmoothness difficulties with the optimization of the cask, and preliminary results are discussed.

Witkowski, W.R.; Eldred, M.S.; Harding, D.C.

1994-09-01T23:59:59.000Z

93

Transport of fallout and reactor radionuclides in the drainage basin of the Hudson River estuary  

SciTech Connect

The transport and fate of Strontium 90, Cesium 137 and Plutonium 239, 240 in the Hudson River Estuary is discussed. Rates of radionuclide deposition and accumulation over time and space are calculated for the Hudson River watershed, estuary, and continental shelf offshore. 37 references, 7 figures, 15 tables. (ACR)

Simpson, H.J.; Linsalata, P.; Olsen, C.R.

1982-01-01T23:59:59.000Z

94

An Innovative Three-Dimensional Heterogeneous Coarse-Mesh Transport Method for Advanced and Generation IV Reactor Core Analysis and Design  

SciTech Connect

This project has resulted in a highly efficient method that has been shown to provide accurate solutions to a variety of 2D and 3D reactor problems. The goal of this project was to develop (1) an accurate and efficient three-dimensional whole-core neutronics method with the following features: based sollely on transport theory, does not require the use of cross-section homogenization, contains a highly accurate and self-consistent global flux reconstruction procedure, and is applicable to large, heterogeneous reactor models, and to (2) create new numerical benchmark problems for code cross-comparison.

Farzad Rahnema

2009-11-12T23:59:59.000Z

95

Reversible Bending Fatigue Test System for Investigating Vibration Integrity of Spent Nuclear Fuel during Transportation  

SciTech Connect

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading during road or rail shipment. Oak Ridge National Laboratory (ORNL) has been developing testing capabilities that can be used to improve the understanding of the impacts on SNF integrity due to vibration loading, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety and security of spent nuclear fuel storage and transport operations. The ORNL developed test system can perform reversible-bending fatigue testing to evaluate both the static and dynamic mechanical response of SNF rods under simulated loads. The testing apparatus is also designed to meet the challenges of hot-cell operation, including remote installation and detachment of the SNF test specimen, in-situ test specimen deformation measurement, and implementation of a driving system suitable for use in a hot cell. The system contains a U-frame set-up equipped with uniquely designed grip rigs, to protect SNF rod and to ensure valid test results, and use of 3 specially designed LVDTs to obtain the in-situ curvature measurement. A variety of surrogate test rods have been used to develop and calibrate the test system as well as in performing a series of systematic cyclic fatigue tests. The surrogate rods include stainless steel (SS) cladding, SS cladding with cast epoxy, and SS cladding with alumina pellets inserts simulating fuel pellets. Testing to date has shown that the interface bonding between the SS cladding and the alumina pellets has a significant impact on the bending response of the test rods as well as their fatigue strength. The failure behaviors observed from tested surrogate rods provides a fundamental understanding of the underlying failure mechanisms of the SNF surrogate rod under vibration which has not been achieved previously. The newly developed device is scheduled to be installed in the hot-cell in summer 2013 to test high burnup SNF.

Wang, Jy-An John [ORNL] [ORNL; Wang, Hong [ORNL] [ORNL; Bevard, Bruce Balkcom [ORNL] [ORNL; Howard, Rob L [ORNL] [ORNL; Flanagan, Michelle [U.S. Nuclear Regulatory Commission] [U.S. Nuclear Regulatory Commission

2013-01-01T23:59:59.000Z

96

Fluidized bed steam reactor including two horizontal cyclone separators and an integral recycle heat exchanger  

SciTech Connect

A reactor is described comprising: a vessel; a first furnace section disposed in said vessel; a second furnace section disposed in said vessel; means in each of said furnace sections for receiving a combustible fuel for generating heat and combustion gases; a first heat recovery area located adjacent said furnace sections; a second heat recovery area located adjacent said furnace sections; means for passing said combustion gases from said first furnace section to said first heat recovery area; and means for passing said combustion gases from said second furnace section to said second heat recovery area.

Gorzegno, W.P.

1993-06-15T23:59:59.000Z

97

Transportation  

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

Transportation Transportation Transportation of Depleted Uranium Materials in Support of the Depleted Uranium Hexafluoride Conversion Program Issues associated with transport of depleted UF6 cylinders and conversion products. Conversion Plan Transportation Requirements The DOE has prepared two Environmental Impact Statements (EISs) for the proposal to build and operate depleted uranium hexafluoride (UF6) conversion facilities at its Portsmouth and Paducah gaseous diffusion plant sites, pursuant to the National Environmental Policy Act (NEPA). The proposed action calls for transporting the cylinder at ETTP to Portsmouth for conversion. The transportation of depleted UF6 cylinders and of the depleted uranium conversion products following conversion was addressed in the EISs.

98

Enhanced Hydrogen Production Integrated with CO2 Separation in a Single-Stage Reactor  

SciTech Connect

High purity hydrogen is commercially produced from syngas by the Water Gas Shift Reaction (WGSR) in high and low temperature shift reactors using iron oxide and copper catalysts respectively. However, the WGSR is thermodynamically limited at high temperatures towards hydrogen production necessitating excess steam addition and catalytic operation. In the calcium looping process, the equilibrium limited WGSR is driven forward by the incessant removal of CO{sub 2} by-product through the carbonation of calcium oxide. At high pressures, this process obviates the need for a catalyst and excess steam requirement, thereby removing the costs related to the procurement and deactivation of the catalyst and steam generation. Thermodynamic analysis for the combined WGS and carbonation reaction was conducted. The combined WGS and carbonation reaction was investigated at varying pressures, temperatures and S/C ratios using a bench scale reactor system. It was found that the purity of hydrogen increases with the increase in pressure and at a pressure of 300 psig, almost 100% hydrogen is produced. It was also found that at high pressures, high purity hydrogen can be produced using stoichiometric quantities of steam. On comparing the catalytic and non catalytic modes of operation in the presence of calcium oxide, it was found that there was no difference in the purity of hydrogen produced at elevated pressures. Multicyclic reaction and regeneration experiments were also conducted and it was found that the purity of hydrogen remains almost constant after a few cycles.

Shwetha Ramkumar; Mahesh Iyer; Danny Wong; Himanshu Gupta; Bartev Sakadjian; Liang-Lhih Fan

2008-09-30T23:59:59.000Z

99

Enhanced Hydrogen Production Integrated with CO2 Separation in a Single-Stage Reactor  

SciTech Connect

Hydrogen production from coal gasification can be enhanced by driving the equilibrium limited Water Gas Shift reaction forward by incessantly removing the CO{sub 2} by-product via the carbonation of calcium oxide. This project aims at using the OSU patented high-reactivity mesoporous precipitated calcium carbonate sorbent for removing the CO{sub 2} product. Preliminary experiments demonstrate the show the superior performance of the PCC sorbent over other naturally occurring calcium sorbents. Gas composition analyses show the formation of 100% pure hydrogen. Novel calcination techniques could lead to smaller reactor footprint and single-stage reactors that can achieve maximum theoretical H{sub 2} production for multicyclic applications. Sub-atmospheric calcination studies reveal the effect of vacuum level, diluent gas flow rate, thermal properties of the diluent gas and the sorbent loading on the calcination kinetics which play an important role on the sorbent morphology. Steam, which can be easily separated from CO{sub 2}, is envisioned to be a potential diluent gas due to its enhanced thermal properties. Steam calcination studies at 700-850 C reveal improved sorbent morphology over regular nitrogen calcination. A mixture of 80% steam and 20% CO{sub 2} at ambient pressure was used to calcine the spent sorbent at 820 C thus lowering the calcination temperature. Regeneration of calcium sulfide to calcium carbonate was achieved by carbonating the calcium sulfide slurry by bubbling CO{sub 2} gas at room temperature.

Mahesh Iyer; Himanshu Gupta; Danny Wong; Liang-Shih Fan

2005-09-30T23:59:59.000Z

100

Transportation  

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

Health Risks » Transportation Health Risks » Transportation DUF6 Health Risks line line Accidents Storage Conversion Manufacturing Disposal Transportation Transportation A discussion of health risks associated with transport of depleted UF6. Transport Regulations and Requirements In the future, it is likely that depleted uranium hexafluoride cylinders will be transported to a conversion facility. For example, it is currently anticipated that the cylinders at the ETTP Site in Oak Ridge, TN, will be transported to the Portsmouth Site, OH, for conversion. Uranium hexafluoride has been shipped safely in the United States for over 40 years by both truck and rail. Shipments of depleted UF6 would be made in accordance with all applicable transportation regulations. Shipment of depleted UF6 is regulated by the

Note: This page contains sample records for the topic "transport reactor integrated" 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

The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986  

SciTech Connect

This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

Smith, R.R.

1986-01-01T23:59:59.000Z

102

Integrated Gasification Combined Cycle Dynamic Model: H2S Absorption/Stripping, Water?Gas Shift Reactors, and CO2 Absorption/Stripping  

Science Journals Connector (OSTI)

Integrated Gasification Combined Cycle Dynamic Model: H2S Absorption/Stripping, Water?Gas Shift Reactors, and CO2 Absorption/Stripping ... Future chemical plants may be required to have much higher flexibility and agility than existing process facilities in order to be able to handle new hybrid combinations of power and chemical units. ...

Patrick J. Robinson; William L. Luyben

2010-04-26T23:59:59.000Z

103

Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications  

SciTech Connect

This report is a preliminary comparison of conventional and potential HTGR-integrated processesa in several common industrial areas: ? Producing electricity via a traditional power cycle ? Producing hydrogen ? Producing ammonia and ammonia-derived products, such as fertilizer ? Producing gasoline and diesel from natural gas or coal ? Producing substitute natural gas from coal, and ? Steam-assisted gravity drainage (extracting oil from tar sands).

Lee Nelson

2009-10-01T23:59:59.000Z

104

Integrated Graduate Education & Research Traineeships (IGERT): Transportation Technology & Policy Final Grant Report  

E-Print Network (OSTI)

Technology & Policy Final Grant Report Patricia L.and Research Traineeships Grant TRANSPORTATION TECHNOLOGYThe UC Davis IGERT grant for Transportation Technology and

Mokhtarian, Patricia L; Tolentino, Joan S.

2005-01-01T23:59:59.000Z

105

Integrated Graduate Education & Research Traineeships: Transportation Technology & Policy Final Grant Report  

E-Print Network (OSTI)

Technology & Policy Final Grant Report Patricia L.and Research Traineeships Grant TRANSPORTATION TECHNOLOGYThe UC Davis IGERT grant for Transportation Technology and

Mokhtarian, Patricia L; Tolentino, Joan

2005-01-01T23:59:59.000Z

106

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

107

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

108

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

109

TRITIUM PERMEATION AND TRANSPORT IN THE GASOLINE PRODUCTION SYSTEM COUPLED WITH HIGH TEMPERATURE GAS-COOLED REACTORS (HTGRS)  

SciTech Connect

This paper describes scoping analyses on tritium behaviors in the HTGR-integrated gasoline production system, which is based on a methanol-to-gasoline (MTG) plant. In this system, the HTGR transfers heat and electricity to the MTG system. This system was analyzed using the TPAC code, which was recently developed by Idaho National Laboratory. The global sensitivity analyses were performed to understand and characterize tritium behaviors in the coupled HTGR/MTG system. This Monte Carlo based random sampling method was used to evaluate maximum 17,408 numbers of samples with different input values. According to the analyses, the average tritium concentration in the product gasoline is about 3.05×10-3 Bq/cm3, and 62 % cases are within the tritium effluent limit (= 3.7x10-3 Bq/cm3[STP]). About 0.19% of released tritium is finally transported from the core to the gasoline product through permeations. This study also identified that the following four parameters are important concerning tritium behaviors in the HTGR/MTG system: (1) tritium source, (2) wall thickness of process heat exchanger, (3) operating temperature, and (4) tritium permeation coefficient of process heat exchanger. These four parameters contribute about 95 % of the total output uncertainties. This study strongly recommends focusing our future research on these four parameters to improve modeling accuracy and to mitigate tritium permeation into the gasol ine product. If the permeation barrier is included in the future study, the tritium concentration will be significantly reduced.

Chang H. Oh; Eung S. Kim; Mike Patterson

2011-05-01T23:59:59.000Z

110

Maglev vehicles and superconductor technology: Integration of high-speed ground transportation into the air travel system  

SciTech Connect

This study was undertaken to (1) evaluate the potential contribution of high-temperature superconductors (HTSCs) to the technical and economic feasibility of magnetically levitated (maglev) vehicles, (2) determine the status of maglev transportation research in the United States and abroad, (3) identify the likelihood of a significant transportation market for high-speed maglev vehicles, and (4) provide a preliminary assessment of the potential energy and economic benefits of maglev systems. HTSCs should be considered as an enhancing, rather than an enabling, development for maglev transportation because they should improve reliability and reduce energy and maintenance costs. Superconducting maglev transportation technologies were developed in the United States in the late 1960s and early 1970s. Federal support was withdrawn in 1975, but major maglev transportation programs were continued in Japan and West Germany, where full-scale prototypes now carry passengers at speeds of 250 mi/h in demonstration runs. Maglev systems are generally viewed as very-high-speed train systems, but this study shows that the potential market for maglev technology as a train system, e.g., from one downtown to another, is limited. Rather, aircraft and maglev vehicles should be seen as complementing rather than competing transportation systems. If maglev systems were integrated into major hub airport operations, they could become economical in many relatively high-density US corridors. Air traffic congestion and associated noise and pollutant emissions around airports would also be reduced. 68 refs., 26 figs., 16 tabs.

Johnson, L.R.; Rote, D.M.; Hull, J.R.; Coffey, H.T.; Daley, J.G.; Giese, R.F.

1989-04-01T23:59:59.000Z

111

Status of the development of a fully integrated code system for the simulation of high temperature reactor cores  

Science Journals Connector (OSTI)

Abstract The HTR code package (HCP) is a new code system, which couples a variety of stand-alone codes for the simulation of different aspects of HTR. HCP will allow the steady-state and transient operating conditions of a 3D reactor core to be simulated including new features such as spatially resolved fission product release calculations or production and transport of graphite dust. For this code the latest programming techniques and standards are applied. As a first step an object-oriented data model was developed which features a high level of readability because it is based on problem-specific data types like Nuclide, Reaction, ReactionHandler, CrossSectionSet, etc. Those classes help to encapsulate and therefore hide specific implementations, which are not relevant with respect to physics. HCP will make use of one consistent data library for which an automatic generation tool was developed. The new data library consists of decay information, cross sections, fission yields, scattering matrices etc. for all available nuclides (e.g. ENDF/B-VII.1). The data can be stored in different formats such as binary, ASCII or XML. The new burn up code TNT (Topological Nuclide Transmutation) applies graph theory to represent nuclide chains and to minimize the calculation effort when solving the burn up equations. New features are the use of energy-dependent fission yields or the calculation of thermal power for decay, fission and capture reactions. With STACY (source term analysis code system) the fission product release for steady state as well as accident scenarios can be simulated for each fuel batch. For a full-core release calculation several thousand fuel elements are tracked while passing through the core. This models the stochastic behavior of a pebble bed in a realistic manner. In this paper we report on the current status of the HCP and present first results, which prove the applicability of the selected approach.

Stefan Kasselmann; Claudia Druska; Stefan Herber; Stephan Jühe; Florian Keller; Daniela Lambertz; Jingjing Li; Sarah Scholthaus; Dunfu Shi; Andre Xhonneux; Hans-Josef Allelein

2014-01-01T23:59:59.000Z

112

Integrated Catalytic Conversion of ?-Valerolactone to Liquid Alkenes for Transportation Fuels  

Science Journals Connector (OSTI)

...for Transportation Fuels 10.1126/science...Chemical and Biological Engineering, University of...synthesis of renewable fuels remains a challenging...corn ethanol and biodiesel, have the capacity...of transportation fuels from biomass: chemistry...catalysts, and engineering. Chem. Rev. 106...

Jesse Q. Bond; David Martin Alonso; Dong Wang; Ryan M. West; James A. Dumesic

2010-02-26T23:59:59.000Z

113

Integration of Nontraditional Isotopic Systems Into Reaction-Transport Models of EGS For Exploration, Evaluation of Water-Rock Interaction, and Impacts of Water Chemistry on Reservoir Sustainability  

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

Integration of Nontraditional Isotopic Systems Into Reaction-Transport Models of EGS For Exploration, Evaluation of Water-Rock Interaction, and Impacts of Water Chemistry on Reservoir Sustainability presentation at the April 2013 peer review meeting held in Denver, Colorado.

114

Integrated fuel performance and thermal-hydraulic sub-channel models for analysis of sodium fast reactors  

E-Print Network (OSTI)

Sodium Fast Reactors (SFR) show promise as an effective way to produce clean safe nuclear power while properly managing the fuel cycle. Accurate computer modeling is an important step in the design and eventual licensing ...

Fricano, Joseph William

2012-01-01T23:59:59.000Z

115

Oil- and Coal-Based Sea Transportation Needs: An Integrated Forecasting Approach  

Science Journals Connector (OSTI)

Due to the global economic crisis and increasing environmental issues (agreements to...2...emissions), the overall transportation logistics sector is currently suffering from one of the most severe recessions (as...

Y. H. Venus Lun; Olli-Pekka Hilmola; Alexander M. Goulielmos…

2013-01-01T23:59:59.000Z

116

Towards integrated sustainable transportation profile: a case study of Gharb El-Balad district, Assiut City, Egypt  

Science Journals Connector (OSTI)

Despite transportation is a key necessity for humans, it was argued a significant contributor to environmental degradation. Responding to that, drafts of sustainable transportation (ST) were sketched, guiding principles were argued, and rating systems were pursued. However, the selection of the appropriate method to implement ST is a challengeable task; not only due to the diversity and multidimensionality of local contexts, but also due to the numerous attributes of ST and their varying relative weights. This paper aims at defining an integrated profile of ST. Firstly, the motive behind tackling the issue was introduced and the methodology was worked out. Definitions, guiding principles, indicators, and rating systems were reviewed. After that, taxonomy of ST indicators was carried out, the interrelationship of key-attributes was investigated, and strategic directions to ST were proposed, to be applied over a selected case study. At last, a discussion took place highlighting the opportunities and challenges.

Khaled Ali Youssef; Moataz Mohmoud

2011-01-01T23:59:59.000Z

117

The Integration of a Structural Water-Gas-Shift Catalyst with a Vanadium Alloy Hydrogen Transport Device  

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

9 9 The InTegraTIon of a STrucTural WaTer- gaS-ShIfT caTalyST WITh a VanadIum alloy hydrogen TranSporT deVIce Description The purpose of this project is to produce a scalable device that simultaneously performs both water-gas-shift (WGS) and hydrogen separation from a coal-derived synthesis gas stream. The justification of such a system is the improved efficiency for the overall production of hydrogen. Removing hydrogen from the synthesis gas (syngas) stream allows the WGS reaction to convert more carbon monoxide (CO) to carbon dioxide (CO 2 ) and maximizes the total hydrogen produced. An additional benefit is the reduction in capital cost of plant construction due to the removal of one step in the process by integrating WGS with the membrane separation device.

118

Transportation  

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

Due to limited parking, all visitors are strongly encouraged to: Due to limited parking, all visitors are strongly encouraged to: 1) car-pool, 2) take the Lab's special conference shuttle service, or 3) take the regular off-site shuttle. If you choose to use the regular off-site shuttle bus, you will need an authorized bus pass, which can be obtained by contacting Eric Essman in advance. Transportation & Visitor Information Location and Directions to the Lab: Lawrence Berkeley National Laboratory is located in Berkeley, on the hillside directly above the campus of University of California at Berkeley. The address is One Cyclotron Road, Berkeley, California 94720. For comprehensive directions to the lab, please refer to: http://www.lbl.gov/Workplace/Transportation.html Maps and Parking Information: On Thursday and Friday, a limited number (15) of barricaded reserved parking spaces will be available for NON-LBNL Staff SNAP Collaboration Meeting participants in parking lot K1, in front of building 54 (cafeteria). On Saturday, plenty of parking spaces will be available everywhere, as it is a non-work day.

119

Mercury + VisIt: Integration of a Real-Time Graphical Analysis Capability into a Monte Carlo Transport Code  

SciTech Connect

Validation of the problem definition and analysis of the results (tallies) produced during a Monte Carlo particle transport calculation can be a complicated, time-intensive processes. The time required for a person to create an accurate, validated combinatorial geometry (CG) or mesh-based representation of a complex problem, free of common errors such as gaps and overlapping cells, can range from days to weeks. The ability to interrogate the internal structure of a complex, three-dimensional (3-D) geometry, prior to running the transport calculation, can improve the user's confidence in the validity of the problem definition. With regard to the analysis of results, the process of extracting tally data from printed tables within a file is laborious and not an intuitive approach to understanding the results. The ability to display tally information overlaid on top of the problem geometry can decrease the time required for analysis and increase the user's understanding of the results. To this end, our team has integrated VisIt, a parallel, production-quality visualization and data analysis tool into Mercury, a massively-parallel Monte Carlo particle transport code. VisIt provides an API for real time visualization of a simulation as it is running. The user may select which plots to display from the VisIt GUI, or by sending VisIt a Python script from Mercury. The frequency at which plots are updated can be set and the user can visualize the simulation results as it is running.

O'Brien, M J; Procassini, R J; Joy, K I

2009-03-09T23:59:59.000Z

120

Recovery Act: Beneficial CO{sub 2} Capture in an Integrated Algal Biorefinery for Renewable Generation and Transportation Fuels  

SciTech Connect

DOE DE-FE0001888 Award, Phase 2, funded research, development, and deployment (RD&D) of Phycal’s pilot-scale, algae to biofuels, bioproducts, and processing facility in Hawai’i. Phycal’s algal-biofuel and bioproducts production system integrates several novel and mature technologies into a system that captures and reuses industrially produced carbon dioxide emissions, which would otherwise go directly to the atmosphere, for the manufacture of renewable energy products and bioproducts from algae (note that these algae are not genetically engineered). At the end of Phase 2, the project as proposed was to encompass 34 acres in Central Oahu and provide large open ponds for algal mass culturing, heterotrophic reactors for the Heteroboost™ process, processing facilities, water recycling facilities, anaerobic digestion facilities, and other integrated processes. The Phase 2 award was divided into two modules, Modules 1 & 2, where the Module 1 effort addressed critical scaling issues, tested highest risk technologies, and set the overall infrastructure needed for a Module 2. Phycal terminated the project prior to executing construction of the first Module. This Final Report covers the development research, detailed design, and the proposed operating strategy for Module 1 of Phase 2.

Lane, Christopher; Hampel, Kristin; Rismani-Yazdi, Hamid; Kessler, Ben; Moats, Kenneth; Park, Jonathan; Schwenk, Jacob; White, Nicholas; Bakhit, Anis; Bargiel, Jeff; Allnutt, F.C.

2014-03-31T23:59:59.000Z

Note: This page contains sample records for the topic "transport reactor integrated" 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

Formation and sustainment of internal transport barriers in the International Thermonuclear Experimental Reactor with the baseline heating mix  

SciTech Connect

Plasmas with internal transport barriers (ITBs) are a potential and attractive route to steady-state operation in ITER. These plasmas exhibit radially localized regions of improved confinement with steep pressure gradients in the plasma core, which drive large bootstrap current and generate hollow current profiles and negative magnetic shear. This work examines the formation and sustainment of ITBs in ITER with electron cyclotron heating and current drive. The time-dependent transport simulations indicate that, with a trade-off of the power delivered to the equatorial and to the upper launcher, the sustainment of steady-state ITBs can be demonstrated in ITER with the baseline heating configuration.

Poli, Francesca M.; Kessel, Charles E. [Princeton Plasma Physics laboratory, Princeton, New Jersey 08543 (United States)] [Princeton Plasma Physics laboratory, Princeton, New Jersey 08543 (United States)

2013-05-15T23:59:59.000Z

122

Determination of Optimal Process Flowrates and Reactor Design for Autothermal Hydrogen Production in a Heat-Integrated Ceramic Microchannel Network  

E-Print Network (OSTI)

emissions [19]. Hence, hydrogen can be produced on large scale from biomass feedstocks in centralized facilities and subsequently distributed at fueling stations and/or community locations as a universal clean fuel for transportation and power...

Damodharan, Shalini

2012-07-16T23:59:59.000Z

123

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

124

Integrated electronic transport and thermometry at milliKelvin temperatures and in strong magnetic fields  

SciTech Connect

We fabricated a He-3 immersion cell for transport measurements of semiconductor nanostructures at ultra low temperatures and in strong magnetic fields. We have a new scheme of field-independent thermometry based on quartz tuning fork Helium-3 viscometry which monitors the local temperature of the sample's environment in real time. The operation and measurement circuitry of the quartz viscometer is described in detail. We provide evidence that the temperature of two-dimensional electron gas confined to a GaAs quantum well follows the temperature of the quartz viscometer down to 4 mK.

Samkharadze, N.; Kumar, A.; Csathy, G. A. [Department of Physics, Purdue University, West Lafayette, Indiana 47907 (United States); Manfra, M. J. [Department of Physics, Birck Nanotechnology Center, School of Materials Engineering, and School of Electrical and Computer Engineering, Purdue University, West Lafayette, Indiana 47907 (United States); Pfeiffer, L. N.; West, K. W. [Princeton University, Princeton, New Jersey 08544 (United States)

2011-05-15T23:59:59.000Z

125

Are the transport fuel retail markets regionally integrated in Spain? Evidence from price transmission  

Science Journals Connector (OSTI)

Abstract In this paper we explore whether the Spanish retail fuel markets are integrated at the regional level. We perform a comparative analysis of the transmission of international wholesale fuel prices to retail fuel prices. Our results are in favor of market segmentation, since the degree of cost pass-through differs noticeably across provinces (NUTS 3) and this outcome is clearly robust to the exclusion of the island provinces. We also found that cost pass-through is more similar for those provinces belonging to the same autonomous community (NUTS 2). It is suggested that different regulations and criteria regarding the granting of administrative authorizations from the autonomous communities could be hindering the integration of geographical markets.

Jacint Balaguer; Jordi Ripollés

2014-01-01T23:59:59.000Z

126

Strategy for the Integration of Hydrogen as a Vehicle Fuel into the Existing Natural Gas Vehicle Fueling Infrastructure of the Interstate Clean Transportation Corridor Project: 22 April 2004--31 August 2005  

SciTech Connect

Evaluates opportunities to integrate hydrogen into the fueling stations of the Interstate Clean Transportation Corridor--an existing network of LNG fueling stations in California and Nevada.

Gladstein, Neandross and Associates

2005-09-01T23:59:59.000Z

127

POTENTIAL IMPACT OF INTERFACIAL BONDING EFFICIENCY ON USED NUCLEAR FUEL VIBRATION INTEGRITY DURING NORMAL TRANSPORTATION  

SciTech Connect

Finite element analysis (FEA) was used to investigate the impacts of interfacial bonding efficiency at pellet pellet and pellet clad interfaces on surrogate of used nuclear fuel (UNF) vibration integrity. The FEA simulation results were also validated and benchmarked with reversible bending fatigue test results on surrogate rods consisting of stainless steel (SS) tubes with alumina-pellet inserts. Bending moments (M) are applied to the FEA models to evaluate the system responses of the surrogate rods. From the induced curvature, , the flexural rigidity EI can be estimated as EI=M/ . The impacts of interfacial bonding efficiency include the moment carrying capacity distribution between pellets and clad and cohesion influence on the flexural rigidity of the surrogate rod system. The result also indicates that the immediate consequences of interfacial de-bonding are a load carrying capacity shift from the fuel pellets to the clad and a reduction of the composite rod flexural rigidity. Therefore, the flexural rigidity of the surrogate rod and the bending moment bearing capacity between the clad and fuel pellets are strongly dependent on the efficiency of interfacial bonding at the pellet pellet and pellet clad interfaces. FEA models will be further used to study UNF vibration integrity.

Jiang, Hao [ORNL] [ORNL; Wang, Jy-An John [ORNL] [ORNL; Wang, Hong [ORNL] [ORNL

2014-01-01T23:59:59.000Z

128

Chamber transport  

SciTech Connect

Heavy ion beam transport through the containment chamber plays a crucial role in all heavy ion fusion (HIF) scenarios. Here, several parameters are used to characterize the operating space for HIF beams; transport modes are assessed in relation to evolving target/accelerator requirements; results of recent relevant experiments and simulations of HIF transport are summarized; and relevant instabilities are reviewed. All transport options still exist, including (1) vacuum ballistic transport, (2) neutralized ballistic transport, and (3) channel-like transport. Presently, the European HIF program favors vacuum ballistic transport, while the US HIF program favors neutralized ballistic transport with channel-like transport as an alternate approach. Further transport research is needed to clearly guide selection of the most attractive, integrated HIF system.

OLSON,CRAIG L.

2000-05-17T23:59:59.000Z

129

Sustainability and Transport  

E-Print Network (OSTI)

2005. Integrating Sustainability into the Trans- portationTHOUGHT PIECE Sustainability and Transport by Richardof the concept of sustainability to transport planning. In

Gilbert, Richard

2006-01-01T23:59:59.000Z

130

Overview of the integration and operation of the USA/9904/B(U)F-85 RTG transportation system (RTGTS) for Cassini and future missions  

SciTech Connect

The USA/9904/B(U)F-85, Radioisotope Thermoelectric Generator (RTG) Transportation System (RTGTS) was designed and tested by Westinghouse Hanford Company but was transferred to EG and G Mound Applied Technologies to transport Radioisotope Thermoelectric Generators (RTGs) in support of the Cassini mission. EG and G Mound Applied Technologies is also the RTGTS custodian. Since the RTGTS is a new system, careful scrutiny must be applied not only to the integration of the system into Mound's operations but also the operation of the system so as to prevent any adverse affects to the performance of the RTGs. This paper details specific precautions that have been applied to the integration and operation of the RTGTS to protect the Cassini RTGs during loading, onloading, transportation, offloading, and unloading.

Miller, Roger G. [EG and G Mound Applied Technologies P.O. Box 3000 Miamisburg, Ohio 45343-3000 (United States)

1997-01-10T23:59:59.000Z

131

Source Term Estimation of Radioxenon Released from the Fukushima Dai-ichi Nuclear Reactors Using Measured Air Concentrations and Atmospheric Transport Modeling  

SciTech Connect

Systems designed to monitor airborne radionuclides released from underground nuclear explosions detected radioactive fallout from the Fukushima Daiichi nuclear accident in March 2011. Atmospheric transport modeling (ATM) of plumes of noble gases and particulates were performed soon after the accident to determine plausible detection locations of any radioactive releases to the atmosphere. We combine sampling data from multiple International Modeling System (IMS) locations in a new way to estimate the magnitude and time sequence of the releases. Dilution factors from the modeled plume at five different detection locations were combined with 57 atmospheric concentration measurements of 133-Xe taken from March 18 to March 23 to estimate the source term. This approach estimates that 59% of the 1.24×1019 Bq of 133-Xe present in the reactors at the time of the earthquake was released to the atmosphere over a three day period. Source term estimates from combinations of detection sites have lower spread than estimates based on measurements at single detection sites. Sensitivity cases based on data from four or more detection locations bound the source term between 35% and 255% of available xenon inventory.

Eslinger, Paul W. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Biegalski, S. [Univ. of Texas at Austin, TX (United States); Bowyer, Ted W. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Cooper, Matthew W. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Haas, Derek A. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hayes, James C. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hoffman, Ian [Radiation Protection Bureau, Health Canada, Ottawa, ON (Canada); Korpach, E. [Radiation Protection Bureau, Health Canada, Ottawa, ON (Canada); Yi, Jing [Radiation Protection Bureau, Health Canada, Ottawa, ON (Canada); Miley, Harry S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Rishel, Jeremy P. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Ungar, R. Kurt [Radiation Protection Bureau, Health Canada, Ottawa, ON (Canada); White, Brian [Radiation Protection Bureau, Health Canada, Ottawa, ON (Canada); Woods, Vincent T. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

2014-01-01T23:59:59.000Z

132

1 MODELING THE PERFORMANCE OF ULTRAVIOLET REACTOR IN EULERIAN AND LAGRANGIAN FRAMEWORKS  

E-Print Network (OSTI)

CFD models for simulating the performance of ultraviolet (UV) reactors for micro-organism inactivation were developed in Eulerian and Lagrangian frameworks, taking into account hydrodynamics, kinetics, and radiation field within UV reactor. In the Lagrangian framework, micro-organisms were treated as discrete particles where the trajectory was predicted by integrating the force balance on the particle. In the Eulerian framework, the conservation equation of species (microorganisms) was solved along with the transport equations. The fluid flow was characterized experimentally using particle image velocimetry (PIV) flow visualization techniques and modeled using CFD for a UV reactor prototype model. The performance of annular UV reactors with an inlet parallel and perpendicular to the reactor axis were investigated. The results indicated that the fluid flow distribution within the reactor volume can significantly affect the reactor performance. Both the Eulerian and Lagrangian models were used to obtain complimentary information on the reactors; while the Lagrangian method provided an estimation of the UV-fluence distribution and the trajectory of species, the Eulerian approach showed the concentration distribution and local photo-reaction rates. The combined information can be used to predict and monitor reactor performance and to improve the reactor design.

Angelo Sozzi; Fariborz Taghipour

2006-01-01T23:59:59.000Z

133

Fuel handling apparatus for a nuclear reactor  

DOE Patents (OSTI)

Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

Hawke, Basil C. (Solana Beach, CA)

1987-01-01T23:59:59.000Z

134

Simulated Performance of the Integrated PNAR and SINRD Detector Designed for Spent Fuel Measurements at the Fugen Reactor in Japan  

SciTech Connect

Objective is to investigate the use of Passive Neutron Albedo Reactivity (PNAR) and Self-Interrogation Neutron Resonance Densitometry (SINRD) to quantify fissile content in FUGEN spent fuel assemblies (FAs). Methodology used is: (1) Detector was designed using fission chambers (FCs); (2) Optimized design via MCNPX simulations; and (3) Plan to build and field test instrument in FY13. Significance was to improve safeguards verification of spent fuel assemblies in water and increase sensitivity to partial defects. MCNPX simulations were performed to optimize the design of the SINRD+PNAR detector. PNAR ratio was less sensitive to FA positioning than SINRD and SINRD ratio was more sensitive to Pu fissile mass than PNAR. Significance was that the integration of these techniques can be used to improve verification of spent fuel assemblies in water.

Lafleur, Adrienne M. [Los Alamos National Laboratory; Ulrich, Timothy J. II [Los Alamos National Laboratory; Menlove, Howard O. [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory; Seya, Michio [Japan Atomic Energy Agency; Bolind, Alan M. [Japan Atomic Energy Agency

2012-07-13T23:59:59.000Z

135

Overview of Options to Integrate Stationary Power Generation from Fuel Cells with Hydrogen Demand for the Transportation Sector  

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

Overview of lessons learned, integration, barriers, enablers, federal incentives, state programs, and benefits

136

TRANSPORTATION: INTEGRAL TO CREATING  

E-Print Network (OSTI)

steam-propelled vessel (Savannah) crossed Atlantic · 1843 - 1st wagon train departed Independence, MOLuhan "I think the internal combustion engine will disappear from the streets of our cities in the next

Minnesota, University of

137

Three-dimensional calculations of neutron streaming in the beam tubes of the ORNL HFIR (High Flux Isotope Reactor) Reactor  

SciTech Connect

The streaming of neutrons through the beam tubes in High Flux Isotope Reactor at Oak Ridge National Laboratory has resulted in a reduction of the fracture toughness of the reactor vessel. As a result, an evaluation of vessel integrity was undertaken in order to determine if the reactor can be operated again. As a part of this evaluation, three-dimensional neutron transport calculations were performed to obtain fluxes at points of interest in the wall of the vessel. By comparing the calculated and measured activation of dosimetry specimens from the vessel surveillance program, it was determined that the calculated flux shape was satisfactory to transpose the surveillance data to the locations in the vessel. A bias factor was applied to correct for the average C/E ratio of 0.69. 8 refs., 7 figs., 3 tabs.

Childs, R.L.; Rhoades, W.A.; Williams, L.R.

1988-01-01T23:59:59.000Z

138

E-Print Network 3.0 - atr reactor Sample Search Results  

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

(ITM) Reactor - Air Products and Chemicals, Inc. Autothermal Reforming (ATR) - Praxair Inc. 12... day H2 Ion Transport Membrane Reactor (ITM) production unit...

139

Reactor and Nuclear Systems Division (RNSD)  

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

RNSD Home RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Staff Details (CV/Bios) Publications Org Chart Contact Us ORNL Staff Only Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Reactor and Nuclear Systems Division News Highlights U.S. Rep. Fleischmann touts ORNL as national energy treasure Martin Peng wins Fusion Power Associates Leadership Award

140

Engineering Development of Ceramic Membrane Reactor  

E-Print Network (OSTI)

ceramic Ion Transport Membrane (ITM) reactor system for low-cost conversion of natural gas to hydrogen;7 A Revolutionary Technology Using Ceramic Membranes Ion Transport Membranes (ITM) ­ Non-porous multiEngineering Development of Ceramic Membrane Reactor Systems for Converting Natural Gas to Hydrogen

Note: This page contains sample records for the topic "transport reactor integrated" 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

Integrated polymer solar cells in serial architecture with patterned charge-transporting MoOx for miniature high-voltage sources  

Science Journals Connector (OSTI)

We develop miniature high-voltage sources from polymer solar cells (PSCs) with charge-transporting molybdenum oxide (MoOx) integrated in a serial architecture through sacrificial layer (SL)-assisted patterning. The MoOx layer, being patterned by the lift-off process of the SL of a hydrophobic fluorinated-polymer, as a hole transporting layer plays a critical role on the reduction of the dark current and the increase of a high open circuit voltage of an integrated PSC array. The underlying mechanism lies primarily on the elimination of the lateral charge pathways in the MoOx layer in the presence of the electrode interconnection. Two miniature voltage sources consisting of 20 PSCs and 50 PSCs are demonstrated in the operation of a liquid crystal display and an organic field-effect transistor, respectively. Our SL-assisted integration approach will be directly applicable for implementing the self-power sources made of the PSCs into a wide range of the electronic and optoelectronic devices.

Seong-Min Cho; Chang-Min Keum; Hea-Lim Park; Min-Hoi Kim; Jin-Hyuk Bae; Sin-Doo Lee

2014-01-01T23:59:59.000Z

142

The performance of ENDF/B-V. 2 nuclear data for fast reactor calculations  

SciTech Connect

Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations.

Atkinson, C.A.; Collins, P.J.

1987-01-01T23:59:59.000Z

143

Nuclear Fuel Cycle Integrated System Analysis  

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

Fuel Cycle Integrated System Analysis Fuel Cycle Integrated System Analysis Abdellatif M. Yacout Argonne National Laboratory Nuclear Engineering Division The nuclear fuel cycle is a complex system with multiple components and activities that are combined to provide nuclear energy to a variety of end users. The end uses of nuclear energy are diverse and include electricity, process heat, water desalination, district heating, and possibly future hydrogen production for transportation and energy storage uses. Components of the nuclear fuel cycle include front end components such as uranium mining, conversion and enrichment, fuel fabrication, and the reactor component. Back end of the fuel cycle include used fuel coming out the reactor, used fuel temporary and permanent storage, and fuel reprocessing. Combined with those components there

144

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

145

Sandia National Laboratories: renewable energy integration  

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

Grid Integration, Infrastructure Security, Microgrid, News, News & Events, Partnership, Renewable Energy, SMART Grid, Transmission Grid Integration, Transportation Energy Under...

146

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

147

Reactors: Modern-Day Alchemy - Argonne's Nuclear Science and Technology  

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

Achievements > Achievements > Legacy > Reactors: Modern-Day Alchemy 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

148

Achievements: Nuclear Reactors designed/built by Argonne National  

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

Achievements > Achievements > Argonne National Laboratory 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

149

Integrated Ceramic Membrane System for Hydrogen Production  

SciTech Connect

Phase I was a technoeconomic feasibility study that defined the process scheme for the integrated ceramic membrane system for hydrogen production and determined the plan for Phase II. The hydrogen production system is comprised of an oxygen transport membrane (OTM) and a hydrogen transport membrane (HTM). Two process options were evaluated: 1) Integrated OTM-HTM reactor – in this configuration, the HTM was a ceramic proton conductor operating at temperatures up to 900°C, and 2) Sequential OTM and HTM reactors – in this configuration, the HTM was assumed to be a Pd alloy operating at less than 600°C. The analysis suggested that there are no technical issues related to either system that cannot be managed. The process with the sequential reactors was found to be more efficient, less expensive, and more likely to be commercialized in a shorter time than the single reactor. Therefore, Phase II focused on the sequential reactor system, specifically, the second stage, or the HTM portion. Work on the OTM portion was conducted in a separate program. Phase IIA began in February 2003. Candidate substrate materials and alloys were identified and porous ceramic tubes were produced and coated with Pd. Much effort was made to develop porous substrates with reasonable pore sizes suitable for Pd alloy coating. The second generation of tubes showed some improvement in pore size control, but this was not enough to get a viable membrane. Further improvements were made to the porous ceramic tube manufacturing process. When a support tube was successfully coated, the membrane was tested to determine the hydrogen flux. The results from all these tests were used to update the technoeconomic analysis from Phase I to confirm that the sequential membrane reactor system can potentially be a low-cost hydrogen supply option when using an existing membrane on a larger scale. Phase IIB began in October 2004 and focused on demonstrating an integrated HTM/water gas shift (WGS) reactor to increase CO conversion and produce more hydrogen than a standard water gas shift reactor would. Substantial improvements in substrate and membrane performance were achieved in another DOE project (DE-FC26-07NT43054). These improved membranes were used for testing in a water gas shift environment in this program. The amount of net H2 generated (defined as the difference of hydrogen produced and fed) was greater than would be produced at equilibrium using conventional water gas shift reactors up to 75 psig because of the shift in equilibrium caused by continuous hydrogen removal. However, methanation happened at higher pressures, 100 and 125 psig, and resulted in less net H2 generated than would be expected by equilibrium conversion alone. An effort to avoid methanation by testing in more oxidizing conditions (by increasing CO2/CO ratio in a feed gas) was successful and net H2 generated was higher (40-60%) than a conventional reactor at equilibrium at all pressures tested (up to 125 psig). A model was developed to predict reactor performance in both cases with and without methanation. The required membrane area depends on conditions, but the required membrane area is about 10 ft2 to produce about 2000 scfh of hydrogen. The maximum amount of hydrogen that can be produced in a membrane reactor decreased significantly due to methanation from about 2600 scfh to about 2400 scfh. Therefore, it is critical to eliminate methanation to fully benefit from the use of a membrane in the reaction. Other modeling work showed that operating a membrane reactor at higher temperature provides an opportunity to make the reactor smaller and potentially provides a significant capital cost savings compared to a shift reactor/PSA combination.

Schwartz, Joseph; Lim, Hankwon; Drnevich, Raymond

2010-08-05T23:59:59.000Z

150

Graduate Certificate in Transportation  

E-Print Network (OSTI)

Graduate Certificate in Transportation Nohad A. Toulan School of Urban Studies and Planning of Engineering and Computer Science integrated transportation systems. The Graduate Certificate in Transportation their capabilities. Students in the program can choose among a wide range of relevant courses in transportation

Bertini, Robert L.

151

REVIEW OF SCIENTIFIC INSTRUMENTS 82, 053902 (2011) Integrated electronic transport and thermometry at milliKelvin temperatures  

E-Print Network (OSTI)

of semiconductor nanostructures at ultra low temperatures and in strong magnetic fields. We have a new scheme to 4 mK. © 2011 American Institute of Physics. [doi:10.1063/1.3586766] I. INTRODUCTION One of the most excitations obeying non-Abelian statistics.3 Extending transport experiments to ultra low temperatures

Manfra, Michael J.

152

Intelligent Transportation Systems - Center for Transportation Analysis  

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

Intelligent Transportation Systems Intelligent Transportation Systems The Center for Transportation Analysis does specialty research and development in intelligent transportation systems. Intelligent Transportation Systems (ITS) are part of the national strategy for improving the operational safety, efficiency, and security of our nation's highways. Since the early 1990s, ITS has been the umbrella under which significant efforts have been conducted in research, development, testing, deployment and integration of advanced technologies to improve the measures of effectiveness of our national highway network. These measures include level of congestion, the number of accidents and fatalities, delay, throughput, access to transportation, and fuel efficiency. A transportation future that includes ITS will involve a significant improvement in these

153

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

154

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

155

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

156

Three-dimensional discrete ordinates radiation transport calculations of neutron fluxes for beginning-of-cycle at several pressure vessel surveillance positions in the high flux isotope reactor  

SciTech Connect

The objective of this research was to determine improved thermal, epithermal, and fast fluxes and several responses at mechanical test surveillance location keys 2, 4, 5, and 7 of the pressure vessel of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) for the beginning of the fuel cycle. The purpose of the research was to provide essential flux data in support of radiation embrittlement studies of the pressure vessel shell and beam tubes at some of the important locations.

Pace, J.V. III; Slater, C.O.; Smith, M.S.

1993-11-01T23:59:59.000Z

157

1896 IEEE TRANSACTIONS ON INTELLIGENT TRANSPORTATION SYSTEMS, VOL. 14, NO. 4, DECEMBER 2013 Probabilistic Integration of Intensity and Depth  

E-Print Network (OSTI)

of the system to different object classes. We apply our method to the problem of detecting vehicles from or semiautonomous driving systems require a detailed modeling of the vehicle's surroundings to detect potentially Probabilistic Integration of Intensity and Depth Information for Part-Based Vehicle Detection Alexandros Makris

Paris-Sud XI, Université de

158

Fast-ignition transport studies: Realistic electron source, integrated particle-in-cell and hydrodynamic modeling, imposed magnetic fields  

SciTech Connect

Transport modeling of idealized, cone-guided fast ignition targets indicates the severe challenge posed by fast-electron source divergence. The hybrid particle-in-cell (PIC) code Zuma is run in tandem with the radiation-hydrodynamics code Hydra to model fast-electron propagation, fuel heating, and thermonuclear burn. The fast electron source is based on a 3D explicit-PIC laser-plasma simulation with the PSC code. This shows a quasi two-temperature energy spectrum and a divergent angle spectrum (average velocity-space polar angle of 52 Degree-Sign ). Transport simulations with the PIC-based divergence do not ignite for >1 MJ of fast-electron energy, for a modest (70 {mu}m) standoff distance from fast-electron injection to the dense fuel. However, artificially collimating the source gives an ignition energy of 132 kJ. To mitigate the divergence, we consider imposed axial magnetic fields. Uniform fields {approx}50 MG are sufficient to recover the artificially collimated ignition energy. Experiments at the Omega laser facility have generated fields of this magnitude by imploding a capsule in seed fields of 50-100 kG. Such imploded fields will likely be more compressed in the transport region than in the laser absorption region. When fast electrons encounter increasing field strength, magnetic mirroring can reflect a substantial fraction of them and reduce coupling to the fuel. A hollow magnetic pipe, which peaks at a finite radius, is presented as one field configuration which circumvents mirroring.

Strozzi, D. J.; Tabak, M.; Larson, D. J.; Divol, L.; Kemp, A. J.; Bellei, C.; Marinak, M. M.; Key, M. H. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, California 94550 (United States)

2012-07-15T23:59:59.000Z

159

Early Exploration - Reactors designed/built by Argonne National Laboratory  

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

Early Exploration Early Exploration 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

Integrated Kinetic Simulation of Laser-Plasma Interactions, Fast-Electron Generation and Transport in Fast Ignition  

SciTech Connect

We present new results on the physics of short-pulse laser-matter interaction of kilojoule-picosecond pulses at full spatial and temporal scale, using a new approach that combines a 3D collisional electromagnetic Particle-in-Cell code with an MHD-hybrid model of high-density plasma. In the latter, collisions damp out plasma waves, and an Ohm's law with electron inertia effects neglected determines the electric field. In addition to yielding orders of magnitude in speed-up while avoiding numerical instabilities, this allows us to model the whole problem in a single unified framework: the laser-plasma interaction at sub-critical densities, energy deposition at relativistic critical densities, and fast-electron transport in solid densities. Key questions such as the multi-picosecond temporal evolution of the laser energy conversion into hot electrons, the impact of return currents on the laser-plasma interaction, and the effect of self-generated electric and magnetic fields on electron transport will be addressed. We will report applications to current experiments.

Kemp, A; Cohen, B; Divol, L

2009-11-16T23:59:59.000Z

Note: This page contains sample records for the topic "transport reactor integrated" 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

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

162

RAMI Analysis Program Design and Research for CFETR (Chinese Fusion Engineering Testing Reactor) Tokamak Machine  

Science Journals Connector (OSTI)

Chinese Fusion Engineering Testing Reactor (CFETR) is a test reactor which shall be constructed by National Integration Design Group for Magnetic Confinement Fusion Reactor of China with an ambitious scientific ...

Shijun Qin; Yuntao Song; Damao Yao; Yuanxi Wan; Songtao Wu…

2014-10-01T23:59:59.000Z

163

Thermographic analysis of polyurethane foams integrated with phase change materials designed for dynamic thermal insulation in refrigerated transport  

Science Journals Connector (OSTI)

Abstract The dispersion process of a micro-encapsulated phase change material (n-tetradecane) into a polyurethane foam was studied in order to develop a micro-composite insulating material with both low thermal conductivity and latent heat storage properties. The maximum weight content of micro-capsules added to the cellular matrix was 13.5%. Dynamic thermal properties of hybrid foams were investigated by means of a thermographic analysis. This was found to be a very effective diagnostic technique in detecting the change in heat transfer rate across the micro-composite foam in an indirect way, i.e. by measuring how the surface temperature changes over time under heat irradiation. Such a material would be of interest in the field of transport of perishable goods, particularly those requiring a controlled regime of carriage/storage temperatures.

Andrea Tinti; Antonella Tarzia; Alessandra Passaro; Riccardo Angiuli

2014-01-01T23:59:59.000Z

164

Potential Impact of Interfacial Bonding Efficiency on High-Burnup Spent Nuclear Fuel Vibration Integrity during Normal Transportation  

SciTech Connect

Finite element analysis (FEA) was used to investigate the impacts of interfacial bonding efficiency at pellet pellet and pellet clad interfaces on spent nuclear fuel (SNF) vibration integrity. The FEA simulation results were also validated and benchmarked with reverse bending fatigue test results on surrogate rods consisting of stainless steel (SS) tubes with alumina-pellet inserts. Bending moments (M) are applied to the FEA models to evaluate the system responses of the surrogate rods. From the induced curvature, , the flexural rigidity EI can be estimated as EI=M/ . The impacts of interfacial bonding efficiency on SNF vibration integrity include the moment carrying capacity distribution between pellets and clad and the impact of cohesion on the flexural rigidity of the surrogate rod system. The result also indicates that the immediate consequences of interfacial de-bonding are a load carrying capacity shift from the fuel pellets to the clad and a reduction of the composite rod flexural rigidity. Therefore, the flexural rigidity of the surrogate rod and the bending moment bearing capacity between the clad and fuel pellets are strongly dependent on the efficiency of interfacial bonding at the pellet pellet and pellet clad interfaces. The above-noted phenomenon was calibrated and validated by reverse bending fatigue testing using a surrogate rod system.

Jiang, Hao [ORNL] [ORNL; Wang, Jy-An John [ORNL] [ORNL; Wang, Hong [ORNL] [ORNL

2014-01-01T23:59:59.000Z

165

A Local Incident Flux Response Expansion Transport Method for Coupling to the Diffusion Method in Cylindrical Geometry  

SciTech Connect

A local incident flux response expansion transport method is developed to generate transport solutions for coupling to diffusion theory codes regardless of their solution method (e.g., fine mesh, nodal, response based, finite element, etc.) for reactor core calculations in both two-dimensional (2-D) and three-dimensional (3-D) cylindrical geometries. In this approach, a Monte Carlo method is first used to precompute the local transport solution (i.e., response function library) for each unique transport coarse node, in which diffusion theory is not valid due to strong transport effects. The response function library is then used to iteratively determine the albedo coefficients on the diffusion-transport interfaces, which are then used as the coupling parameters within the diffusion code. This interface coupling technique allows a seamless integration of the transport and diffusion methods. The new method retains the detailed heterogeneity of the transport nodes and naturally constructs any local solution within them by a simple superposition of local responses to all incoming fluxes from the contiguous coarse nodes. A new technique is also developed for coupling to fine-mesh diffusion methods/codes. The local transport method/module is tested in 2-D and 3-D pebble-bed reactor benchmark problems consisting of an inner reflector, an annular fuel region, and a controlled outer reflector. It is found that the results predicted by the transport module agree very well with the reference fluxes calculated directly by MCNP in both benchmark problems.

Dingkang Zhang; Farzad Rahnema; Abderrafi M. Ougouag

2013-09-01T23:59:59.000Z

166

The MC21 Monte Carlo Transport Code  

SciTech Connect

MC21 is a new Monte Carlo neutron and photon transport code currently under joint development at the Knolls Atomic Power Laboratory and the Bettis Atomic Power Laboratory. MC21 is the Monte Carlo transport kernel of the broader Common Monte Carlo Design Tool (CMCDT), which is also currently under development. The vision for CMCDT is to provide an automated, computer-aided modeling and post-processing environment integrated with a Monte Carlo solver that is optimized for reactor analysis. CMCDT represents a strategy to push the Monte Carlo method beyond its traditional role as a benchmarking tool or ''tool of last resort'' and into a dominant design role. This paper describes various aspects of the code, including the neutron physics and nuclear data treatments, the geometry representation, and the tally and depletion capabilities.

Sutton TM, Donovan TJ, Trumbull TH, Dobreff PS, Caro E, Griesheimer DP, Tyburski LJ, Carpenter DC, Joo H

2007-01-09T23:59:59.000Z

167

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"...

168

Argonne step closer to safer nuclear reactor  

Science Journals Connector (OSTI)

Argonne step closer to safer nuclear reactor ... "A key technological link" toward development of meltdown-immune nuclear reactors is now in the demonstration phase at Argonne National Laboratory near Chicago. ... The technique is part of Argonne's continuing interest in the sodium-cooled integral fast reactor (IFR), whose immunity to meltdown derives from molten sodium's function as a heat sink and the use of metallic fuel that conducts heat better than conventional oxide fuels. ...

WARD WORTHY

1988-05-30T23:59:59.000Z

169

What are Intelligent Transportation Intelligent Transportation Systems (ITS) are  

E-Print Network (OSTI)

What are Intelligent Transportation Systems? Intelligent Transportation Systems (ITS) are existing, combined in innovative ways, integrated into the management of our multimodal transportation system aimed at saving lives, time, and resources. Transportation is the backbone of our society-- the movement of people

Bertini, Robert L.

170

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

171

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...

172

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

173

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

174

Chicago Pile reactors create enduring research legacy - Argonne's  

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

Chicago Pile reactors create enduring research Chicago Pile reactors create enduring research legacy 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

175

Early Argonne reactor lit the way for worldwide nuclear industry -  

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

Early Argonne reactor lit the way for worldwide Early Argonne reactor lit the way for worldwide nuclear industry 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

176

Nuclear Reactor Technologies | Department of Energy  

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

Reactor Technologies Reactor Technologies Nuclear Reactor Technologies TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Small Modular Reactor Technologies Small modular reactors can also be made in factories and transported to sites where they would be ready to "plug and play" upon arrival, reducing both capital costs and construction times. The smaller size also makes these reactors ideal for small electric grids and for locations that

177

Strategy for the Integration of Hydrogen as a Vehicle Fuel into the Existing Natural Gas Vehicle Fueling Infrastructure of the Interstate Clean Transportation Corridor Project: 22 April 2004--31 August 2005  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

national laboratory of the U.S. Department of Energy national laboratory of the U.S. Department of Energy Office of Energy Efficiency & Renewable Energy National Renewable Energy Laboratory Innovation for Our Energy Future Subcontract Report Strategy for the Integration of NREL/SR-540-38720ďż˝ Hydrogen as a Vehicle Fuel into September 2005 ďż˝ the Existing Natural Gas Vehicle ďż˝ Fueling Infrastructure of the ďż˝ Interstate Clean Transportation ďż˝ Corridor Project ďż˝ April 22, 2004 - August 31, 2005 Gladstein, Neandross & Associates ďż˝ Santa Monica, California ďż˝ NREL is operated by Midwest Research Institute â—Ź Battelle Contract No. DE-AC36-99-GO10337 Strategy for the Integration of Hydrogen as a Vehicle Fuel into the Existing Natural Gas Vehicle Fueling Infrastructure of the Interstate Clean Transportation

178

PROTEUS - Simulation Toolset for Reactor Physics and Fuel Cycle Analysis  

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

Simulation Toolset for Simulation Toolset for Reactor Physics and Fuel Cycle Analysis PROTEUS Faster and more accurate neutronics calculations enable optimum reactor design... Argonne National Laboratory's powerful reactor physics toolset, PROTEUS, empowers users to create optimal reactor designs quickly, reliably and accurately. ...Reducing costs for designers of fast spectrum reactors. PROTEUS' long history of validation provides confidence in predictive simulations Argonne's simulation tools have more than 30 years of validation history against numerous experiments and measurements. The tools within PROTEUS work together, using the same interface files for easier integration of calculations. Multi-group Fast Reactor Cross Section Processing: MC 2 -3 No other fast spectrum multigroup generation tool

179

SUPPLEMENT ANALYSIS OF FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL  

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

FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL TRANSPORTATION ALONG OTHER THAN~. PRESENTATIVE ROUTE FROM CONCORD NAVAL WEAPO~~ STATION TO IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LADORA TORY Introduction The Department of Energy is planning to transport foreign research reactor spent nuclear fuel by rail from the Concord Naval Weapons Station (CNWS), Concord, California, to the Idaho National Engineering and Environmental Laboratory (INEEL). The environmental analysis supporting the decision to transport, by rail or truck, foreign research reactor spent nuclear fuel from CNWS to the INEEL is contained in +he Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliftration Policy Concerning Foreign Research Reactor

180

Light Water Reactor Sustainability Technical Documents | Department of  

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

Reactor Technologies » Light Water Reactor Reactor Technologies » Light Water Reactor Sustainability Program » Light Water Reactor Sustainability Technical Documents Light Water Reactor Sustainability Technical Documents April 30, 2013 LWRS Program and EPRI Long-Term Operations Program - Joint R&D Plan To address the challenges associated with pursuing commercial nuclear power plant operations beyond 60 years, the U.S. Department of Energy's (DOE) Office of Nuclear Energy (NE) and the Electric Power Research Institute (EPRI) have established separate but complementary research and development programs: DOE-NE's Light Water Reactor Sustainability (LWRS) Program and EPRI's Long-Term Operations (LTO) Program. April 30, 2013 Light Water Reactor Sustainability Program - Integrated Program Plan The Light Water Reactor Sustainability (LWRS) Program is a research and

Note: This page contains sample records for the topic "transport reactor integrated" 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

Final analysis of the GCFR radial blanket and shield integral experiment  

SciTech Connect

An integral experiment has been performed for verification of radiation transport methods and nuclear data used in the design of the radial shield for the proposed gas-cooled fast breeder reactor demonstration plant. The experiment was conducted at the ORNL Tower Shielding Facility and consisted of integral and spectral measurements of the neutron and gamma-ray flux transmitted through slabs of materials which modeled a GCFR-type radial blanket and radial shield. Both UO/sub 2/ and ThO/sub 2/ blankets were investigated as well as several shield designs comprising stainless steel, graphite, and boronated graphite.

Ingersoll, D.T.; Williams, L.R.

1981-04-01T23:59:59.000Z

182

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling  

SciTech Connect

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

Koch, M.; Kazimi, M.S.

1991-04-01T23:59:59.000Z

183

Parallel Monte Carlo reactor neutronics  

SciTech Connect

The issues affecting implementation of parallel algorithms for large-scale engineering Monte Carlo neutron transport simulations are discussed. For nuclear reactor calculations, these include load balancing, recoding effort, reproducibility, domain decomposition techniques, I/O minimization, and strategies for different parallel architectures. Two codes were parallelized and tested for performance. The architectures employed include SIMD, MIMD-distributed memory, and workstation network with uneven interactive load. Speedups linear with the number of nodes were achieved.

Blomquist, R.N.; Brown, F.B.

1994-03-01T23:59:59.000Z

184

The development of a remote monitoring system for the Nuclear Science Center reactor.  

E-Print Network (OSTI)

??With funding provided by Nuclear Energy Research Initiative (NERI), design of Secure, Transportable, Autonomous Reactors (STAR) to aid countries with insufficient energy supplies is underway.… (more)

Jiltchenkov, Dmitri Victorovich

2012-01-01T23:59:59.000Z

185

Development of Nuclear Reactor remote Monitoring software (NRM) for the Star project.  

E-Print Network (OSTI)

??As a response to the needs of developing countries to meet their rapidly growing energy requirements, the Safe, Transportable, Autonomous Reactor (STAR) program originated. This… (more)

Gautier, Vincent Charles

2012-01-01T23:59:59.000Z

186

FCT Technology Validation: Integrated Projects  

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

Integrated Projects to Integrated Projects to someone by E-mail Share FCT Technology Validation: Integrated Projects on Facebook Tweet about FCT Technology Validation: Integrated Projects on Twitter Bookmark FCT Technology Validation: Integrated Projects on Google Bookmark FCT Technology Validation: Integrated Projects on Delicious Rank FCT Technology Validation: Integrated Projects on Digg Find More places to share FCT Technology Validation: Integrated Projects on AddThis.com... Home Transportation Projects Stationary/Distributed Generation Projects Integrated Projects DOE Projects Non-DOE Projects Quick Links Hydrogen Production Hydrogen Delivery Hydrogen Storage Fuel Cells Manufacturing Codes & Standards Education Systems Analysis Contacts Integrated Projects To maximize overall system efficiencies, reduce costs, and optimize

187

Transportation Issues and Resolutions Compilation of Laboratory  

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

Transportation Issues and Resolutions Compilation of Laboratory Transportation Issues and Resolutions Compilation of Laboratory Transportation Work Package Reports Transportation Issues and Resolutions Compilation of Laboratory Transportation Work Package Reports The Transportation Team identified the retrievability and subcriticality safety functions to be of primary importance to the transportation of UNF after extended storage and to transportation of high burnup fuel. The tasks performed and described herein address issues related to retrievability and subcriticality; integrity of cladding (embrittled, high burnup cladding, loads applied to cladding during transport), criticality analyses of failed UNF within transport packages, moderator exclusion concepts, stabilization of cladding with canisters for criticality control;

188

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

189

Coal Gasification in a Transport Reactor  

Science Journals Connector (OSTI)

These simulations were used to compare the response of coals gasified to those combusted substoichiometrically, to evaluate the optimum operating conditions and to predict the performance in larger-scale units with less heat loss. ... Entrained-flow gasifiers use high temperatures (1350?1550 °C) and gasify coals in 2?3 s. ... Kinetic studies were carried out to elucidate the mechanisms of steam and CO2 gasification of char and the interactions of these gasifying agents. ...

Lawrence J. Shadle; Esmail R. Monazam; Michael L. Swanson

2001-05-25T23:59:59.000Z

190

Simulation of household in-home and transportation energy use : an integrated behavioral model for estimating energy consumption at the neighborhood scale  

E-Print Network (OSTI)

Household in-home activities and out-of-home transportation are two major sources of urban energy consumption. In light of China's rapid urbanization and income growth, changing lifestyles and consumer patterns - evident ...

Yu, Feifei, S.M. Massachusetts Institute of Technology

2013-01-01T23:59:59.000Z

191

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

192

Influence of seeded impurities on the fusion reactor operation  

Science Journals Connector (OSTI)

In order to study the influence of seeded impurities on ITER like reactor operation the COREDIV code has been extended to include the transport of several sputtered and/or injected impurities. In the COREDIV c...

R. Stankiewicz; R. Zagórski

2004-03-01T23:59:59.000Z

193

Energy Systems Integration  

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

Systems Integration Systems Integration Ben Kroposki, PhD, PE Director, Energy Systems Integration National Renewable Energy Laboratory 2 Reducing investment risk and optimizing systems in a rapidly changing energy world * Increasing penetration of variable RE in grid * Increasing ultra high energy efficiency buildings and controllable loads * New data, information, communications and controls * Electrification of transportation and alternative fuels * Integrating energy storage (stationary and mobile) and thermal storage * Interactions between electricity/thermal/fuels/data pathways * Increasing system flexibility and intelligence Current Energy Systems Future Energy Systems Why Energy Systems Integration? 3 Energy Systems Integration Continuum Scale Appliance (Plug)

194

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

195

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

196

Why Nuclear Energy? - Reactors designed/built by Argonne National  

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

Nuclear Energy: Nuclear Energy: Why Nuclear Energy? 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

197

Proposed design requirements for high-integrity containers used to store, transport, and dispose of high-specific-activity, low-level radioactive wastes from Three Mile Island Unit II  

SciTech Connect

This report develops proposed design requirements for high integrity containers used to store, transport and/or dispose of high-activity, low-level radioactive wastes from Three Mile Island Unit II. The wastes considered are the dewatered resins produced by the EPICOR II waste treatment system used to clean-up the auxiliary building water. The radioactivity level of some of these EPICOR II liners is 1300 curies per container. These wastes may be disposed of in an intermediate depth burial (10 to 20 meter depth) facility. The proposed container design requirements are directed to ensure isolation of the waste and protection of the public health and safety.

Vigil, M.G.; Allen, G.C.; Pope, R.B.

1981-04-01T23:59:59.000Z

198

Light Water Reactor Sustainability (LWRS) Program | Department of Energy  

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

Light Water Reactor Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program The Light Water Reactor Sustainability (LWRS) Program is developing the scientific basis to extend existing nuclear power plant operating life beyond the current 60-year licensing period and ensure long-term reliability, productivity, safety, and security. The program is conducted in collaboration with national laboratories, universities, industry, and international partners. Idaho National Laboratory serves as the Technical Integration Office and coordinates the research and development (R&D) projects in the following pathways: Materials Aging and Degradation Assessment, Advanced Instrumentation, Information, and Control Systems

199

Kinetic model of the MTG process taking into account the catalyst deactivation. Reactor simulation  

Science Journals Connector (OSTI)

A kinetic model for the deactivation of catalyst (based on a HZSM5 zeolite) in the transformation of methanol into gasoline is proposed from results obtained in an isothermal fixed bed integral reactor. The kinetic model takes into account the effect of the composition of the lumps of oxygenates, light olefins and rest of products on the catalyst deactivation along the reactor. The model allows for simulating the integral reactor and for studying the influence of the operating conditions on selectivity towards different lumps in the MTG process. The resukts have been experimentally proven in an isothermal integral reactor and are in agreement with the results of coke deposition along the reactor.

A.G. Gayubo; P.L. Benito; A.T. Aguayo; M. Castilla; J. Bilbao

1996-01-01T23:59:59.000Z

200

Tritium issues in commercial pressurized water reactors  

SciTech Connect

Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

Jones, G. [Constellation Energy Group, R.E. Ginna Nuclear Power Plant, Ontario, NY (United States)

2008-07-15T23:59:59.000Z

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


201

SRS Small Modular Reactors  

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

202

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

203

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

204

Integrated Dynamic Simulation for Process Optimization and Control  

E-Print Network (OSTI)

wherever possible ­ Radiative heat transfer ­ Mass balance ­ Boundary layer transport ­ Surface adsorption;Schematics of Polysilicon RT-CVD Reactor MFC gas cylinder heating lamps RTP reactor RTP pumps 1st stage 2nd-level description ­ Reduced-order models to represent high complexity (e.g., reactor fluid dynamics, heat transfer

Rubloff, Gary W.

205

Biological solutions to transport network design  

Science Journals Connector (OSTI)

...might provide integrated decision...centralized control (Davidson...transport system. Second...difference between treatment (T) and...allometry of plant vascular systems. Nature...removing waste products. Animal and plant transport systems are branching...

2007-01-01T23:59:59.000Z

206

Fluid transport container  

DOE Patents (OSTI)

An improved fluid container for the transport, collection, and dispensing of a sample fluid that maintains the fluid integrity relative to the conditions of the location at which it is taken. More specifically, the invention is a fluid sample transport container that utilizes a fitting for both penetrating and sealing a storage container under controlled conditions. Additionally, the invention allows for the periodic withdrawal of portions of the sample fluid without contamination or intermixing from the environment surrounding the sample container. 13 figs.

DeRoos, B.G.; Downing, J.P. Jr.; Neal, M.P.

1995-11-14T23:59:59.000Z

207

Uranium Oxide Aerosol Transport in Porous Graphite  

SciTech Connect

The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactor’s lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

Blanchard, Jeremy; Gerlach, David C.; Scheele, Randall D.; Stewart, Mark L.; Reid, Bruce D.; Gauglitz, Phillip A.; Bagaasen, Larry M.; Brown, Charles C.; Iovin, Cristian; Delegard, Calvin H.; Zelenyuk, Alla; Buck, Edgar C.; Riley, Brian J.; Burns, Carolyn A.

2012-01-23T23:59:59.000Z

208

Automated Transportation Management System (ATMS) | Department...  

Office of Environmental Management (EM)

of Energy's (DOE's) Automated Transportation Management System is an integrated web-based logistics management system allowing users to manage inbound and outbound freight...

209

An autonomous long-term fast reactor system and the principal design limitations of the concept  

E-Print Network (OSTI)

Actinides MOX Mixed OXide MSR Molten-Salt Reactors NERI Nuclear Energy Research Initiative vii PWR Pressurized Water Reactor RGPu Reactor-Grade Plutonium SCNES Self-Consistent Nuclear Energy System STAR Secure Transportable Autonomous Reactor... of LWR?s, the drastic increase of Am and Cm inventories are observed after uranium fuel irradiation and the second recycling of MOX fuel.1 Therefore, partitioning and transmutation of the recovered MA?s could significantly reduce the long...

Tsvetkova, Galina Valeryevna

2004-09-30T23:59:59.000Z

210

Liquid metal systems development: reactor vessel support structure evaluation  

SciTech Connect

Results of an evaluation of support structures for the reactor vessel are reported. The U ring, box ring, integral ring, tee ring and tangential beam supports were investigated. The U ring is the recommended vessel support structure configuration.

McEdwards, J.A.

1981-01-01T23:59:59.000Z

211

MOCABA: a general Monte Carlo-Bayes procedure for improved predictions of integral functions of nuclear data  

E-Print Network (OSTI)

MOCABA is a combination of Monte Carlo sampling and Bayesian updating algorithms for the prediction of integral functions of nuclear data, such as reactor power distributions or neutron multiplication factors. Similarly to the established Generalized Linear Least Squares (GLLS) methodology, MOCABA offers the capability to utilize integral experimental data to reduce the prior uncertainty of integral observables. The MOCABA approach, however, does not involve any series expansions and, therefore, does not suffer from the breakdown of first-order perturbation theory for large nuclear data uncertainties. This is related to the fact that, in contrast to the GLLS method, the updating mechanism within MOCABA is applied directly to the integral observables without having to "adjust" any nuclear data. A central part of MOCABA is the nuclear data Monte Carlo program NUDUNA, which performs random sampling of nuclear data evaluations according to their covariance information and converts them into libraries for transport code systems like MCNP or SCALE. What is special about MOCABA is that it can be applied to any integral function of nuclear data, and any integral measurement can be taken into account to improve the prediction of an integral observable of interest. In this paper we present two example applications of the MOCABA framework: the prediction of the neutron multiplication factor of a water-moderated PWR fuel assembly based on 21 criticality safety benchmark experiments and the prediction of the power distribution within a toy model reactor containing 100 fuel assemblies.

Axel Hoefer; Oliver Buss; Maik Hennebach; Michael Schmid; Dieter Porsch

2014-11-12T23:59:59.000Z

212

Current Photovoltaics Applications—Transport  

Science Journals Connector (OSTI)

Most transport systems demand high reliability as personal safety is often at stake. System design and overall integrity is thus the primary requirement. These factors are to be considered in detail. Users will d...

P. R. Wolfe

1981-01-01T23:59:59.000Z

213

Transportation Services  

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

Transportation Services Transporting nuclear materials within the United States and throughout the world is a complicated and sometimes highly controversial effort requiring...

214

Local Transportation  

E-Print Network (OSTI)

Local Transportation. Transportation from the Airport to Hotel. There are two types of taxi companies that operate at the airport: special and regular taxis (

215

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

216

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

217

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

218

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

219

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

220

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

Note: This page contains sample records for the topic "transport reactor integrated" 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

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

222

Development of an integrated fission product release and transport code for spatially resolved full-core calculations of V/HTRs  

Science Journals Connector (OSTI)

Abstract The computer codes FRESCO-I, FRESCO-II, PANAMA and SPATRA developed at Forschungszentrum Jülich in Germany in the early 1980s are essential tools to predict the fission product release from spherical fuel elements and the TRISO fuel performance, respectively, under given normal or accidental conditions. These codes are able to calculate a conservative estimation of the source term, i.e. quantity and duration of radionuclide release. Recently, these codes have been reversed engineered, modernized (FORTRAN 95/2003) and combined to form a consistent code named STACY (Source Term Analysis Code System). STACY will later become a module of the V/HTR Code Package (HCP). In addition, further improvements have been implemented to enable more detailed calculations. For example the distinct temperature profile along the pebble radius is now taken into account and coated particle failure rates can be calculated under normal operating conditions. In addition, the absolute fission product release of an V/HTR pebble bed core can be calculated by using the newly developed burnup code Topological Nuclide Transformation (TNT) replacing the former rudimentary approach. As a new functionality, spatially resolved fission product release calculations for normal operating conditions as well as accident conditions can be performed. In case of a full-core calculation, a large number of individual pebbles which follow a random path through the reactor core can be simulated. The history of the individual pebble is recorded, too. Main input data such as spatially resolved neutron fluxes and fluid dynamics data are provided by the VSOP code. Capabilities of the FRESCO-I and SPATRA code which allow for the simulation of the redistribution of fission products within the primary circuit and the deposition of fission products on graphitic and metallic surfaces are also available in STACY. In this paper, details of the STACY model and first results for its application to the 200 MW(th) HTR-Module are presented.

Andre Xhonneux; Hans-Josef Allelein

2014-01-01T23:59:59.000Z

223

Light Water Reactors [Corrosion and Mechanics of Materials] - Nuclear  

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

Light Water Reactors Light Water Reactors Capabilities Materials Testing Environmentally Assisted Cracking (EAC) of Reactor Materials Corrosion Performance/Metal Dusting Overview Light Water Reactors Fatigue Testing of Carbon Steels and Low-Alloy Steels Environmentally Assisted Cracking of Ni-Base Alloys Irradiation-Induced Stress Corrosion Cracking of Austenitic Stainless Steels Steam Generator Tube Integrity Program Air Oxidation Kinetics for Zr-based Alloys Fossil Energy Fusion Energy Metal Dusting Publications List Irradiated Materials Steam Generator Tube Integrity Other Facilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Corrosion and Mechanics of Materials Light Water Reactors Bookmark and Share To continue safe operation of current LWRs, the aging degradation of the

224

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

225

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

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. 1 figure.

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

1994-12-20T23:59:59.000Z

227

Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight  

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

Safety Culture in the US Nuclear Regulatory Commission's Reactor Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process September 19, 2012 Presenter: Undine Shoop, Chief, Health Physics and Human Performance Branch, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission Topics covered: Purpose of the Reactor Oversight Process (ROP) ROP Framework Safety Culture within the ROP Safety Culture Assessments Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process More Documents & Publications A Commissioner's Perspective on USNRC Actions in Response to the Fukushima Nuclear Accident Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11

228

U.S. domestic reactor conversion program  

SciTech Connect

The RERTR U.S. Domestic Conversion program continues in its support of the Global Treat Reduction Initiative (GTRI) to convert seven U.S reactors to low enriched uranium (LEU) by 2010. These reactors are located at the University of Florida, Texas A and M University, Purdue University, Washington State University, Oregon State University, the University of Wisconsin, and the Idaho National Laboratory. The reactors located at the University of Florida and Texas A and M Nuclear Science Center were successfully converted to LEU in September of 2006 through an integrated and collaborative effort involving INL, Argonne National Laboratory (ANL), DOE (headquarters and the field office), the Nuclear Regulatory Commission (NRC), the universities, and the contractors involved in analyses, fuel design and fabrication, and spent nuclear fuel (SNF) shipping and disposition. With this work completed and in anticipation of other impending conversion projects, a meeting was established to engage the project participants in a structured discussion to capture the lessons learned. The objectives of this meeting were to document the observations, insights, issues, concerns, and ideas of those involved in the reactor conversions so that future efforts could be conducted with greater effectiveness, efficiency, and with fewer challenges. The lessons learned from completing the University of Florida and Texas A and M conversions, the Purdue reactor conversion status, and an overview of the upcoming reactor conversions will be presented at the meeting. (author)

Meyer, Dana M.; Woolstenhulme, Eric C. [Idaho National Laboratory, Idaho Falls, Idaho 83415 (United States)

2008-07-15T23:59:59.000Z

229

Subsurface Uranium Fate and Transport: Integrated Experiments and Modeling of Coupled Biogeochemical Mechanisms of Nanocrystalline Uraninite Oxidation by Fe(III)-(hydr)oxides - Project Final Report  

SciTech Connect

Subsurface bacteria including sulfate reducing bacteria (SRB) reduce soluble U(VI) to insoluble U(IV) with subsequent precipitation of UO2. We have shown that SRB reduce U(VI) to nanometer-sized UO2 particles (1-5 nm) which are both intra- and extracellular, with UO2 inside the cell likely physically shielded from subsequent oxidation processes. We evaluated the UO2 nanoparticles produced by Desulfovibrio desulfuricans G20 under growth and non-growth conditions in the presence of lactate or pyruvate and sulfate, thiosulfate, or fumarate, using ultrafiltration and HR-TEM. Results showed that a significant mass fraction of bioreduced U (35-60%) existed as a mobile phase when the initial concentration of U(VI) was 160 µM. Further experiments with different initial U(VI) concentrations (25 - 900 ?M) in MTM with PIPES or bicarbonate buffers indicated that aggregation of uraninite depended on the initial concentrations of U(VI) and type of buffer. It is known that under some conditions SRB-mediated UO2 nanocrystals can be reoxidized (and thus remobilized) by Fe(III)-(hydr)oxides, common constituents of soils and sediments. To elucidate the mechanism of UO2 reoxidation by Fe(III) (hydr)oxides, we studied the impact of Fe and U chelating compounds (citrate, NTA, and EDTA) on reoxidation rates. Experiments were conducted in anaerobic batch systems in PIPES buffer. Results showed EDTA significantly accelerated UO2 reoxidation with an initial rate of 9.5?M day-1 for ferrihydrite. In all cases, bicarbonate increased the rate and extent of UO2 reoxidation with ferrihydrite. The highest rate of UO2 reoxidation occurred when the chelator promoted UO2 and Fe(III) (hydr)oxide dissolution as demonstrated with EDTA. When UO2 dissolution did not occur, UO2 reoxidation likely proceeded through an aqueous Fe(III) intermediate as observed for both NTA and citrate. To complement to these laboratory studies, we collected U-bearing samples from a surface seep at the Rifle field site and have measured elevated U concentrations in oxic iron-rich sediments. To translate experimental results into numerical analysis of U fate and transport, a reaction network was developed based on Sani et al. (2004) to simulate U(VI) bioreduction with concomitant UO2 reoxidation in the presence of hematite or ferrihydrite. The reduction phase considers SRB reduction (using lactate) with the reductive dissolution of Fe(III) solids, which is set to be microbially mediated as well as abiotically driven by sulfide. Model results show the oxidation of HS– by Fe(III) directly competes with UO2 reoxidation as Fe(III) oxidizes HS– preferentially over UO2. The majority of Fe reduction is predicted to be abiotic, with ferrihydrite becoming fully consumed by reaction with sulfide. Predicted total dissolved carbonate concentrations from the degradation of lactate are elevated (log(pCO2) ~ –1) and, in the hematite system, yield close to two orders-of-magnitude higher U(VI) concentrations than under initial carbonate concentrations of 3 mM. Modeling of U(VI) bioreduction with concomitant reoxidation of UO2 in the presence of ferrihydrite was also extended to a two-dimensional field-scale groundwater flow and biogeochemically reactive transport model for the South Oyster site in eastern Virginia. This model was developed to simulate the field-scale immobilization and subsequent reoxidation of U by a biologically mediated reaction network.

Peyton, Brent M. [Montana State University; Timothy, Ginn R. [University of California Davis; Sani, Rajesh K. [South Dakota School of Mines and Technology

2013-08-14T23:59:59.000Z

230

The design of a compact integral medium size PWR : the CIRIS  

E-Print Network (OSTI)

The International Reactor Innovative and Secure (IRIS) is an advanced medium size, modular integral light water reactor design, rated currently at 1000 MWt. IRIS design has been under development by over 20 organizations ...

Shirvan, Koroush

2010-01-01T23:59:59.000Z

231

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),

232

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

233

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

234

Transportation Politics and Policy  

U.S. Energy Information Administration (EIA) Indexed Site

Reducing Greenhouse Reducing Greenhouse Gas Emissions from U.S. Transportation Steven Plotkin, Argonne National Laboratory (co-author is David Greene of Oak Ridge) 2011 EIA Energy Conference May 26-27, 2011 Washington, DC Overview  Presentation based on recent report from the Pew Center on Global Climate Change  Task: Assess the potential to substantially reduce transportation's GHG emissions by 2035 & 2050.  Base Case: Annual Energy Outlook 2010 Reference Case, extended to 2050  Three scenarios with differing assumptions about technological progress, policy initiatives, and public attitudes  Rely on existing studies to estimate impacts  Scenario analysis uses Kaya method to integrate policy impacts and avoid

235

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

236

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

237

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

238

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

239

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data  

SciTech Connect

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Jung, Y. S.; Joo, H. G. [Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of); Yoon, J. I. [KEPCO Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon (Korea, Republic of)

2013-07-01T23:59:59.000Z

240

Benchmark Data Through The International Reactor Physics Experiment Evaluation Project (IRPHEP)  

SciTech Connect

The International Reactor Physics Experiments Evaluation Project (IRPhEP) was initiated by the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency’s (NEA) Nuclear Science Committee (NSC) in June of 2002. The IRPhEP focus is on the derivation of internationally peer reviewed benchmark models for several types of integral measurements, in addition to the critical configuration. While the benchmarks produced by the IRPhEP are of primary interest to the Reactor Physics Community, many of the benchmarks can be of significant value to the Criticality Safety and Nuclear Data Communities. Benchmarks that support the Next Generation Nuclear Plant (NGNP), for example, also support fuel manufacture, handling, transportation, and storage activities and could challenge current analytical methods. The IRPhEP is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and is closely coordinated with the ICSBEP. This paper highlights the benchmarks that are currently being prepared by the IRPhEP that are also of interest to the Criticality Safety Community. The different types of measurements and associated benchmarks that can be expected in the first publication and beyond are described. The protocol for inclusion of IRPhEP benchmarks as ICSBEP benchmarks and for inclusion of ICSBEP benchmarks as IRPhEP benchmarks is detailed. The format for IRPhEP benchmark evaluations is described as an extension of the ICSBEP format. Benchmarks produced by the IRPhEP add new dimension to criticality safety benchmarking efforts and expand the collection of available integral benchmarks for nuclear data testing. The first publication of the "International Handbook of Evaluated Reactor Physics Benchmark Experiments" is scheduled for January of 2006.

J. Blair Briggs; Dr. Enrico Sartori

2005-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport reactor integrated" 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

Aerosol Resuspension Model for MELCOR for Fusion and Very High Temperature Reactor Applications  

SciTech Connect

Dust is generated in fusion reactors from plasma erosion of plasma facing components within the reactor’s vacuum vessel (VV) during reactor operation. This dust collects in cooler regions on interior surfaces of the VV. Because this dust can be radioactive, toxic, and/or chemically reactive, it poses a safety concern, especially if mobilized by the process of resuspension during an accident and then transported as an aerosol though out the reactor confinement building, and possibly released to the environment. A computer code used at the Idaho National Laboratory (INL) to model aerosol transport for safety consequence analysis is the MELCOR code. A primary reason for selecting MELCOR for this application is its aerosol transport capabilities. The INL Fusion Safety Program (FSP) organization has made fusion specific modifications to MELCOR. Recent modifications include the implementation of aerosol resuspension models in MELCOR 1.8.5 for Fusion. This paper presents the resuspension models adopted and the initial benchmarking of these models.

B.J. Merrill

2011-01-01T23:59:59.000Z

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

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

245

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

246

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

247

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

248

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

249

The physics of magnetic fusion reactors  

Science Journals Connector (OSTI)

During the past two decades there have been substantial advances in magnetic fusion research. On the experimental front, progress has been led by the mainline tokamaks, which have achieved reactor-level values of temperature and plasma pressure. Comparable progress, when allowance is made for their smaller programs, has been made in complementary configurations such as the stellarator, reversed-field pinch and field-reversed configuration. In this paper, the status of understanding of the physics of toroidal plasmas is reviewed. It is shown how the physics performance, constrained by technological and economic realities, determines the form of reference toroidal reactors. A comparative study of example reactors is not made, because the level of confidence in projections of their performance varies widely, reflecting the vastly different levels of support which each has received. Success with the tokamak has led to the initiation of the International Thermonuclear Experimental Reactor project. It is designed to produce 1500 MW of fusion power from a deuterium-tritium plasma for pulses of 1000 s or longer and to demonstrate the integration of the plasma and nuclear technologies needed for a demonstration reactor.

John Sheffield

1994-07-01T23:59:59.000Z

250

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...

251

Transportation Agency Tool to Analyze Benefits of Living Snow Fences  

E-Print Network (OSTI)

Transportation Agency Tool to Analyze Benefits of Living Snow Fences 5/31/12 Transportation Agency/31/12 Transportation Agency Tool to Analyze Benefits of Living Snow Fences Center for Integrated Natural Resources, Mobility, & Transportation Authority Benefits, Farmer Costs, & Carbon Impacts Focus Groups and Surveys

Minnesota, University of

252

Parallel Reacting Flow Calculations for Chemical Vapor Deposition Reactor Design 1  

E-Print Network (OSTI)

National Laboratories Albuquerque, NM 87185­1111 (To be published in Proceedings of the International at the synthesis of two important research areas: 3D flow and transport modeling of reactors and the simulationParallel Reacting Flow Calculations for Chemical Vapor Deposition Reactor Design 1 Andrew G

Devine, Karen

253

A three dimensional corner balance method for spatial discretization of the transport equation  

E-Print Network (OSTI)

(ANS) reactor, currently under development at Oak Ridge National Laboratory (ORNL), is an excellent example of a real reactor with a complicated three-dimensional geometry. At ORNL, transport problems for this reactor have been modeled using TWODANT... reactor could be represented by a 45' slice with reflecting boundaries. We began with an S4 quadrature set to permit comparison against the analyses performed by ORNL. SNAC allows for different quadrature sets in each energy group, and numerical...

Richardson, Rebecca Lynn

1994-01-01T23:59:59.000Z

254

Transportation Demand  

Gasoline and Diesel Fuel Update (EIA)

page intentionally left blank page intentionally left blank 69 U.S. Energy Information Administration | Assumptions to the Annual Energy Outlook 2011 Transportation Demand Module The NEMS Transportation Demand Module estimates transportation energy consumption across the nine Census Divisions (see Figure 5) and over ten fuel types. Each fuel type is modeled according to fuel-specific technology attributes applicable by transportation mode. Total transportation energy consumption is the sum of energy use in eight transport modes: light-duty vehicles (cars and light trucks), commercial light trucks (8,501-10,000 lbs gross vehicle weight), freight trucks (>10,000 lbs gross vehicle weight), buses, freight and passenger aircraft, freight and passenger rail, freight shipping, and miscellaneous

255

Office of Secure Transportation Activities  

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

16th, 2012 16th, 2012 WIPP Knoxville, TN OFFICE OF SECURE TRANSPORTATION Agency Integration Briefing Our Mission To provide safe and secure ground and air transportation of nuclear weapons, nuclear weapons components, and special nuclear materials and conduct other missions supporting the national security of the United States of America. 3 5 OST's Commitment to Transportation Safety and Security Over three decades of safe, secure transport of nuclear weapons and special nuclear material to and from military locations and DOE facilities More than 140 million miles traveled Over three decades and 240,000 flight hours of accident-free flying Professionalism We conduct ourselves and our operations with the highest standards of professionalism and integrity.

256

WIPP Transportation  

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

Transuranic Waste Transportation Container Documents Documents related to transuranic waste containers and packages. CBFO Tribal Program Information about WIPP shipments across...

257

Transportation Security  

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

Preliminary Draft - For Review Only 1 Transportation Security Draft Annotated Bibliography Review July 2007 Preliminary Draft - For Review Only 2 Work Plan Task * TEC STG Work...

258

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010  

SciTech Connect

The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

Snead, Lance Lewis [ORNL; Besmann, Theodore M [ORNL; Collins, Emory D [ORNL; Bell, Gary L [ORNL

2010-10-01T23:59:59.000Z

259

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

260

A GASFLOW analysis of a steam explosion accident in a typical light-water reactor confinement building  

SciTech Connect

Steam over-pressurization resulting from ex-vessel steam explosion (fuel-coolant interaction) may pose a serious challenge to the integrity of a typical light-water reactor confinement building. If the steam generation rate exceeds the removal capacity of the Airborne Activity Confinement System, confinement over pressurization occurs. Thus, there is a large potential for an uncontrolled and unfiltered release of fission products from the confinement atmosphere to the environment at the time of the steam explosion. The GASFLOW computer code was used to analyze the effects of a hypothetical steam explosion and the transport of steam and hydrogen throughout a typical light-water reactor confinement building. The effects of rapid pressurization and the resulting forces on the internal structures and the heat exchanger service bay hatch covers were calculated. Pressurization of the ventilation system and the potential damage to the ventilation fans and high-efficiency particulate air filters were assessed. Because of buoyancy forces and the calculated confinement velocity field, the hydrogen diffuses and mixes in the confinement atmosphere but tends to be transported to its upper region.

Travis, J.R. [ESSI Inc. (United States); Wilson, T.L.; Spore, J.W.; Lam, K.L. [Los Alamos National Lab., NM (United States); Rao, D.V. [SEA Inc. (United States)

1994-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport reactor integrated" 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

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

262

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

263

Post-doc: Modelling & Control of Continuous Reactors  

E-Print Network (OSTI)

development in mechatronics and microsystems, sustainable industrial processes, transportation systems scope is a demonstration of continuous reactors with in-line analytics for fine chemical production for improvement of operation and actuation and will provide data for sensor development and control guidelines. 4

Langendoen, Koen

264

Near-Neoclassical Transport & Enhanced Stability  

E-Print Network (OSTI)

magnetic shear configurations are particularly attractive for advanced tokamak reactors -- predicted and advanced tokamak physics Outline · Formation · Transport · MHD Stability · Future Directions TFTR #12;6 5 4 by Shaing and Hazeltine; Hinton and Kim modification of Hirshman-Sigmar equations · comparison with Full

265

NREL: Transportation Research - Systems Analysis and Integration  

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

cell vehicles, and other alternative fuel vehicles. Using a suite of simulation and analysis tools, NREL conducts technical analyses of promising vehicle technologies to find...

266

Transportation Market Distortions  

E-Print Network (OSTI)

of Highways, Volpe National Transportation Systems Center (Evaluating Criticism of Transportation Costing, VictoriaFrom Here: Evaluating Transportation Diversity, Victoria

Litman, Todd

2006-01-01T23:59:59.000Z

267

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

268

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

269

Spring 2010 National Transportation Stakeholder Forum Meetings, Illinois |  

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

0 National 0 National Transportation Stakeholder Forum Meetings, Illinois Spring 2010 National Transportation Stakeholder Forum Meetings, Illinois NTSF Spring 2010 Agenda Final Agenda NTSF Presentations Applying Risk Communication to the Transportation of Radioactive Materials Department of Energy Office of Science Transportation Overview Department of Transportation Pipeline and Hazardous Materials Safety Administration Activities EM Waste and Materials Disposition & Transportation National Transportation Stakeholders Forum Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management Status and Future of TRANSCOM Transportation Emergency Preparedness Program - Making A Difference Waste Isolation Pilot Plant Status and Plans - 2010 Meeting Summary Meeting Summary Notes

270

The development of an integrated multistage fluid bed retorting process. [Kentort II process  

SciTech Connect

This report summarizes the progress made on the development of an integrated multistage fluidized bed retorting process (KENTORT II) during the period of April 1, 1992 through June 30, 1992. The KENTORT II process includes integral fluidized bed zones for pyrolysis, gasification, and combustion of the oil shale. The purpose of this program is to design and test the KENTORT II process at the 50-lb/hr scale. The raw oil shale sample for the program was mined, prepared, characterized and stored this quarter. The shale that was chosen was from the high-grade zone of the Devonian Cleveland Member of the Ohio Shale in Montgomery County, Kentucky. The shale was mined and then transported to the contractor's crushing facility where it was crushed, double-screened, and loaded into 85 55-gal barrels. The barrels, containing a total of 25-30 tons of shale, were transported to the (CAER) Center for Applied Energy Research where the shale was double-screened, analyzed and stored. A major objective of the program is the study of solid-induced secondary coking and cracking reactions. A valved fluidized bed reactor has been the primary apparatus used for this study prior to this quarter, but two additional techniques have been initiated this quarter for the study of other aspects of this issue. First, the two-stage hydropyrolysis reactor at the University of Strathclyde, Glasgow, Scotland, was used to study the coking tendency of shale oil vapors under a wide range of pyrolysis and hydropyrolysis conditions. This work enabled us to examine secondary reactions under high pressure conditions (up to 150 bar) which were previously unavailable. Second, the development of a fixed bed reactor system was initiated at the CAER to study the coking and cracking characteristics of model compounds. A fixed bed apparatus was necessary because the conversion of model compounds was too low in the fluidized bed apparatus.

Carter, S.D.; Taulbee, D.N.; Robl, T.L.; Hower, J.C.

1992-08-01T23:59:59.000Z

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

Parametric study on maximum transportable distance and cost for thermal energy transportation using various coolants  

SciTech Connect

The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as district heating, desalination, hydrogen production and other process heat applications, etc. The process heat industry/facilities will be located outside the nuclear island due to safety measures. This thermal energy from the reactor has to be transported a fair distance. In this study, analytical analysis was conducted to identify the maximum distance that thermal energy could be transported using various coolants such as molten-salts, helium and water by varying the pipe diameter and mass flow rate. The cost required to transport each coolant was also analyzed. The coolants analyzed are molten salts (such as: KClMgCl2, LiF-NaF-KF (FLiNaK) and KF-ZrF4), helium and water. Fluoride salts are superior because of better heat transport characteristics but chloride salts are most economical for higher temperature transportation purposes. For lower temperature water is a possible alternative when compared with He, because low pressure He requires higher pumping power which makes the process very inefficient and economically not viable for both low and high temperature application.

Su-Jong Yoon; Piyush Sabharwall

2014-07-01T23:59:59.000Z

273

OXYGEN TRANSPORT CERAMIC MEMBRANES  

SciTech Connect

In the present quarter, oxygen transport perovskite ceramic membranes are evaluated for strength and fracture in oxygen gradient conditions. Oxygen gradients are created in tubular membranes by insulating the inner surface from the reducing environment by platinum foils. Fracture in these test conditions is observed to have a gradient in trans and inter-granular fracture as opposed to pure trans-granular fracture observed in homogeneous conditions. Fracture gradients are reasoned to be due to oxygen gradient set up in the membrane, variation in stoichiometry across the thickness and due to varying decomposition of the parent perovskite. The studies are useful in predicting fracture criterion in actual reactor conditions and in understanding the initial evolution of fracture processes.

Dr. Sukumar Bandopadhyay; Dr. Nagendra Nagabhushana

2002-07-01T23:59:59.000Z

274

KBR transport gasifier  

SciTech Connect

The KBR Transport Gasifier is an advanced circulating fluidized bed reactor designed to operate at higher circulation rates, velocities and riser densities than a conventional circulating fluidized bed and is based on KBR's extensive fluid bed catalytic cracking experience. The KBR Transport Gasifier is currently being tested at the Power Systems Development Facility (PSDF), an engineering scale demonstration of advanced coal-fired power systems and high temperature, high-pressure gas filtration systems. The KBR Transport Gasifier was operated for three years as a pressurized combustor until coal gasification testing began in September 1999. Through September 2005, the Transport Gasifier has achieved over 7,700 hours of coal gasification. A total of 6,320 hours of gasification were with Powder River Basin coal and 750 hours were with North Dakota lignite. Additional hours were devoted to bituminous coals from Utah, Illinois, Indiana and Alabama. Most testing occurred in air blown gasification mode. It has also been tested for a total of 1,722 hours in oxygen-blown mode. The gasifier has operated at temperatures from 1,500 to 1,950{sup o}F and at pressures of up to 250 psig with coal rates of 2,500 to 5,000 pounds per hour, yielding commercially projected turbine inlet syngas heating values of up to 147 Btu/SCF in air-blown gasification and up to 298 Btu/SCF in oxygen-blown gasification. Carbon conversion has been as high as 98%. 7 refs., 8 figs., 1 tab.

NONE

2005-07-01T23:59:59.000Z

275

Convective equilibrium and mixing-length theory for stellarator reactors  

SciTech Connect

In high ..beta.. stellarator and tokamak reactors, the plasma pressure gradient in some regions of the plasma may exceed the critical pressure gradient set by ballooning instabilities. In these regions, convective cells break out to enhance the transport. As a result, the pressure gradient can rise only slightly above the critical gradient and the plasma is in another state of equilibrium - ''convective equilibrium'' - in these regions. Although the convective transport cannot be calculated precisely, it is shown that the density and temperature profiles in the convective region can still be estimated. A simple mixing-length theory, similar to that used for convection in stellar interiors, is introduced in this paper to provide a qualitative description of the convective cells and to show that the convective transport is highly efficient. A numerical example for obtaining the density and temperature profiles in a stellarator reactor is given.

Ho, D.D.M.; Kulsrud, R.M.

1985-09-01T23:59:59.000Z

276

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

277

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

278

Experimental method for reactor-noise measurements of effective beta. [LMFBR  

SciTech Connect

A variance-to-mean noise technique, modified to eliminate systematic errors from drifting of reactor power, has been used to infer integral values of effective beta for uranium and plutonium fueled fast reactor modk-ups. The measurement technique, including corrections for a finite detector-electrometer time response, is described together with preliminary beta measurement results.

Bennett, E.F.

1981-09-01T23:59:59.000Z

279

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor  

SciTech Connect

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

2010-09-01T23:59:59.000Z

280

1. INTRODUCTION High-energy fusion-product (fp) transport (e.g., alpha particle  

E-Print Network (OSTI)

1 1. INTRODUCTION High-energy fusion-product (fp) transport (e.g., alpha particle transport in D-T plasmas) is a central issue in fusion reactor de- velopment. Important effects dependent on fp transport-7 are concerned with fp wall bombardment and focus on two types of charged, high-energy fp losses from

Hively, Lee M.

Note: This page contains sample records for the topic "transport reactor integrated" 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

Surrogate Spent Nuclear Fuel Vibration Integrity Investigation  

SciTech Connect

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Bevard, Bruce Balkcom [ORNL; Howard, Rob L [ORNL

2014-01-01T23:59:59.000Z

282

INL/EXT-14-33257 Light Water Reactor Sustainability Program  

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

57 Light Water Reactor Sustainability Program 3D J-Integral Capability in Grizzly September 2014 DOE Office of Nuclear Energy DISCLAIMER This information was prepared as an account...

283

Airflow and Pollutant Transport  

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

Computational fluid dynamics flow diagram Computational fluid dynamics flow diagram Airflow and Pollutant Transport Research on airflow and pollutant transport integrates experimental and modeling research in order to understand the dispersion of airborne pollutants in buildings. The work applies to reducing health risks (for example, in the event of a toxic release in an occupied space), as well as to improving energy efficiency and occupant comfort. Investigators also conduct research to characterize and better understand the sources of airborne volatile, semi-volatile and particulate organic pollutants in the indoor environment, and studies of the physical and chemical processes that govern indoor air pollutant concentrations and exposures. The motivation is to contribute to the reduction of potential

284

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

285

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

286

Natural Fueling of a Tokamak Fusion Reactor  

E-Print Network (OSTI)

A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is discussed. In H-mode plasmas dominated by ion- temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is demonstrated using the three-dimensional toroidal electromagnetic gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport.

Wan, Weigang; Chen, Yang; Perkins, Francis W

2009-01-01T23:59:59.000Z

287

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

288

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

289

Simulator platform for fast reactor operation and safety technology demonstration  

SciTech Connect

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

2012-07-30T23:59:59.000Z

290

Integrated Design of Chemical Processes and Utility Systems  

E-Print Network (OSTI)

The pinch concept for integrated heat recovery networks has recently become established in chemical process design. This paper presents an overview of the concept and shows how it has now been extended to total process design (reactors, separators...

Linnhoff, B.

291

Structural Integrity Assessment of Very High Temperature Nuclear...  

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

Structural Integrity Assessment of Very High Temperature Nuclear Reactor Core Components Oct 20 2014 09:00 AM - 10:00 AM Gyanendar Singh, The University of Minnesota, Minneapolis...

292

Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements  

SciTech Connect

The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

J. D. Bess; T. L. Maddock; M. A. Marshall

2011-09-01T23:59:59.000Z

293

Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC  

SciTech Connect

A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area.

Flach, G.P.

2000-02-11T23:59:59.000Z

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

Transportation fuels from synthetic gas  

SciTech Connect

Twenty-five experimental Fischer-Tropsch synthesis runs were made with 14 different catalysts or combinations of catalysts using a Berty reactor system. Two catalysts showed increased selectivity to transportation fuels compared to typical Fischer-Tropsch catalysts. With a catalyst consisting of 5 wt % ruthenium impregnated on a Y zeolite (run number 24), 63 to 70 wt % of the hydrocarbon product was in the gasoline boiling range. Using a 0.5 wt % ruthenium on alumina catalyst (run number 22), 64 to 78 wt % of the hydrocarbon product was in the diesel fuel boiling range. Not enough sample was produced to determine the octane number of the gasoline from run number 24, but it is probably somewhat better than typical Fischer-Tropsch gasoline (approx. 50) and less than unleaded gasoline (approx. 88). The diesel fuel produced in run number 22 consisted of mostly straight chained paraffins and should be an excellent transportation fuel without further refining. The yield of transportation fuels from biomass via gasification and the Fischer-Tropsch synthesis with the ruthenium catalysts identified in the previous paragraph is somewhat less, on a Btu basis, than methanol (via gasification) and wood oil (PERC and LBL processes) yields from biomass. However, the products of the F-T synthesis are higher quality transportation fuels. The yield of transportation fuels via the F-T synthesis is similar to the yield of gasoline via methanol synthesis and the Mobil MTG process.

Baker, E.G.; Cuello, R.

1981-08-01T23:59:59.000Z

296

EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear  

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

2: Urgent-Relief Acceptance of Foreign Research Reactor Spent 2: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel SUMMARY This EA evaluates the environmental impacts of a proposal to accept 409 spent fuel elements from eight foreign research reactors in seven European countries. The spent fuel would be shipped across the ocean in spent fuel transportation casks from the country of origin to one or more United States eastern seaboard ports. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD April 22, 1994 EA-0912: Finding of No Significant Impact Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel April 22, 1994 EA-0912: Final Environmental Assessment Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

297

A Transportation Risk Assessment Tool for Analyzing the Transport of Spent Nuclear Fuel and High-Level Radioactive Waste to the Proposed Yucca Mountain Repository  

SciTech Connect

The Yucca Mountain Transportation Database was developed as a data management tool for assembling and integrating data from multiple sources to compile the potential transportation impacts presented in the Draft Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada (DEIS). The database uses the results from existing models and codes such as RADTRAN, RISKIND, INTERLINE, and HIGHWAY to estimate transportation-related impacts of transporting spent nuclear fuel and high-level radioactive waste from commercial reactors and U. S. Department of Energy (DOE) facilities to Yucca Mountain. The source tables in the database are compendiums of information from many diverse sources including: radionuclide quantities for each waste type; route and route characteristics for rail, legal-weight truck, heavy haul. truck, and barge transport options; state-specific accident and fatality rates for routes selected for analysis; packaging and shipment data by waste type; unit risk factors; the complex behavior of the packaged waste forms in severe transport accidents; and the effects of exposure to radiation or the isotopic specific effects of radionclides should they be released in severe transportation accidents. The database works together with the codes RADTRAN (Neuhauser, et al, 1994) and RISKlND (Yuan, et al, 1995) to calculate incident-free dose and accident risk. For the incident-free transportation scenario, the database uses RADTRAN and RISKIND-generated data to calculate doses to offlink populations, onlink populations, people at stops, crews, inspectors, workers at intermodal transfer stations, guards at overnight stops, and escorts, as well as non-radioactive pollution health effects. For accident scenarios, the database uses RADTRAN-generated data to calculate dose risks based on ingestion, inhalation, resuspension, immersion (cloudshine), and groundshine as well as non-radioactive traffic fatalities. The Yucca Mountain EIS Transportation Database was developed using Microsoft Access 97{trademark} software and the Microsoft Windows NT{trademark} operating system. The database consists of tables for storing data, forms for selecting data for querying, and queries for retrieving the data in a predefined format. Database queries retrieve records based on input parameters and are used to calculate incident-free and accident doses using unit risk factors obtained from RADTRAN results. The next section briefly provides some background that led to the development of the database approach used in preparing the Yucca Mountain DEIS. Subsequent sections provide additional details on the database structure and types of impacts calculated using the database.

Ralph Best; T. Winnard; S. Ross; R. Best

2001-08-17T23:59:59.000Z

298

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

299

verification & Validation of High-Order Short-Characteristics-Based Deterministic Transport Methodology on Unstructured Grids  

SciTech Connect

The research team has developed a practical, high-order, discrete-ordinates, short characteristics neutron transport code for three-dimensional configurations represented on unstructured tetrahedral grids that can be used for realistic reactor physics applications at both the assembly and core levels. This project will perform a comprehensive verification and validation of this new computational tool against both a continuous-energy Monte Carlo simulation (e.g. MCNP) and experimentally measured data, an essential prerequisite for its deployment in reactor core modeling. Verification is divided into three phases. The team will first conduct spatial mesh and expansion order refinement studies to monitor convergence of the numerical solution to reference solutions. This is quantified by convergence rates that are based on integral error norms computed from the cell-by-cell difference between the code’s numerical solution and its reference counterpart. The latter is either analytic or very fine- mesh numerical solutions from independent computational tools. For the second phase, the team will create a suite of code-independent benchmark configurations to enable testing the theoretical order of accuracy of any particular discretization of the discrete ordinates approximation of the transport equation. For each tested case (i.e. mesh and spatial approximation order), researchers will execute the code and compare the resulting numerical solution to the exact solution on a per cell basis to determine the distribution of the numerical error. The final activity comprises a comparison to continuous-energy Monte Carlo solutions for zero-power critical configuration measurements at Idaho National Laboratory’s Advanced Test Reactor (ATR). Results of this comparison will allow the investigators to distinguish between modeling errors and the above- listed discretization errors introduced by the deterministic method, and to separate the sources of uncertainty.

Azmy, Yousry; Wang, Yaqi

2013-12-20T23:59:59.000Z

300

Energy and Transportation Science Division (ETSD)  

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

Contact Us Contact Us Research Groups Building Technologies Research & Integration Fuels, Engines, & Emissions Research Center for Transportation Analysis Center for Sustainable Industry and Manufacturing Working with Us Employment Opportunities Organization Chart ETSD Staff Only Research Groups Building Technologies Research & Integration Fuels, Engines, & Emissions Research Center for Transportation Analysis Center for Sustainable Industry and Manufacturing Energy and Transportation Science Division News and Events Studies quantify the effect of increasing highway speed on fuel economy WUFI ("Warme und Feuchte Instationar," or transient heat and moisture). A family of PC-based software tools jointly developed by Germany's Fraunhofer Institute for Building Physics and ORNL,...

Note: This page contains sample records for the topic "transport reactor integrated" 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

High temperature gas cooled reactor steam-methane reformer design  

SciTech Connect

The concept of the long distance transportation of process heat energy from a High Temperature Gas Cooled Reactor (HTGR) heat source, based on the steam-methane reforming reaction, is being evaluated by the Department of Energy as an energy source/application for use early in the 21st century. This paper summaries the design of a helium heated steam reformer utilized in conjunction with an intermediate loop, 850/degree/C reactor outlet temperature, HTGR process heat plant concept. This paper also discusses various design considerations leading to the mechanical design features, the thermochemical performance, the materials selection and the structural design analysis. 12 refs.

Impellezzeri, J.R.; Drendel, D.B.; Odegaard, T.K.

1981-01-01T23:59:59.000Z

302

Space-reactor electric systems: subsystem technology assessment  

SciTech Connect

This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

Anderson, R.V.; Bost, D.; Determan, W.R.

1983-03-29T23:59:59.000Z

303

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...

304

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

305

A review of experiments and results from the transient reactor test (TREAT) facility.  

SciTech Connect

The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop.

Deitrich, L. W.

1998-07-28T23:59:59.000Z

306

A Generalized Adjoint Framework for Sensitivity and Global Error Estimation in Time-Dependent Nuclear Reactor Simulations 1  

E-Print Network (OSTI)

-Dependent Nuclear Reactor Simulations 1 H. F. Striplinga, , M. Anitescub , M. L. Adamsa aNuclear Engineering Bateman and transport equations, which govern the time-dependent neutronic behavior of a nuclear reactor framework for computing the adjoint variable to nuclear engineering problems gov- erned by a set

Anitescu, Mihai

307

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

308

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

309

Reactor Pressure Vessel Head Packaging & Disposal  

SciTech Connect

Reactor Pressure Vessel (RPV) Head replacements have come to the forefront due to erosion/corrosion and wastage problems resulting from the susceptibility of the RPV Head alloy steel material to water/boric acid corrosion from reactor coolant leakage through the various RPV Head penetrations. A case in point is the recent Davis-Besse RPV Head project, where detailed inspections in early 2002 revealed significant wastage of head material adjacent to one of the Control Rod Drive Mechanism (CRDM) nozzles. In lieu of making ASME weld repairs to the damaged head, Davis-Besse made the decision to replace the RPV Head. The decision was made on the basis that the required weld repair would be too extensive and almost impractical. This paper presents the packaging, transport, and disposal considerations for the damaged Davis-Besse RPV Head. It addresses the requirements necessary to meet Davis Besse needs, as well as the regulatory criteria, for shipping and burial of the head. It focuses on the radiological characterization, shipping/disposal package design, site preparation and packaging, and the transportation and emergency response plans that were developed for the Davis-Besse RPV Head project.

Wheeler, D. M.; Posivak, E.; Freitag, A.; Geddes, B.

2003-02-26T23:59:59.000Z

310

ASME Material Challenges for Advanced Reactor Concepts  

SciTech Connect

This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at higher temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.

Piyush Sabharwall; Ali Siahpush

2013-07-01T23:59:59.000Z

311

Process simulation of refinery units including chemical reactors  

Science Journals Connector (OSTI)

Process simulation methods for design and operation of refinery units are well established as long as no chemical reactors are included. The feedstocks are divided into pseudo-components which enables calculation of phase equilibria and transport properties. When chemical reactors are present some chemical conversion takes place which obviously affects the nature of the pseudo-components and their properties. The stream leaving the reactor will not only be of a different composition than the stream entering the reactor but in addition, the pseudo-components making up the outlet stream will also have other physical properties than the ones in the inlet stream. These changes affect not only the reactor unit but also the simulation of the whole flow-sheet. The paper presents a detailed model for an adiabatic distillate hydrotreater which takes into account the elemental composition of the feed. A special simulation strategy has been developed to incorporate such reactor units into process simulators. Finally, the simulation strategy is illustrated for a hydrotreating plant.

Jens A. Hansen; Barry H. Cooper

1992-01-01T23:59:59.000Z

312

Neurotransmitter Transporters  

E-Print Network (OSTI)

at specialized synaptic junctions where electrical excitability in the form of an action potential is translated membrane of neurons and glial cells. Transporters harness electrochemical gradients to force the movement.els.net #12;The response produced when a transmitter interacts with its receptors, the synaptic potential

Bergles, Dwight

313

Memory effects in turbulent transport  

E-Print Network (OSTI)

In the mean-field theory of magnetic fields, turbulent transport, i.e. the turbulent electromotive force, is described by a combination of the alpha effect and turbulent magnetic diffusion, which are usually assumed to be proportional respectively to the mean field and its spatial derivatives. For a passive scalar there is just turbulent diffusion, where the mean flux of concentration depends on the gradient of the mean concentration. However, these proportionalities are approximations that are valid only if the mean field or the mean concentration vary slowly in time. Examples are presented where turbulent transport possesses memory, i.e. where it depends crucially on the past history of the mean field. Such effects are captured by replacing turbulent transport coefficients with time integral kernels, resulting in transport coefficients that depend effectively on the frequency or the growth rate of the mean field itself. In this paper we perform numerical experiments to find the characteristic timescale (or memory length) of this effect as well as simple analytical models of the integral kernels in the case of passive scalar concentrations and kinematic dynamos. The integral kernels can then be used to find self-consistent growth or decay rates of the mean fields. In mean-field dynamos the growth rates and cycle periods based on steady state values of alpha effect and turbulent diffusivity can be quite different from the actual values.

Alexander Hubbard; Axel Brandenburg

2008-11-17T23:59:59.000Z

314

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

315

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

316

NREL: Transportation Research - News  

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

News NREL provides a number of transportation and hydrogen news sources. Transportation News Find news stories that highlight NREL's transportation research, development, and...

317

The siting of UK nuclear reactors  

Science Journals Connector (OSTI)

Choosing a suitable site for a nuclear power station requires the consideration and balancing of several factors. Some 'physical' site characteristics, such as the local climate and the potential for seismic activity, will be generic to all reactors designs, while others, such as the availability of cooling water, the area of land required and geological conditions capable of sustaining the weight of the reactor and other buildings will to an extent be dependent on the particular design of reactor chosen (or alternatively the reactor design chosen may to an extent be dependent on the characteristics of an available site). However, one particularly interesting tension is a human and demographic one. On the one hand it is beneficial to place nuclear stations close to centres of population, to reduce transmission losses and other costs (including to the local environment) of transporting electricity over large distances from generator to consumer. On the other it is advantageous to place nuclear stations some distance away from such population centres in order to minimise the potential human consequences of a major release of radioactive materials in the (extremely unlikely) event of a major nuclear accident, not only in terms of direct exposure but also concerning the management of emergency planning, notably evacuation.This paper considers the emergence of policies aimed at managing this tension in the UK. In the first phase of nuclear development (roughly speaking 1945–1965) there was a highly cautious attitude, with installations being placed in remote rural locations with very low population density. The second phase (1965–1985) saw a more relaxed approach, allowing the development of AGR nuclear power stations (which with concrete pressure vessels were regarded as significantly safer) closer to population centres (in 'semi-urban' locations, notably at Hartlepool and Heysham). In the third phase (1985–2005) there was very little new nuclear development, Sizewell B (the first and so far only PWR power reactor in the UK) being colocated with an early Magnox station on the rural Suffolk coast. Renewed interest in nuclear new build from 2005 onward led to a number of sites being identified for new reactors before 2025, all having previously hosted nuclear stations and including the semi-urban locations of the 1960s and 1970s. Finally, some speculative comments are made as to what a 'fifth phase' starting in 2025 might look like.

Malcolm Grimston; William J Nuttall; Geoff Vaughan

2014-01-01T23:59:59.000Z

318

Generation of spatial weighting functions for ex-core detectors by adjoint transport calculation  

SciTech Connect

The response rate of ex-core detectors depends on the power level, power distribution, and reactor configuration. For the analysis of the detector response rate for various core power distributions, it is important to generate spatial weighting functions that are insensitive to small changes in the core power distribution and determined by the reactor configuration. Two-dimensional discrete ordinates adjoint transport calculations are used to calculate the core axial weighting functions. The effects of the reactors operating conditions on the core axial weighting functions are analyzed, and it is found that the soluble boron concentration in the reactor coolant has little effect while the core power level affects the core axial weighting functions significantly. A comparison between the results of the adjoint and forward transport calculations shows an excellent agreement. However, the adjoint transport method provides more detailed data and requires less computing time than forward transport method.

Ahn, Joon Gi; Cho, Nam Zin (Korea Advanced Inst. of Science and Tech., Taejon (Korea, Republic of). Dept. of Nuclear Engineering); Kuh, Jung Eui (Korea Atomic Energy Research Inst., Taejon (Korea, Republic of). Safety Analysis Division)

1993-07-01T23:59:59.000Z

319

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

320

Effects of light water reactor coolant environment on the fatigue lives of  

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

Effects of light water reactor coolant environment on the fatigue lives of Effects of light water reactor coolant environment on the fatigue lives of reactor materials July 8, 2013 A metal component can become progressively degraded, and its structural integrity can be adversely impacted when it is subjected to repeated fluctuating loads, or fatigue loading. Fatigue loadings on nuclear reactor pressure vessel components can occur because of changes in pressure and temperature caused by transients during operation, such as reactor startup or shutdown and turbine trip events. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code recognizes fatigue as a possible cause of failure of reactor materials and provides rules for designing nuclear power plant components to avoid fatigue failures. For various materials, the ASME Code defines the

Note: This page contains sample records for the topic "transport reactor integrated" 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

Radiation Transport Calculations and Simulations  

SciTech Connect

This article is an introduction to the Monte Carlo method as used in particle transport. After a description at an elementary level of the mathematical basis of the method, the Boltzmann equation and its physical meaning are presented, followed by Monte Carlo integration and random sampling, and by a general description of the main aspects and components of a typical Monte Carlo particle transport code. In particular, the most common biasing techniques are described, as well as the concepts of estimator and detector. After a discussion of the different types of errors, the issue of Quality Assurance is briefly considered.

Fasso, Alberto; /SLAC; Ferrari, A.; /CERN

2011-06-30T23:59:59.000Z

322

Transportation Security  

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

For Review Only 1 Transportation Security Draft Annotated Bibliography Review July 2007 Preliminary Draft - For Review Only 2 Work Plan Task * TEC STG Work Plan, dated 8/2/06, Product #16, stated: "Develop an annotated bibliography of publicly-available documents related to security of radioactive material transportation." * Earlier this year, a preliminary draft annotated bibliography on this topic was developed by T-REX , UNM, to initially address this STG Work Plan Task. Preliminary Draft - For Review Only 3 Considerations in Determining Release of Information * Some "Publicly-available" documents could potentially contain inappropriate information according to standards set by DOE information security policy and DOE Guides. - Such documents would not be freely

323

Transportation Issues  

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

Issues Issues and Resolutions - Compilation of Laboratory Transportation Work Package Reports Prepared for U.S. Department of Energy Used Fuel Disposition Campaign Compiled by Paul McConnell Sandia National Laboratories September 30, 2012 FCRD-UFD-2012-000342 Transportation Issues and Resolutions ii September 2012 Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any

324

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

325

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

326

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

327

Reactor Safety Research Programs Quarterly Report October - December 1981  

SciTech Connect

This document summarizes the work performed by Pacific Northwest laboratory (PNL) from October 1 through December 31, 1981, for the Division of Accident Evaluation, 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 serviceinduced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-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-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and post accident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S. K.

1982-03-01T23:59:59.000Z

328

Reactor technology assessment and selection utilizing systems engineering approach  

SciTech Connect

The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

Zolkaffly, Muhammed Zulfakar; Han, Ki-In [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-02-12T23:59:59.000Z

329

Core transport studies in fusion devices  

E-Print Network (OSTI)

The turbulence in magnetically confined fusion plasmas has important and non-trivial effects on the quality of the energy confinement. These effects are hard to make a quantitative assessment of analytically. The problem investigated in this article is the transport of energy and particles, in particular impurities, in a Tokamak plasma. Impurities from the walls of the plasma vessel cause energy losses if they reach the plasma core. It is therefore important to understand the transport mechanisms to prevent impurity accumulation and minimize losses. This is an area of research where turbulence plays a major role and is intimately associated with the performance of future fusion reactors, such as ITER.

Strand, Pär; Nordman, Hans

2010-01-01T23:59:59.000Z

330

Policy Research TRANSPORTATION  

E-Print Network (OSTI)

Policy Research TRANSPORTATION CENTER Thestate's transportation system is central to its ability movement of goods to maintain and enhance global economic competitiveness. An effective transportation, TTI has identified the following set of initial transportation issues which must be better understood

331

Cryogenic neutron moderator on mesitylene pellets for IBR-2 reactor  

E-Print Network (OSTI)

Cryogenic neutron moderator on mesitylene pellets for IBR-2 reactor Anan'ev V., Belyakov A mesitylene receiver 1 2 3 4 5 6 7 9 8 HeHe #12;Cryogenic transport pipeline for moderators of 2-3 and 7- 10 of the IBR-2M cryogenic moderator 1 2 3 5 4 1 ­ camera-imitator of cryogenic moderator, 2 ­thermal exchanger

Titov, Anatoly

332

DECAY HEAT CONDITIONS OF CURRENT AND NEXT GENERATION REACTORS  

E-Print Network (OSTI)

(RNSD) of Oak Ridge National Laboratory (ORNL).? Scale has 89 computational modules including 3 deterministic and 3 Monte Carlo radiation transport solvers. These modules are selected based on the user?s desired solution strategy. ?Scale includes... user interfaces to make it easy to use, and also it can plot three-dimensions of the model which helps the user acquire desired results (ORNL, 2011). 5 Analysis In order to simulate the reactor models, Scale requires user inputs including...

Choe, JongSoo 1985-

2012-05-04T23:59:59.000Z

333

Insolation integrator  

DOE Patents (OSTI)

An electric signal representative of the rate of insolation is integrated to determine if it is adequate for operation of a solar energy collection system.

Dougherty, John J. (Norristown, PA); Rudge, George T. (Lansdale, PA)

1980-01-01T23:59:59.000Z

334

Integrated astrophysical modeling  

SciTech Connect

In this project, we have developed prototype techniques for defining and extending a variety of astrophysical modeling capabilities, including those involving multidimensional hydrodynamics, complex transport, and flexibly-coupled equation-of state and nuclear reaction networks. As expected, this project is having both near-term payoffs in understanding complex astrophysical phenomena, as well as significant spin-offs in terms of people and ideas to related ASCI code efforts. Most of our work in the first part of this project was focused on the modularization, extension, and initial integration of 4 previously separate and incommensurate codes: the stellar evolution/explosion code KEPLER; the non-LTE spectral line transport code, EDDINGTON, used for modeling supernovae spectra; the 3-D smooth particle hydro code, PIP; and the discontinuous-finite-element, 3D hydro module from the lCF3D code.

Weaver, T.A., Eastman, R.G., Dubois, P., Eltgroth, P.G., Gentile, N., Jedamzik, K., Wilson, J.R.

1997-06-03T23:59:59.000Z

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

Representative Source Terms and the Influence of Reactor Attributes on Functional Containment in Modular High-Temperature Gas-Cooled Reactors  

SciTech Connect

Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

D. A. Petti; Hans Gougar; Dick Hobbins; Pete Lowry

2013-11-01T23:59:59.000Z

337

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...

338

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.

339

9 - Microporous silica membranes: fundamentals and applications in membrane reactors for hydrogen separation  

Science Journals Connector (OSTI)

Abstract: This chapter discusses the research and development of membrane reactors, incorporating microporous silica-based membranes, specifically for hydrogen production. Microporous silica membranes are first introduced alongside a discussion of relevant gas transport mechanisms, membrane performance parameters, membrane reactor designs and membrane reactor performance metrics. This is followed by an in-depth analysis of the various research investigations where silica membrane reactors have been used to produce hydrogen and/or syngas from hydrocarbon reforming reactions. Of particular importance here is the hydrothermal instability of silica-based membranes at the required operating temperatures and so the chapter closes by presenting the future research trends and industrial design challenges and considerations of silica-based membrane reactors.

S. Smart; J. Beltramini; J.C. Diniz da Costa; S.P. Katikaneni; T. Pham

2013-01-01T23:59:59.000Z

340

Exposure conditions of reactor internals of Rovno VVER-440 NPP units 1 and 2  

SciTech Connect

Results of determination of irradiation conditions for vessel internals of VVER-440 reactor No. 1 and 2 at Rovno Nuclear Power Plant, obtained by specialists at Inst. for Nuclear Research Kyiv (Ukraine)), and Nuclear Research Inst. Rez (Czech Republic)), are presented. To calculate neutron transport, detailed calculation models of these reactors were prepared. Distribution of neutron flux functionals on the surface of reactor VVER-440 baffle and core barrel for different core loads was studied. Agreement between results obtained by specialists at Inst. for Nuclear Research and at Nuclear Research Inst. is shown. (authors)

Grytsenko, O.V.; Pugach, S.M.; Diemokhin, V.L.; Bukanov, V.N. [Inst. for Nuclear Research, Kyiv, 03680 (Ukraine); Marek, M.; Vandlik, S. [Nuclear Research Inst. Rez Plc., Rez, 25068 (Czech Republic)

2011-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport reactor integrated" 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

A fixed-bed reactor modeling study on the methanation of CO2  

Science Journals Connector (OSTI)

Abstract The methanation of carbon dioxide has gained renewed interest during the last years as a possible technology to synthesize a feasible chemical energy carrier. This modeling study aims at a basic understanding of the aspects relevant for designing an externally cooled fixed-bed reactor for the methanation of a pure, stoichiometric feed gas. It is shown that the reaction rates and the exothermicity (?H° = ?165 kJ/mol) prevent a fixed-bed reactor of technical dimensions to be operated at high conversions without runaway of the reactor. The model predictions of differently detailed pseudo-homogeneous reactor models and a heterogeneous reactor model where the intraparticle transport of mass is described according to a dusty-gas approach are compared to assess the needed level of detail in terms of modeling the heat transfer, fluid flow characteristics and transport resistances on the pellet scale. Under specific conditions, intraparticle mass transfer and external heat transfer need to be considered for describing the temperature and concentration profiles adequately. The study is completed by modeling a fixed-bed membrane reactor as an example of a structured reactor that offers improved temperature control by separated and controlled feeding of hydrogen and carbon dioxide.

David Schlereth; Olaf Hinrichsen

2014-01-01T23:59:59.000Z

345

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

346

OXYGEN TRANSPORT CERAMIC MEMBRANES  

SciTech Connect

In the present quarter, the possibility of using a more complex interfacial engineering approach to the development of reliable and stable oxygen transport perovskite ceramic membranes/metal seals is discussed. Experiments are presented and ceramic/metal interactions are characterized. Crack growth and fracture toughness of the membrane in the reducing conditions are also discussed. Future work regarding this approach is proposed are evaluated for strength and fracture in oxygen gradient conditions. Oxygen gradients are created in tubular membranes by insulating the inner surface from the reducing environment by platinum foils. Fracture in these test conditions is observed to have a gradient in trans and inter-granular fracture as opposed to pure trans-granular fracture observed in homogeneous conditions. Fracture gradients are reasoned to be due to oxygen gradient set up in the membrane, variation in stoichiometry across the thickness and due to varying decomposition of the parent perovskite. The studies are useful in predicting fracture criterion in actual reactor conditions and in understanding the initial evolution of fracture processes.

Dr. Sukumar Bandopadhyay; Dr. Nagendra Nagabhushana

2003-01-01T23:59:59.000Z

347

Secondary heat exchanger design and comparison for advanced high temperature reactor  

SciTech Connect

Next generation nuclear reactors such as the advanced high temperature reactor (AHTR) are designed to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. This study considers two different types of heat exchangers - helical coiled heat exchanger and printed circuit heat exchanger - as possible options for the AHTR secondary heat exchangers with distributed load analysis and comparison. Comparison is provided for all different cases along with challenges and recommendations. (authors)

Sabharwall, P. [Idaho National Laboratory, Idaho Falls, ID 83415-3860 (United States); Kim, E. S. [Seoul National Univ., P.O. Box 1625, Idaho Falls, ID 83415-3860 (United States); Siahpush, A.; McKellar, M.; Patterson, M. [Idaho National Laboratory, Idaho Falls, ID 83415-3860 (United States)

2012-07-01T23:59:59.000Z

348

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

349

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

350

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

351

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

352

Metrology/viewing system for next generation fusion reactors  

SciTech Connect

Next generation fusion reactors require accurate measuring systems to verify sub-millimeter alignment of plasma-facing components in the reactor vessel. A metrology system capable of achieving such accuracy must be compatible with the vessel environment of high gamma radiation, high vacuum, elevated temperature, and magnetic field. This environment requires that the system must be remotely deployed. A coherent, frequency modulated laser radar system is being integrated with a remotely operated deployment system to meet these requirements. The metrology/viewing system consists of a compact laser transceiver optics module which is linked through fiber optics to the laser source and imaging units that are located outside of the harsh environment. The deployment mechanism is a telescopic-mast positioning system. This paper identifies the requirements for the International Thermonuclear Experimental Reactor metrology and viewing system, and describes a remotely operated precision ranging and surface mapping system.

Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Menon, M.M. [Oak Ridge National Lab., TN (United States); Dagher, M.A. [Boeing Rocketdyne Div., Canoga Park, CA (United States)

1997-02-01T23:59:59.000Z

353

Generation III reactors safety requirements and the design solutions  

SciTech Connect

Nuclear energy's public acceptance, and hence its development, depends on its safety. As a reactor designer, we will first briefly remind the basic safety principles of nuclear reactors' design. We will then show how the industry, and in particular Areva with its EPR, made design evolution in the wake of the Three Miles Island accident in 1979. In particular, for this new generation of reactors, severe accidents are taken into account beyond the standard design basis accidents. Today, Areva's EPR meets all so-called 'generation III' safety requirements and was licensed by several nuclear safety authorities in the world. Many innovative solutions are integrated in the EPR, some of which will be introduced here.

Felten, P. [Areva NP (France)

2009-03-31T23:59:59.000Z

354

Transportation in Shanghai: A Decision Support System to Move towards Sustainability.  

E-Print Network (OSTI)

??An excellent transportation system is integral for Shanghai as it aims for sustainable development. Decision-making has a far-reaching impact on transportation, which should be improved… (more)

Quchen, Xu; Yanping, Zhuang

2010-01-01T23:59:59.000Z

355

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

356

On-Board Vehicle, Cost Effective Hydrogen Enhancement Technology for Transportation PEM Fuel Cells  

SciTech Connect

Final Report of On-Board Vehicle, Cost Effective Hydrogen Enhancement Technology for Transportation PEM Fuel Cells. The objective of this effort was to technologically enable a compact, fast start-up integrated Water Gas Shift-Pd membrane reactor for integration into an On Board Fuel Processing System (FPS) for an automotive 50 kWe PEM Fuel Cell (PEM FC). Our approach was to: (1) use physics based reactor and system level models to optimize the design through trade studies of the various system design and operating parameters; and (2) synthesize, characterize and assess the performance of advanced high flux, high selectivity, Pd alloy membranes on porous stainless steel tubes for mechanical strength and robustness. In parallel and not part of this program we were simultaneously developing air tolerant, high volumetric activity, thermally stable Water Gas Shift catalysts for the WGS/membrane reactor. We identified through our models the optimum WGS/membrane reactor configuration, and best Pd membrane/FPS and PEM FC integration scheme. Such a PEM FC power plant was shown through the models to offer 6% higher efficiency than a system without the integrated membrane reactor. The estimated FPS response time was < 1 minute to 50% power on start-up, 5 sec transient response time, 1140 W/L power density and 1100 W/kg specific power with an estimated production cost of $35/kW. Such an FPS system would have a Catalytic Partial Oxidation System (CPO) rather than the slower starting Auto-Thermal Reformer (ATR). We found that at optimum WGS reactor configuration that H{sub 2} recovery efficiencies of 95% could be achieved at 6 atm WGS pressure. However optimum overall fuel to net electrical efficiency ({approx}31%) is highest at lower fuel processor efficiency (67%) with 85% H{sub 2} recovery because less parasitic power is needed. The H{sub 2} permeance of {approx}45 m{sup 3}/m{sup 2}-hr-atm{sup 0.5} at 350 C was assumed in these simulations. In the laboratory we achieved a H{sub 2} permeance of 50 m{sup 3}/(m{sup 2}-hr-atm{sup 0.5}) with a H{sub 2}/N{sub 2} selectivity of 110 at 350 C with pure Pd. We also demonstrated that we could produce Pd-Ag membranes. Such alloy membranes are necessary because they aren't prone to the Pd-hydride {alpha}-{beta} phase transition that is known to cause membrane failure in cyclic operation. When funding was terminated we were on track to demonstrated Pd-Ag alloy deposition on a nano-porous ({approx}80 nm) oxide layer supported on porous stainless steel tubing using a process designed for scale-up.

Thomas H. Vanderspurt; Zissis Dardas; Ying She; Mallika Gummalla; Benoit Olsommer

2005-12-30T23:59:59.000Z

357

LEDSGP/Transportation Toolkit/Strategies/Avoid | Open Energy Information  

Open Energy Info (EERE)

LEDSGP/Transportation Toolkit/Strategies/Avoid LEDSGP/Transportation Toolkit/Strategies/Avoid < LEDSGP‎ | Transportation Toolkit‎ | Strategies(Redirected from Transportation Toolkit/Strategies/Avoid) Jump to: navigation, search LEDSGP Logo.png Transportation Toolkit Home Tools Training Contacts Avoid, Shift, Improve Framework The avoid, shift, improve (ASI) framework enables development stakeholders to holistically design low-emission transport strategies by assessing opportunities to avoid the need for travel, shift to less carbon-intensive modes, and improve on conventional technologies, infrastructure, and policies. Avoid Trips and Reduce Travel Demand Transportation Assessment Toolkit Bikes Spain licensed cropped.jpg Avoid trips taken and reduce travel demand by integrating land use planning, transport infrastructure planning, and transport demand

358

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

359

The International Reactor Physics Experiment Evaluation Project (IRPHEP)  

SciTech Connect

Since the beginning of the Nuclear Power industry, numerous experiments concerned with nuclear energy and technology have been performed at different research laboratories, worldwide. These experiments required a large investment in terms of infrastructure, expertise, and cost; however, many were performed without a high degree of attention to archival of results for future use. The degree and quality of documentation varies greatly. There is an urgent need to preserve integral reactor physics experimental data, including measurement methods, techniques, and separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. If the data are compromised, it is unlikely that any of these experiments will be repeated again in the future. The International Reactor Physics Evaluation Project (IRPhEP) was initiated, as a pilot activity in 1999 by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. The purpose of the IRPhEP is to provide an extensively peer reviewed set of reactor physics related integral benchmark data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next generation reactors and establish the safety basis for operation of these reactors. A short history of the IRPhEP is presented and its purposes are discussed in this paper. Accomplishments of the IRPhEP, including the first publication of the IRPhEP Handbook, are highlighted and the future of the project outlined.

J. Blair Briggs; Enrico Sartori; Lori Scott

2006-09-01T23:59:59.000Z

360

The International Reactor Physics Experiment Evaluation Project (IRPhEP)  

SciTech Connect

Since the beginning of the Nuclear Power industry, numerous experiments concerned with nuclear energy and technology have been performed at different research laboratories, worldwide. These experiments required a large investment in terms of infrastructure, expertise, and cost; however, many were performed without a high degree of attention to archival of results for future use. The degree and quality of documentation varies greatly. There is an urgent need to preserve integral reactor physics experimental data, including measurement methods, techniques, and separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. If the data are compromised, it is unlikely that any of these experiments will be repeated again in the future. The International Reactor Physics Evaluation Project (IRPhEP) was initiated, as a pilot activity in 1999 by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. The purpose of the IRPhEP is to provide an extensively peer reviewed set of reactor physics related integral benchmark data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next generation reactors and establish the safety basis for operation of these reactors. A short history of the IRPhEP is presented and its purposes are discussed in this paper. Accomplishments of the IRPhEP, including the first publication of the IRPhEP Handbook, are highlighted and the future of the project outlined. (authors)

Blair Briggs, J. [Idaho National Laboratory, 2525 North Fremont, Idaho Falls, ID 83415-3860 (United States); Sartori, E. [OECD, Nuclear Energy Agency, Le Seine Saint-Germain, 12 Boulevard des Iles, F-92310 Issy-les-Moulineaux (France); Scott, L. [Cover to Cover, 1015 Cedar Hills Blvd., Belle Vernon, PA 15012 (United States)

2006-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport reactor integrated" 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

Advanced thermionic reactor systems design code  

SciTech Connect

An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance.

Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C. (Department of Nuclear Engineering, Radiation Center, C116, Oregon State University, Corvallis, Oregon 97331-5902 (US))

1991-01-01T23:59:59.000Z

362

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

363

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

364

VIM continuous energy Monte Carlo transport code  

SciTech Connect

VIM is a continuous energy neutron and photon transport code. VIM solves the steady-state neutron or photon transport problem in any detailed three-dimensional geometry using either continuous energy-dependent ENDF nuclear data or multigroup cross sections. Neutron transport is carried out in a criticality mode, or in a fixed source mode (optionally incorporating subcritical multiplication). Photon transport is simulated in the fixed source mode. The geometry options are infinite medium, combinatorial geometry, and hexagonal or rectangular lattices of combinatorial geometry unit cells, and rectangular lattices of cells of assembled plates. Boundary conditions include vacuum, specular and white reflection, and periodic boundaries for reactor cell calculations. VIM was developed primarily as a reactor criticality code. Its tally and edit features are very easy to use, and automatically provide fission, fission production, absorption, capture, elastic scattering, inelastic scattering, and (n,2n) reaction rates for each edit region, edit energy group, and isotope, as well as the corresponding macroscopic information, including group scalar fluxes. Microscopic and macroscopic cross sections, including microscopic P{sub N} group-to-group cross sections are also easily produced.

Blomquist, R.N. [Argonne National Lab., IL (United States)

1995-12-31T23:59:59.000Z

365

ORNL/TM-2008/129 Generation IV Reactors Integrated  

E-Print Network (OSTI)

. L. Snead, ORNL W. E. Windes, INL T. E. McGreevy, Caterpillar R. Soto, INL J. K. Wright, INL D. Morgan, University of Wisconsin K. Sridharan, University of Wisconsin R. N. Wright, INL R. K. Nanstad. Ghoniem, UCLA T. L. Sham D. F. Wilson, ORNL Yutai Katoh, ORNL L. L. Snead, ORNL W. E. Windes, INL T. E. Mc

Pennycook, Steve

366

Economic prospects of the Integral Fast Reactor (IFR) fuel cycle  

SciTech Connect

The IFR fuel cycle based on pyroprocessing involves only few operational steps and the batch-oriented process equipment systems are compact. This results in major cost reductions in all of three areas of reprocessing, fabrication, and waste treatment. This document discusses the economic aspects of this fuel cycle.

Chang, Y.I.; Till, C.E.

1991-01-01T23:59:59.000Z

367

Systems Integration  

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

Through the SunShot Initiative, the U.S. Department of Energy (DOE) supports the development of innovative, cost-effective solutions that allow increasing amounts of solar energy to integrate...

368

Integration elements  

Science Journals Connector (OSTI)

Market-based integration is simple: Do whatever you want, the rest is up to the market. This model of the individual and its relation to others best suits the logic of the consumer society ?self-orientation an...

Dr. Eric Dieth

2011-01-01T23:59:59.000Z

369

Procurement Integrity  

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

employment with certain bidders or offerors. This chapter is intended to act as a primer for all DOE employees on issues related to procurement integrity. As such, not all...

370

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

371

Business Case for Early Orders of New Nuclear Reactors Executive Overview  

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

Early Orders of New Nuclear Reactors Early Orders of New Nuclear Reactors Executive Overview Page EX - 1 Business Case for New Business Case for New Nuclear Power Plants Nuclear Power Plants Bringing Public and Private Resources Together for Nuclear Energy Mitigating Critical Risks on Early Orders for New Reactors Briefing for NERAC October 1, 2002 Disclaimer: This draft report was prepared to help the Department of Energy determine the barriers related to the deployment of new nuclear power plants but does not necessarily represent the views or policy of the Department. Business Case for Early Orders of New Nuclear Reactors Executive Overview Page EX - 2 Integrated Project Team Process * Integrated project team (IPT) approach facilitated consideration of complex issues involved in the project and to ensure contractor access to important data from NE.

372

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

373

Chemical factors affecting fission product transport in severe LMFBR accidents  

SciTech Connect

This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly.

Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

1984-10-01T23:59:59.000Z

374

Environmentally assisted cracking of light-water reactor materials  

SciTech Connect

Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator tubing in PWRs and the stress corrosion cracking (SCC) of austenitic SS piping in BWRs have been the most visible and most expensive examples of EAC in LWRs, and the repair and replacement of steam generators and recirculation piping has cost hundreds of millions of dollars. However, other problems associated with the effects of the environment on reactor structures and components am important concerns in operating plants and for extended reactor lifetimes. Cast duplex austenitic-ferritic SSs are used extensively in the nuclear industry to fabricate pump casings and valve bodies for LWRs and primary coolant piping in many PWRs. Embrittlement of the ferrite phase in cast duplex SS may occur after 10 to 20 years at reactor operating temperatures, which could influence the mechanical response and integrity of pressure boundary components during high strain-rate loading (e.g., seismic events). The problem is of most concern in PWRs where slightly higher temperatures are typical and cast SS piping is widely used.

Chopra, O.K.; Chung, H.M.; Kassner, T.F.; Shack, W.J.

1996-02-01T23:59:59.000Z

375

Compact Reversed-Field Pinch Reactors (CRFPR): preliminary engineering considerations  

SciTech Connect

The unique confinement physics of the Reversed-Field Pinch (RFP) projects to a compact, high-power-density fusion reactor that promises a significant reduction in the cost of electricity. The compact reactor also promises a factor-of-two reduction in the fraction of total cost devoted to the reactor plant equipment (i.e., fusion power core (FPC) plus support systems). In addition to operational and developmental benefits, these physically smaller systems can operate economically over a range of total power output. After giving an extended background and rationale for the compact fusion approaches, key FPC subsystems for the Compact RFP Reactor (CRFPR) are developed, designed, and integrated for a minimum-cost, 1000-MWe(net) system. Both the problems and promise of the compact, high-power-density fusion reactor are quantitatively evaluated on the basis of this conceptual design. The material presented in this report both forms a framework for a broader, more expanded conceptual design as well as suggests directions and emphases for related research and development.

Hagenson, R.L.; Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Embrechts, M.J.; Schnurr, N.M.; Battat, M.E.; LaBauve, R.J.; Davidson, J.W.

1984-08-01T23:59:59.000Z

376

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

377

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

378

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)...

379

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...

380

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

Note: This page contains sample records for the topic "transport reactor integrated" 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

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

382

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

383

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...

384

Metropolitan Transport Planning Collaboration in Decentralized Indonesia. Case Study of Greater Yogyakarta.  

E-Print Network (OSTI)

??Indonesia has witnessed the emergence of metropolitan areas whose boundaries stretch beyond administratively defined local authorities. This prompts the need of integrated metropolitan transport planning… (more)

Diwangkari, Andyan

2014-01-01T23:59:59.000Z

385

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

386

NREL: Energy Systems Integration - Integrated Deployment Workshop  

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

Integrated Deployment Workshop Integrated Deployment Workshop The Energy Systems Integration Facility workshop, Integrated Deployment, was held August 21 - 23, 2012 at the National Renewable Energy Laboratory in Golden, Colorado. Each day of the workshop, which included a tour of the Energy Systems Integration Facility, focused on a different topic: Day 1: Utility-Scale Renewable Integration Day 2: Distribution-Level Integration Day 3: Isolated and Islanded Grid Systems The agenda and presentations from the workshop are below. Agenda Energy Systems Integration Facility Overview ESIF Technology Partnerships Integrated Deployment Model Integrated Deployment and the Energy Systems Integration Facility: Workshop Proceedings Printable Version Energy Systems Integration Home Research & Development

387

Transportation Licenses Available | Tech Transfer | ORNL  

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

Transportation Transportation SHARE Transportation 200301316 Integrated Inverter Control for Multiple Electric Machines 200701874 Power Conversion Apparatus and Method for Hybrid Electric and Electric Vehicle Engines 200701929 Rapid Separation and Detection of Hydrocarbon Fractions in an Exhaust Stream 200701962 Sensor Rapidly Measures the Concentration of Oxygen in Fluids 200802045 Failure Prediction of Complex Structures Under Loading and Environmental Stress 200802130 Optical Backscatter Probe for Sensing Particulate Matter 200802165 Optimized Fuel Formulation and Engine Control Parameters for Advanced Diesel Engine 200802180 Enhancements to WIM for Ease of Use - Modifications and/or Elimination of Load Cell Foot Strap 200802188 Highly efficient 6-stroke engine cycle with water

388

DOE-Idaho's Packaging and Transportation Perspective  

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

Idaho's Packaging and Idaho's Packaging and T t ti P ti Transportation Perspective Richard Provencher Manager DOE Idaho Operations Office DOE Idaho Operations Office Presented to the DOE National Transportation Stakeholders Forum Stakeholders Forum May 12, 2011 DOE's Idaho site ships and receives a wide variety of radioactive materials 2 Engineering Test Reactor vessel excavated, transported across the site and disposed 3 Navy SNF moved from wet to dry storage storage 4 5 Left: Contact-handled TRU shipments Right: A remote-handled TRU shipment Right: A remote handled TRU shipment 6 NAC spent nuclear fuel container on its trailer, prior to installation of its impact limiters 7 Examples of dry (CPP-603) and wet (CPP-666) storage in Idaho (CPP 666) storage in Idaho 8 INL's Materials and Fuels Complex Hot Fuel Examination

389

Ion transport membrane module and vessel system  

DOE Patents (OSTI)

An ion transport membrane system comprising (a) a pressure vessel having an interior, an exterior, an inlet, and an outlet; (b) a plurality of planar ion transport membrane modules disposed in the interior of the pressure vessel and arranged in series, each membrane module comprising mixed metal oxide ceramic material and having an interior region and an exterior region, wherein any inlet and any outlet of the pressure vessel are in flow communication with exterior regions of the membrane modules; and (c) one or more gas manifolds in flow communication with interior regions of the membrane modules and with the exterior of the pressure vessel. The ion transport membrane system may be utilized in a gas separation device to recover oxygen from an oxygen-containing gas or as an oxidation reactor to oxidize compounds in a feed gas stream by oxygen permeated through the mixed metal oxide ceramic material of the membrane modules.

Stein, VanEric Edward (Allentown, PA); Carolan, Michael Francis (Allentown, PA); Chen, Christopher M. (Allentown, PA); Armstrong, Phillip Andrew (Orefield, PA); Wahle, Harold W. (North Canton, OH); Ohrn, Theodore R. (Alliance, OH); Kneidel, Kurt E. (Alliance, OH); Rackers, Keith Gerard (Louisville, OH); Blake, James Erik (Uniontown, OH); Nataraj, Shankar (Allentown, PA); Van Doorn, Rene Hendrik Elias (Obersulm-Willsbach, DE); Wilson, Merrill Anderson (West Jordan, UT)

2012-02-14T23:59:59.000Z

390

Ion transport membrane module and vessel system  

DOE Patents (OSTI)

An ion transport membrane system comprising (a) a pressure vessel having an interior, an exterior, an inlet, and an outlet; (b) a plurality of planar ion transport membrane modules disposed in the interior of the pressure vessel and arranged in series, each membrane module comprising mixed metal oxide ceramic material and having an interior region and an exterior region, wherein any inlet and any outlet of the pressure vessel are in flow communication with exterior regions of the membrane modules; and (c) one or more gas manifolds in flow communication with interior regions of the membrane modules and with the exterior of the pressure vessel.The ion transport membrane system may be utilized in a gas separation device to recover oxygen from an oxygen-containing gas or as an oxidation reactor to oxidize compounds in a feed gas stream by oxygen permeated through the mixed metal oxide ceramic material of the membrane modules.

Stein, VanEric Edward (Allentown, PA); Carolan, Michael Francis (Allentown, PA); Chen, Christopher M. (Allentown, PA); Armstrong, Phillip Andrew (Orefield, PA); Wahle, Harold W. (North Canton, OH); Ohrn, Theodore R. (Alliance, OH); Kneidel, Kurt E. (Alliance, OH); Rackers, Keith Gerard (Louisville, OH); Blake, James Erik (Uniontown, OH); Nataraj, Shankar (Allentown, PA); van Doorn, Rene Hendrik Elias (Obersulm-Willsbach, DE); Wilson, Merrill Anderson (West Jordan, UT)

2008-02-26T23:59:59.000Z

391

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

392

Transportation Security | ornl.gov  

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

Transportation Security SHARE Global Threat Reduction Initiative Transportation Security Cooperation Secure Transport Operations (STOP) Box Security of radioactive material while...

393

Transportation Security | Department of Energy  

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

Transportation Security Transportation Security Transportation Security More Documents & Publications Overview for Newcomers West Valley Demonstration Project Low-Level Waste...

394

Sequestration and Transport of Lignin Monomeric Precursors  

SciTech Connect

Lignin is the second most abundant terrestrial biopolymer after cellulose. It is essential for the viability of vascular plants. Lignin precursors, the monolignols, are synthesized within the cytosol of the cell. Thereafter, these monomeric precursors are exported into the cell wall, where they are polymerized and integrated into the wall matrix. Accordingly, transport of monolignols across cell membranes is a critical step affecting deposition of lignin in the secondarily thickened cell wall. While the biosynthesis of monolignols is relatively well understood, our knowledge of sequestration and transport of these monomers is sketchy. In this article, we review different hypotheses on monolignol transport and summarize the recent progresses toward the understanding of the molecular mechanisms underlying monolignol sequestration and transport across membranes. Deciphering molecular mechanisms for lignin precursor transport will support a better biotechnological solution to manipulate plant lignification for more efficient agricultural and industrial applications of cell wall biomass.

Liu, C.J.; Miao, Y.-C.; Zhang, K.-W.

2011-01-18T23:59:59.000Z

395

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

396

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

397

Strategic Freight Transportation Contract Procurement  

E-Print Network (OSTI)

Based Procurement for Transportation Services, Journal ofCoia, A. , Evolving transportation exchanges, World trade,an Auction Based Transportation Marketplace, Transportation

Nandiraju, Srinivas

2006-01-01T23:59:59.000Z

398

"Educating transportation professionals."  

E-Print Network (OSTI)

"Educating transportation professionals." Michael Demetsky Henry L. Kinnier Professor mjd of Virginia Charlottesville, VA 434.924.7464 Transportation Engineering & Management Research Our group works closely with the Virginia Center for Transportation Innovation and Research (VCTIR), located

Acton, Scott

399

Dynamics of reverse osmosis in a standalone cogenerative nuclear reactor (Part I: reactivity changes)  

Science Journals Connector (OSTI)

The present study considers the dynamic behaviour of the pressurised water reactor safety features, represented by the integrity of the fuel cladding, under some transient cases. A cosine-shaped heating through the fuel is taken with the corresponding coolant lumps, to simulate realistic cases encountered in nuclear reactors. A mathematical model was developed for the Westinghouse 3411 MWth pressurised water reactor, as an example of a familiar design with predominantly published data design. The model consists of two parts. The first part is concerned with the dynamics of the primary side of the reactor, which is described in this paper. The second part is concerned with the secondary side of the plant, which is described elsewhere in this issue. To study the dynamics of the reactor, a model of 17 lumped parameters was used, consisting of first-order differential equations deduced from the first principles considering six groups of delayed neutrons. A computer program was developed using the Runge-Kutta method to solve these equations and to predict the behaviour of the state variables with time. Two case studies were considered as examples for normal transients. The first case study, which represents Part 1 of this study, considers the effect of primary side transient on the system as the reactivity changes. Reactor reactivity changes, including movements of the reactor control rods, which are taken as an example for the effect of the reactor primary side conditions. These reactivity changes vary from 0.0005 up to 0.003, both for positive and negative reactivity. The results of the developed model, which describe the dynamic response of the reactor primary circuit, have been analysed and verified with the relevant models. These results indicate that the reactor components and the integrity of the fuel cladding were attained during different step changes of reactivity.

Aly Karameldin; M.M. Shamloul; M.R. Shaalan; M.H. Esawy

2007-01-01T23:59:59.000Z

400

On-Site Oxy-Lance Size Reduction of South Texas Project Reactor Vessel Heads - 12324  

SciTech Connect

On-Site Oxy-Lance size reduction of mildly radioactive large components has been accomplished at other operating plants. On-Site Oxy-Lance size reduction of more radioactive components like Reactor Vessel Heads had previously been limited to decommissioning projects. Building on past decommissioning and site experience, subcontractors for South Texas Project Nuclear Operating Company (STPNOC) developed an innovative integrated system to control smoke, radioactive contamination, worker dose, and worker safety. STP's innovative, easy to use CEDM containment that provided oxy lance access, smoke control, and spatter/contamination control was the key to successful segmentation for cost-effective and ALARA packaging and transport for disposal. Relative to CEDM milling, STP oxy-lance segmentation saved approximately 40 person- REM accrued during 9,000 hours logged into the radiological controlled area (RCA) during more than 3,800 separate entries. Furthermore there were no personnel contamination events or respiratory uptakes of radioactive material during the course of the entire project. (authors)

Posivak, Edward [WMG, inc. (United States); Keeney, Gilbert; Wheeler, Dean [Shaw Group (United States)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport reactor integrated" 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

New Reactor Physics Benchmark Data in the March 2012 Edition of the IRPhEP Handbook  

SciTech Connect

The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the experimental data are lost, the high cost to repeat many of these measurements may be prohibitive. The purpose of the IRPhEP is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [1]. Several new evaluations have been prepared for inclusion in the March 2012 edition of the IRPhEP Handbook.

John D. Bess; J. Blair Briggs; Jim Gulliford

2012-11-01T23:59:59.000Z

402

Transportation Emergency Preparedness Program (TEPP) | Department of Energy  

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

Transportation Emergency Preparedness Program (TEPP) Transportation Emergency Preparedness Program (TEPP) Transportation Emergency Preparedness Program (TEPP) In an effort to address responder concerns, the Department retooled its approach to emergency responder preparedness and implemented the more simplified and responder-friendly Transportation Emergency Preparedness Program (TEPP). TEPP is a component of the overall comprehensive emergency management system established by DOE Order (DOE O) 151.1, Comprehensive Emergency Management System. TEPP integrates a basic approach to transportation emergency planning and preparedness activities under a single program with the goal to ensure DOE, its operating contractors, and state, tribal, and local emergency responders are prepared to respond promptly, efficiently, and effectively to accidents involving DOE

403

Hazardous Waste Transporter Permits (Connecticut) | Department of Energy  

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

Hazardous Waste Transporter Permits (Connecticut) Hazardous Waste Transporter Permits (Connecticut) Hazardous Waste Transporter Permits (Connecticut) < Back Eligibility Agricultural Commercial Construction Fed. Government Fuel Distributor General Public/Consumer Industrial Installer/Contractor Institutional Investor-Owned Utility Local Government Low-Income Residential Multi-Family Residential Municipal/Public Utility Nonprofit Residential Retail Supplier Rural Electric Cooperative Schools State/Provincial Govt Systems Integrator Transportation Tribal Government Utility Program Info State Connecticut Program Type Siting and Permitting Provider Department of Energy and Environmental Protection Transportation of hazardous wastes into or through the State of Connecticut requires a permit. Some exceptions apply. The regulations provide

404

Transportation of Natural Gas and Petroleum (Nebraska) | Department of  

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

Transportation of Natural Gas and Petroleum (Nebraska) Transportation of Natural Gas and Petroleum (Nebraska) Transportation of Natural Gas and Petroleum (Nebraska) < Back Eligibility Agricultural Commercial Construction Fed. Government Fuel Distributor General Public/Consumer Industrial Installer/Contractor Institutional Investor-Owned Utility Local Government Low-Income Residential Multi-Family Residential Municipal/Public Utility Nonprofit Residential Retail Supplier Rural Electric Cooperative Schools State/Provincial Govt Systems Integrator Transportation Tribal Government Utility Program Info State Nebraska Program Type Siting and Permitting Provider Oil and Gas Conservation Commission This statute enables and regulates the exercise of eminent domain by persons, companies, corporations, or associations transporting crude oil,

405

Analytical modeling of contaminant transport and horizontal well hydraulics  

E-Print Network (OSTI)

transport from one-, two-, and three-dimensional finite sources in a finite-thickness aquifer using Green's function method. A library of unpublished analytical solutions with different finite source geometry is provided. A graphically integrated software...

Park, Eungyu

2004-09-30T23:59:59.000Z

406

Transportation Efficiency Resources  

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

Transportation efficiency reduces travel demand as measured by vehicle miles traveled (VMT). While transportation efficiency policies are often implemented under local governments, national and...

407

Transportation and its Infrastructure  

E-Print Network (OSTI)

cost to mitigate transport’s GHG emissions. There are alsoenergy consumption and GHG mitigation, especially inParis, 2005. ECON, 2003: GHG Emissions from International

2007-01-01T23:59:59.000Z

408

Transportation and its Infrastructure  

E-Print Network (OSTI)

Transport and its infrastructure Chapter 5 Hybrid vehiclesincluding hybrid- Transport and its infrastructure Chapter 5infrastructure Gt CO 2 -eq 1 - Diesels (LDVs) 2 - Hybrids (

2007-01-01T23:59:59.000Z

409

Nonlinear dynamics and stability of boiling water reactors: Part 2 - Quantitative analysis  

SciTech Connect

A physical model of nonlinear boiling water reactor (BWR) dynamics has been developed and employed to calculate the amplitude of limit cycle oscillations and their effects on fuel integrity over a wide range of operating conditions in the Vermont Yankee reactor. These calculations have confirmed that, beyond the threshold for linear stability, the reactor's state variables undergo limit cycle oscillations. This work shows that the amplitudes of these oscillations are very sensitive to changes in operating conditions and are not restricted to small magnitudes as observed in previous stability tests. Consequently, large-amplitude limit cycle oscillations become a possible scenario for BWR operation at low-flow conditions. The effects on fuel integrity of such large-amplitude oscillations have been studied in detail. In particular, it has been shown that limit cycles that oscillate with frequencies higher than 0.25 Hz and that reach the high-power safety scram level of 120 % are not likely to compromise fuel integrity.

March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

1986-06-01T23:59:59.000Z

410

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

411

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

412

integr~1  

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

7 7 AUDIT REPORT THE U.S. DEPARTMENT OF ENERGY' S MANAGEMENT OF RESEARCH AND DEVELOPMENT INTEGRATION MARCH 1998 U.S. DEPARTMENT OF ENERGY OFFICE OF INSPECTOR GENERAL OFFICE OF AUDIT SERVICES DEPARTMENT OF ENERGY Washington, DC 20585 MEMORANDUM FOR THE SECRETARY FROM: Gregory H. Friedman Principal Deputy Inspector General SUBJECT: INFORMATION : Audit Report on "Audit of the Department of Energy's Management of Research and Development Integration" BACKGROUND The Congress, independent task forces, and advisory groups have pointed out the need for the Department to improve its integration of research and development (R&D) projects. In the past, R&D management was carried out by different program offices with the research being

413

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,

414

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

415

Considerations Associated with Reactor Technology Selection for the Next Generation Nuclear Plant Project  

SciTech Connect

At the inception of the Next Generation Nuclear Plant Project and during predecessor activities, alternative reactor technologies have been evaluated to determine the technology that best fulfills the functional and performance requirements of the targeted energy applications and market. Unlike the case of electric power generation where the reactor performance is primarily expressed in terms of economics, the targeted energy applications involve industrial applications that have specific needs in terms of acceptable heat transport fluids and the associated thermodynamic conditions. Hence, to be of interest to these industrial energy applications, the alternative reactor technologies are weighed in terms of the reactor coolant/heat transport fluid, achievable reactor outlet temperature, and practicality of operations to achieve the very high reliability demands associated with the petrochemical, petroleum, metals and related industries. These evaluations have concluded that the high temperature gas-cooled reactor (HTGR) can uniquely provide the required ranges of energy needs for these target applications, do so with promising economics, and can be commercialized with reasonable development risk in the time frames of current industry interest – i.e., within the next 10-15 years.

L.E. Demick

2010-09-01T23:59:59.000Z

416

ITM Syngas and ITM H2: Engineering Development of Ceramic Membrane Reactor Systems for  

E-Print Network (OSTI)

(U.S. DOE) and other members of the ITM Syngas/ITM H2 Team, is developing Ion Transport Membrane (ITM of the ITM membrane to oxygen ions, which diffuse through the membrane under a chemical potential gradientITM Syngas and ITM H2: Engineering Development of Ceramic Membrane Reactor Systems for Converting

417

THE RADIATION SAFETY INFORMATION COMPUTATIONAL CENTER: A RESOURCE FOR REACTOR DOSIMETRY SOFTWARE AND NUCLEAR DATA  

SciTech Connect

The Radiation Safety Information Computational Center (RSICC) was established in 1963 to collect and disseminate computational nuclear technology in the form of radiation transport, shielding and safety software and corresponding nuclear cross sections. Approximately 1700 nuclear software and data packages are in the RSICC collection, and the majority are applicable to reactor dosimetry.

Kirk, Bernadette Lugue [ORNL] [ORNL

2009-01-01T23:59:59.000Z

418

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

419

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

420

Steam Generator Tube Integrity Program [Corrosion and Mechanics of  

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

Steam Generator Tube Steam Generator Tube Integrity Program Capabilities Materials Testing Environmentally Assisted Cracking (EAC) of Reactor Materials Corrosion Performance/Metal Dusting Overview Light Water Reactors Fatigue Testing of Carbon Steels and Low-Alloy Steels Environmentally Assisted Cracking of Ni-Base Alloys Irradiation-Induced Stress Corrosion Cracking of Austenitic Stainless Steels Steam Generator Tube Integrity Program Air Oxidation Kinetics for Zr-based Alloys Fossil Energy Fusion Energy Metal Dusting Publications List Irradiated Materials Steam Generator Tube Integrity Other Facilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Corrosion and Mechanics of Materials Light Water Reactors Bookmark and Share

Note: This page contains sample records for the topic "transport reactor integrated" 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

Integrated System  

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

Integrated Window System Our research activities in the field of high performance windows have led us to conclude that even by using high performance insulating glass units, low conductivity frames, and warm edge spacers, there are still untapped sources for improving energy efficiency in the design and use of residential windows. While such high performance windows are a dramatic improvement over conventional units, they do not reduce conductive losses through wall framing around the window, offer guarantees against excessive wall/window infiltration nor do they adapt to the daily and seasonal potentials for night insulation and summer shading. To meet this need, we have been working on the design, development, and prototyping of Integrated Window Systems (IWS) since 1993. Integrated Window Systems are a form of panelized construction where the wall panel includes an operable or fixed window sash, recessed night insulation, integral solar shading, and is built in a factory setting in order to minimize thermal short circuits and infiltration at joints. IWSs can be built in modular lengths to facilitate their installation with conventional wood frame stick construction or other forms of panelized construction.

422

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

423

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

424

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

425

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

426

An Integrated Marine Propulsion System Utilising TRIGA{sup TM} Fuel  

SciTech Connect

This paper describes the reactor physics, shielding, thermal hydraulics, reactor dynamics and safety studies conducted to develop a proposed Integrated Marine Propulsion System (IMPS) utilising TRIGA{sup TM} type uranium zirconium hydride fuel. The study has demonstrated that the IMPS plant is feasible and meets the design safety principles and safety criteria imposed on the study. (authors)

Manach, G.; Monnez, J-P. [Ecole des Applications Militaires de l'Energie Atomique, Cherbourg (France); Freeman, M.J.; Newell, A.; Brushwood, J.M.; Thompson, A.; Collins, C.; Scholes, N.; Hamilton, P.J.; Beeley, P.A. [Nuclear Department/ Flagship Training Ltd, HMS Sultan, Military Road, Gosport, PO12 3BY (United Kingdom)

2004-07-01T23:59:59.000Z

427

11 California Petroleum Supply, Transportation, Refining and Marketing Trends  

E-Print Network (OSTI)

11 California Petroleum Supply, Transportation, Refining and Marketing Trends Chapter 2 CALIFORNIA PETROLEUM SUPPLY, TRANSPORTATION, REFINING AND MARKETING TRENDS INTRODUCTION California is an integral part of the world oil market as a world-scale petroleum consumer. Historically, about 50 percent of this petroleum

428

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

429

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

430

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

431

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

432

TRANSPORTATION Annual Report  

E-Print Network (OSTI)

2003 CENTER FOR TRANSPORTATION STUDIES Annual Report #12;Center for Transportation Studies University of Minnesota 200 Transportation and Safety Building 511 Washington Avenue S.E. Minneapolis, MN publication is a report of transportation research, education, and outreach activities for the period July

Minnesota, University of

433

Career Map: Transportation Worker  

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

The Wind Program's Career Map provides job description information for Transportation Worker positions.

434

Transportation Organization and Functions  

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

Office of Packaging and Transportation list of organizations and functions, with a list of acronyms.

435

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

436

Key Thermal Fluid Phenomena In Prismatic Gas-Cooled Reactors  

SciTech Connect

Several types of gas-cooled nuclear reactors have been suggested as part of the international Generation IV initiative with the proposed NGNP (Next Generation Nuclear Plant) as one of the main concepts [MacDonald et al., 2003]. Meaningful studies for these designs will require accurate, reliable predictions of material temperatures to evaluate the material capabilities; these temperatures depend on the thermal convection in the core and in other important components. Some of these reactors feature complex geometries and wide ranges of temperatures, leading to significant variations of the gas thermodynamic and transport properties plus possible effects of buoyancy during normal and reduced power operations and loss-of-flow (LOFA) and loss-of-coolant scenarios. Potential issues identified to date include ''hot streaking'' in the lower plenum evolving from ''hot channels'' in prismatic cores. In order to predict thermal hydraulic behavior of proposed designs effectively and efficiently, it is useful to identify the dominant phenomena occurring.

D. M. McEligot; G. E. McCreery; P. D. Bayless; T. D. Marshall

2005-06-01T23:59:59.000Z

437

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

438

Fast neutron fluence of yonggwang nuclear unit 1 reactor pressure vessel  

SciTech Connect

The Code of Federal Regulations, Title 10, Part 50, Appendix H, requires that the neutron dosimetry be present to monitor the reactor vessel throughout plant life. The Ex-Vessel Neutron Dosimetry System has been installed for Yonggwang Nuclear Unit 1 after complete withdrawal of all six in-vessel surveillance capsules. This system has been installed in the reactor cavity annulus in order to measure the fast neutron spectrum coming out through the reactor pressure vessel. Cycle specific neutron transport calculations were performed to obtain the energy dependent neutron flux throughout the reactor geometry including dosimetry positions. Comparisons between calculations and measurements were performed for the reaction rates of each dosimetry sensors and results show good agreements. (authors)

Yoo, C.; Km, B.; Chang, K.; Leeand, S. [Korea Atomic Energy Research Inst., 150 Dukjin-dong, Yuseung-gu, Daejeon 305-353 (Korea, Republic of); Park, J. [Chungnam National Univ., 220 Gung-dong, Yuseung-gu, Daejeon 305-764 (Korea, Republic of)

2006-07-01T23:59:59.000Z

439

Multi-modal Transportation > Highway Transportation > Trucking > Railroad transportation > Public transit > Rural transportation > Rural transit > Freig pipeline transportation > Airport planning and development > Airport maintenance > Bicycle and pedestr  

E-Print Network (OSTI)

Multi-modal Transportation > Highway Transportation > Trucking > Railroad transportation > Public transit > Rural transportation > Rural transit > Freig pipeline transportation > Airport planning and development > Airport maintenance > Bicycle and pedestrian > Ports and waterways >>> Transportation ope

440

Integrated head package for top mounted nuclear instrumentation  

DOE Patents (OSTI)

A nuclear reactor such as a pressurized water reactor has an integrated head package providing structural support and increasing shielding leading toward the vessel head. A reactor vessel head engages the reactor vessel, and a control rod guide mechanism over the vessel head raises and lowers control rods in certain of the thimble tubes, traversing penetrations in the reactor vessel head, and being coupled to the control rods. An instrumentation tube structure includes instrumentation tubes with sensors movable into certain thimble tubes disposed in the fuel assemblies. Couplings for the sensors also traverse penetrations in the reactor vessel head. A shroud is attached over the reactor vessel head and encloses the control rod guide mechanism and at least a portion of the instrumentation tubes when retracted. The shroud forms a structural element of sufficient strength to support the vessel head, the control rod guide mechanism and the instrumentation tube structure, and includes radiation shielding material for limiting passage of radiation from retracted instrumentation tubes. The shroud is thicker at the bottom adjacent the vessel head, where the more irradiated lower ends of retracted sensors reside. The vessel head, shroud and contents thus can be removed from the reactor as a unit and rested safely and securely on a support.

Malandra, Louis J. (McKeesport, PA); Hornak, Leonard P. (Forest Hills, PA); Meuschke, Robert E. (Monroeville, PA)

1993-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport reactor integrated" 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

Graduate Studies Transportation Systems Engineering  

E-Print Network (OSTI)

Graduate Studies Transportation Systems Engineering TRANSPORTATION SYSTEMS The transportation that transportation systems engineering can promote a thriving economy and a better quality of life by ensuring that transportation systems themselves affect the environment through operations, construction, and maintenance

Jacobs, Laurence J.

442

Introduction Transport in disordered graphene  

E-Print Network (OSTI)

Introduction Transport in disordered graphene Summary Ballistic transport in disordered graphene P, Gornyi, Mirlin Ballistic transport in disordered graphene #12;Introduction Transport in disordered graphene Summary Outline 1 Introduction Model Experimental motivation Transport in clean graphene 2

Fominov, Yakov

443

Hanford Single-Shell Tank Integrity Program  

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

Operations Contract Hanford Single Hanford Single- -Shell Shell Hanford Single Hanford Single Shell Shell Tank Integrity Tank Integrity Program Program Herbert S Berman Herbert S Berman Herbert S. Berman Herbert S. Berman July 29, 2009 July 29, 2009 1 Page 1 Tank Operations Contract Introduction * The Hanford site's principle historic mission was plutonium production for the manufacture of nuclear weapons. * Between 1944 and 1988, the site operated nine graphite- moderated light-water production reactors to irradiate moderated, light-water, production reactors to irradiate fuel and produce plutonium. * Four large chemical separations plants were run to extract plutonium from the fuel, and a variety of laboratories, support facilities, and related infrastructure to support production

444

MSIV leakage airborne iodine transport  

SciTech Connect

Gaseous iodine deposits on surfaces exposed to vapors. Basic chemical and physical principles predict this behavior, and several laboratory and in-plant measurements demonstrate the characteristic. An empirical model was developed that describes the deposition, resuspension, and transformation of airborne radioiodine molecular species as a stream containing these forms moves along its pathway. The model uses a data base of measured values of deposition and resuspension rates in its application and describes the conversion of the more reactive inorganic iodine species I[sub 2] to the less reactive organic species CH[sub 3]I as the iodine deposits and resuspends along the path. It also considers radioactive decay and chemical surface bonding during residence on surfaces. For the 8-day [sup 131]I, decay during the airborne portion of the transport is negligible. Verification of the model included measurement tests of long gaseous-activity sampling lines of different diameters, operated at different flow rates and stream temperatures. The model was applied to the streams at a boiling water reactor nuclear power plant to describe the transport through leaking main steam isolation valves (MSIVs), following a loss-of-coolant accident.

Cline, J.E. (Cline Associates Inc., Rockville, MD (United States))

1993-01-01T23:59:59.000Z

445

Cobalt transport at the Vermont Yankee BWR  

SciTech Connect

The behavior of Co-59 (the precursor of Co-60) is of particular interest in BWRs since Co-60 has been shown to be primarily responsible for BWR shutdown radiation levels. To assist in understanding cobalt transport in the BWR, a cobalt solution was injected at the effluent of the condensate demineralized system and was transported to the primary system via the feedwater at the Vermont Yankee BWR. The quantity injected was equivalent to the amount transported during routine operation over approximately 7 days. Sample system modified to eliminate cobalt bearing alloys were demonstrated to provide valid process stream samples for cobalt. Cobalt injected at the effluent of the condensate demineralizers was retained in portions of the feedwater cycle for approximately six hours prior to the initial indication of an increase in cobalt transport to the primary system. This anomaly, which led to considerable difficulty in interpreting concentration variations observed during and subsequent to the injection, is attributed to chemi-sorption of cobalt on iron oxide coated surfaces of the condensate/feedwater system. A similar phenomenon also occurred in the primary system thus eliminating the possibility of establishing an absolute value for the primary system cobalt source strength. Nonetheless, it was possible to infer that cobalt transported to the reactor by the feedwater was the primary source of cobalt available for deposition on the fuel. 14 figs., 13 tabs.

Palino, G.F.; Hobart, R.L.; Wall, P.S.; Sawochka, S.G.; Gardner, J.; Prystupa, M.

1986-05-01T23:59:59.000Z

446

(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

447

NREL: Transportation Research - Sustainable Transportation Basics  

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

Department of Energy's Alternative Fuels Data Center (AFDC) provide an introduction to sustainable transportation. NREL research supports development of electric, hybrid,...

448

Residential Transportation Historical Publications reports, data and  

U.S. Energy Information Administration (EIA) Indexed Site

Historical Publications Historical Publications Residential Transportation reports, data tables and transportation questionnaires Released: May 2008 The Energy Information Administration conducts several core consumption surveys. Among them was the Residential Transportation Energy Consumption Survey (RTECS). RTECS was designed by EIA to provide information on how energy is used by households for personal vehicles. It was an integral part of a series of surveys (i.e., core consumption surveys) designed by EIA to collect data on energy used by end-use economic sectors. The RTECS collected data on the number and type of vehicles used by the household. For each vehicle, data were collected on the number of miles traveled (commonly called VMT) for the year, the number of gallons of fuel consumed, the type of fuel used, the priced paid for fuel, and the number of miles per gallon. Additional electronic releases are available on the Transportation homepage.

449

Admission Guidelines 2015 MS-Ph.D Integrated Degree Program  

E-Print Network (OSTI)

Admission Guidelines 2015 MS-Ph.D Integrated Degree Program (For KAIST Master's Degree Enrolled. of Chemical & Biomolecular Engineering Dept. of Materials Science & Engineering Dept. of Nuclear Transportation Dept. of Electrical Engineering Dept. of Computer Science Dept. of Industrial

Kim, Min H.

450

Admission Guidelines 2014 MS-Ph.D Integrated Degree Program  

E-Print Network (OSTI)

Admission Guidelines 2014 MS-Ph.D Integrated Degree Program (For KAIST Master's Degree Enrolled. of Chemical & Biomolecular Engineering Dept. of Materials Science & Engineering Dept. of Nuclear Transportation Dept. of Electrical Engineering Dept. of Computer Science Dept. of Industrial

Kim, Min H.

451

Integrated 3D Acid Fracturing Model for Carbonate Reservoir Stimulation  

E-Print Network (OSTI)

in integrating fracture propagation, acid transport and dissolution, and well performance models in a seamless fashion for acid fracturing design. In this new approach, the fracture geometry data of a hydraulic fracture is first obtained from commercial models...

Wu, Xi

2014-06-23T23:59:59.000Z

452

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

453

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

454

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

455

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...

456

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

457

Transportation Baseline Schedule  

SciTech Connect

The “1999 National Transportation Program - Transportation Baseline Report” presents data that form a baseline to enable analysis and planning for future Department of Energy (DOE) Environmental Management (EM) waste/material transportation. The companion “1999 Transportation ‘Barriers’ Analysis” analyzes the data and identifies existing and potential problems that may prevent or delay transportation activities based on the data presented. The “1999 Transportation Baseline Schedule” (this report) uses the same data to provide an overview of the transportation activities of DOE EM waste/materials. This report can be used to identify areas where stakeholder interface is needed, and to communicate to stakeholders the quantity/schedule of shipments going through their area. Potential bottlenecks in the transportation system can be identified; the number of packages needed, and the capacity needed at receiving facilities can be planned. This report offers a visualization of baseline DOE EM transportation activities for the 11 major sites and the “Geologic Repository Disposal” site (GRD).

Fawcett, Ricky Lee; John, Mark Earl

2000-01-01T23:59:59.000Z

458

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

459

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

460

Calculation of the Local Neutronic Parameters for CANDU Fuel Bundles Using Transport Methods  

SciTech Connect

For a realistic neutronic evaluation of the CANDU reactor core it is important to accurately perform the local neutronic parameters (i.e. multigroup macroscopic cross sections for the core materials) calculation. This means using codes that allow a good geometric representation of the CANDU fuel bundle and then solving the transport equation. The paper reported here intends to study in detail the local behavior for two types of CANDU fuel, NU{sub 3}7 (Natural Uranium, 37 elements) and SEU{sub 4}3 (Slightly Enriched Uranium, 43 elements, with 1.1 wt% enrichment). The considered fuel types represent fresh and used bundles. The two types of CANDU super-cells are reference NU{sub 3}7, perturbed NU{sub 3}7, reference SEU{sub 4}3 and perturbed SEU{sub 4}3. The perturbed super-cells contain a Mechanical Control Absorber (a very strong reactivity device). For reaching the proposed objective a methodology is used based on WIMS and PIJXYZ codes. WIMS is a standard lattice-cell code, based on transport theory and it is used for producing fuel cell multigroup macroscopic cross sections. For obtaining the fine local neutronic parameters in the CANDU super-cells (k-eff values, local MCA reactivity worth, flux distributions and reaction rates), the PIJXYZ code is used. PIJXYZ is a 3D integral transport code using the first collision probability method and it has been developed for CANDU cell geometry. It is consistent with WIMS lattice-cell calculations and allows a good geometrical representation of the CANDU bundle in three dimensions. The analysis of the neutronic parameters consists of comparing the obtained results with the similar results calculated with the DRAGON code. This comparison shows a good agreement between these results. (authors)

Balaceanu, Victoria; Rizoiu, Andrei; Hristea, Viorel [Institute for Nuclear Research, PO Box 78, PITESTI (Romania)

2006-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport reactor integrated" 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

Isotope Program Transportation | Department of Energy  

Office of Environmental Management (EM)

Isotope Program Transportation Isotope Program Transportation Isotope Program Transportation More Documents & Publications Nuclear Fuel Storage and Transportation Planning Project...

462

Nuclear Transportation Management Services | Department of Energy  

Office of Environmental Management (EM)

Nuclear Transportation Management Services Nuclear Transportation Management Services Nuclear Transportation Management Services More Documents & Publications Transportation and...

463

Transportation Issues and Resolutions Compilation of Laboratory...  

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

Transportation Issues and Resolutions Compilation of Laboratory Transportation Work Package Reports Transportation Issues and Resolutions Compilation of Laboratory Transportation...

464

Transportation | ornl.gov  

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

Transportation Transportation Power Electronics and Electric Machinery Fuels, Engines, Emissions Transportation Analysis Vehicle Systems Energy Storage Propulsion Materials Lightweight Materials Bioenergy Fuel Cell Technologies Clean Energy Home | Science & Discovery | Clean Energy | Research Areas | Transportation SHARE Transportation Research ORNL researcher Jim Szybist uses a variable valve-train engine to evaluate different types of fuels, including ethanol blends, and their effects on the combustion process in an internal combustion engine. Oak Ridge National Laboratory brings together science and technology experts from across scientific disciplines to partner with government and industry in addressing transportation challenges. Research objectives are

465

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.

466

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

467

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

468

NREL: Transmission Grid Integration - Wind Integration Datasets  

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

Wind Integration Datasets The Wind Integration Datasets provide energy professionals with a consistent set of wind profiles for the eastern United States and the western United...

469

Midwestern Radioactive Materials Transportation Committee Agenda...  

Office of Environmental Management (EM)

Midwestern Radioactive Materials Transportation Committee Agenda Midwestern Radioactive Materials Transportation Committee Agenda Midwestern Radioactive Materials Transportation...

470

INL Site Executable Plan for Energy and Transportation Fuels Management  

SciTech Connect

It is the policy of the Department of Energy (DOE) that sustainable energy and transportation fuels management will be integrated into DOE operations to meet obligations under Executive Order (EO) 13423 "Strengthening Federal Environmental, Energy, and Transportation Management," the Instructions for Implementation of EO 13423, as well as Guidance Documents issued in accordance thereto and any modifcations or amendments that may be issued from time to time. In furtherance of this obligation, DOE established strategic performance-based energy and transportation fuels goals and strategies through the Transformational Energy Action Management (TEAM) Initiative, which were incorporated into DOE Order 430.2B "Departmental Energy, Renewable energy, and Transportation Management" and were also identified in DOE Order 450.1A, "Environmental Protection Program." These goals and accompanying strategies are to be implemented by DOE sites through the integration of energy and transportation fuels management into site Environmental Management Systems (EMS).

Ernest L. Fossum

2008-11-01T23:59:59.000Z

471

Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel  

SciTech Connect

The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

Not Available

1994-04-01T23:59:59.000Z

472

MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT  

SciTech Connect

The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and stora