Integral manifolding structure for fuel cell core having parallel gas flow

DOEpatents

Herceg, J.E.

1983-10-12

Disclosed herein are manifolding means for directing the fuel and oxidant gases to parallel flow passageways in a fuel cell core. Each core passageway is defined by electrolyte and interconnect walls. Each electrolyte and interconnect wall consists respectively of anode and cathode materials layered on the opposite sides of electrolyte material, or on the opposite sides of interconnect material. A core wall projects beyond the open ends of the defined core passageways and is disposed approximately midway between and parallel to the adjacent overlaying and underlying interconnect walls to define manifold chambers therebetween on opposite sides of the wall. Each electrolyte wall defining the flow passageways is shaped to blend into and be connected to this wall in order to redirect the corresponding fuel and oxidant passageways to the respective manifold chambers either above or below this intermediate wall. Inlet and outlet connections are made to these separate manifold chambers respectively, for carrying the fuel and oxidant gases to the core, and for carrying their reaction products away from the core.

Alameda-Contra Costa Transit District (AC Transit) Fuel Cell Transit Buses : Evaluation Results Update

DOT National Transportation Integrated Search

2007-10-01

In early 2007, the National Renewable Energy Laboratory (NREL) published a preliminary evaluation results report1 (April through November 2006) on hydrogen fuel cell and diesel buses operating at Alameda-Contra Costa Transit District (AC Transit) in ...

Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stauff, N.E.; Klim, T.K.; Taiwo, T.A.

2013-07-01

A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueledmore » cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to

National Fuel Cell Bus Program: Accelerated Testing Evaluation Report and Appendices, Alameda-Contra Costa Transit District (AC Transit)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chandler, K.; Eudy, L.

2009-01-01

This is an evaluation of hydrogen fuel cell transit buses operating at AC Transit in revenue service since March 20, 2006 compared to similar diesel buses operating from the same depot. This evaluation report includes results from November 2007 through October 2008. Evaluation results include implementation experience, fueling station operation, fuel cell bus operations at Golden Gate Transit, and evaluation results at AC Transit (bus usage, availability, fuel economy, maintenance costs, and roadcalls).

Alameda-Contra Costa Transit District (AC Transit) Fuel Cell Transit Buses : Third Evaluation Report

DOT National Transportation Integrated Search

2008-07-04

This report describes operations at Alameda-Contra Costa Transit District (AC Transit) for three prototype fuel cell buses and six diesel buses operating from the same location. This is the third evaluation report for this site, and it describes new ...

Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

NASA Astrophysics Data System (ADS)

Rouf; Su'ud, Zaki

2016-08-01

Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2015

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eudy, Leslie; Post, Matthew; Gikakis, Christina

This report, published annually, summarizes the progress of fuel cell electric bus (FCEB) development in the United States and discusses the achievements and challenges of introducing fuel cell propulsion in transit. Various stakeholders, including FCEB developers, transit agencies, and system integrators, have expressed the value of this annual status report, which provides a summary of results from evaluations performed by the National Renewable Energy Laboratory. The annual status report tracks the progress of the FCEB industry toward meeting technical targets, documents the lessons learned, and discusses the path forward for commercial viability of fuel cell technology for transit buses. Themore » 2015 summary results primarily focus on the most recent year for each demonstration, from August 2014 through July 2015. The results for these buses account for more than 1,045,000 miles traveled and 83,000 hours of fuel cell power system operation. The primary results presented in the report are from two demonstrations of fuel-cell-dominant bus designs: the Zero Emission Bay Area Demonstration Group led by Alameda-Contra Costa Transit District (AC Transit) in California and the American Fuel Cell Bus Project at SunLine Transit Agency in California.« less

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eudy, Leslie; Post, Matthew; Jeffers, Matthew

This report, published annually, summarizes the progress of fuel cell electric bus development in the United States and discusses the achievements and challenges of introducing fuel cell propulsion in transit. The report provides a summary of results from evaluations performed by the National Renewable Energy Laboratory. Funding for this effort is provided by the U.S. Department of Energy's Fuel Cell Technologies Office within the Office of Energy Efficiency and Renewable Energy and by the U.S. Department of Transportation's Federal Transit Administration. The 2016 summary results primarily focus on the most recent year for each demonstration, from August 2015 through Julymore » 2016. The results for these buses account for more than 550,000 miles traveled and 59,500 hours of fuel cell power system operation. The primary results presented in the report are from three demonstrations of two different fuel-cell-dominant bus designs: Zero Emission Bay Area Demonstration Group led by Alameda-Contra Costa Transit District (AC Transit) in California; American Fuel Cell Bus Project at SunLine Transit Agency in California; and American Fuel Cell Bus Project at the University of California at Irvine.« less

Laser cutting apparatus for nuclear core fuel subassembly

DOEpatents

Walch, Allan P.; Caruolo, Antonio B.

1982-02-23

The object of the invention is to provide a system and apparatus which employs laser cutting to disassemble a nuclear core fuel subassembly. The apparatus includes a gantry frame (C) which straddles the core fuel subassembly (14), an x-carriage (22) travelling longitudinally above the frame which carries a focus head assembly (D) having a vertically moving carriage (46) and a laterally moving carriage (52), a system of laser beam transferring and focusing mirrors carried by the x-carriage and focusing head assembly, and a shroud follower (F) and longitudinal follower (G) for following the shape of shroud (14) to maintain a beam focal point (44) fixed upon the shroud surface for accurate cutting.

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2010

DOT National Transportation Integrated Search

2010-11-11

This past year has been one of transition for the introduction of fuel cell transit buses. The existing generation of fuel cell buses from Van Hool and UTC Power has continued to operate in service at three transit agencies. At the same time, a new g...

Massachusetts Fuel Cell Bus Project: Demonstrating a Total Transit Solution for Fuel Cell Electric Buses in Boston

DOE Office of Scientific and Technical Information (OSTI.GOV)

The Federal Transit Administration's National Fuel Cell Bus Program focuses on developing commercially viable fuel cell bus technologies. Nuvera is leading the Massachusetts Fuel Cell Bus project to demonstrate a complete transit solution for fuel cell electric buses that includes one bus and an on-site hydrogen generation station for the Massachusetts Bay Transportation Authority (MBTA). A team consisting of ElDorado National, BAE Systems, and Ballard Power Systems built the fuel cell electric bus, and Nuvera is providing its PowerTap on-site hydrogen generator to provide fuel for the bus.

Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bromley, B.P.; Hyland, B.

2013-07-01

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (aboutmore » 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)« less

Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bromley, B.P.; Hyland, B.

2013-07-01

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu andmore » Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)« less

Demonstrating a Total Transit Solution for Fuel Cell Electric Buses in Boston

DOT National Transportation Integrated Search

2017-05-01

The Federal Transit Administrations (FTA) National Fuel Cell Bus Program (NFCBP) focuses on developing commercially viable fuel cell bus technologies. Nuvera is leading the Massachusetts Fuel Cell Bus project to demonstrate a complete transit solu...

Solid oxide fuel cell with monolithic core

DOEpatents

McPheeters, Charles C.; Mrazek, Franklin C.

1988-01-01

A solid oxide fuel cell in which fuel and oxidant gases undergo an electrochemical reaction to produce an electrical output includes a monolithic core comprised of a corrugated conductive sheet disposed between upper and lower generally flat sheets. The corrugated sheet includes a plurality of spaced, parallel, elongated slots which form a series of closed, linear, first upper and second lower gas flow channels with the upper and lower sheets within which a fuel gas and an oxidant gas respectively flow. Facing ends of the fuel cell are generally V-shaped and provide for fuel and oxidant gas inlet and outlet flow, respectively, and include inlet and outlet gas flow channels which are continuous with the aforementioned upper fuel gas and lower oxidant gas flow channels. The upper and lower flat sheets and the intermediate corrugated sheet are preferably comprised of ceramic materials and are securely coupled together such as by assembly in the green state and sintering together during firing at high temperatures. A potential difference across the fuel cell, or across a stacked array of similar fuel cells, is generated when an oxidant gas such as air and a fuel such as hydrogen gas is directed through the fuel cell at high temperatures, e.g., between 700.degree. C. and 1100.degree. C.

Nuclear core and fuel assemblies

DOEpatents

Downs, Robert E.

1981-01-01

A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

Alternative Fuels Data Center: Mass Transit

Science.gov Websites

traveled and fuel used by private vehicles. Vehicle fleet managers, corporate decision makers, and public effective incentives for fleet managers and corporate decision makers to build mass transit ridership

Solid oxide fuel cell having monolithic cross flow core and manifolding

DOEpatents

Poeppel, Roger B.; Dusek, Joseph T.

1984-01-01

This invention discloses a monolithic core construction having the flow passageways for the fuel and for the oxidant gases extended transverse to one another, whereby full face core manifolding can be achieved for these gases and their reaction products. The core construction provides that only anode material surround each fuel passageway and only cathode material surround each oxidant passageway, each anode and each cathode further sandwiching at spaced opposing sides electrolyte and interconnect materials to define electrolyte and interconnect walls. Webs of the cathode and anode material hold the electrolyte and interconnect walls spaced apart to define the flow passages. The composite anode and cathode wall structures are further alternately stacked on one another (with the separating electrolyte or interconnect material typically being a single common layer) whereby the fuel passageway and the oxidant passageways are disposed transverse to one another.

Solid oxide fuel cell having monolithic cross flow core and manifolding

DOEpatents

Poeppel, R.B.; Dusek, J.T.

1983-10-12

This invention discloses a monolithic core construction having the flow passageways for the fuel and for the oxidant gases extended transverse to one another, whereby full face core manifolding can be achieved for these gases and their reaction products. The core construction provides that only anode material surround each fuel passageway and only cathode material surround each oxidant passageway, each anode and each cathode further sandwiching at spaced opposing sides electrolyte and interconnect materials to define electrolyte and interconnect walls. Webs of the cathode and anode material hold the electrolyte and interconnect walls spaced apart to define the flow passages. The composite anode and cathode wall structures are further alternately stacked on one another (with the separating electrolyte or interconnect material typically being a single common layer) whereby the fuel passageways and the oxidant passageways are disposed transverse to one another.

National Fuel Cell Bus Program: Accelerated Testing Evaluation Report #2, Alameda-Contra Costa Transit District (AC Transit) and Appendices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eudy, L.; Chandler, K.

2010-06-01

This is an evaluation of hydrogen fuel cell transit buses operating at AC Transit in revenue service since March 20, 2006, comparing similar diesel buses operating from the same depot. It covers November 2007 through February 2010. Results include implementation experience, fueling station operation, evaluation results at AC Transit (bus usage, availability, fuel economy, maintenance costs, and road calls), and a summary of achievements and challenges encountered during the demonstration.

Solid oxide fuel cell with monolithic core

DOEpatents

McPheeters, C.C.; Mrazek, F.C.

1988-08-02

A solid oxide fuel cell in which fuel and oxidant gases undergo an electrochemical reaction to produce an electrical output includes a monolithic core comprised of a corrugated conductive sheet disposed between upper and lower generally flat sheets. The corrugated sheet includes a plurality of spaced, parallel, elongated slots which form a series of closed, linear, first upper and second lower gas flow channels with the upper and lower sheets within which a fuel gas and an oxidant gas respectively flow. Facing ends of the fuel cell are generally V-shaped and provide for fuel and oxidant gas inlet and outlet flow, respectively, and include inlet and outlet gas flow channels which are continuous with the aforementioned upper fuel gas and lower oxidant gas flow channels. The upper and lower flat sheets and the intermediate corrugated sheet are preferably comprised of ceramic materials and are securely coupled together such as by assembly in the green state and sintering together during firing at high temperatures. A potential difference across the fuel cell, or across a stacked array of similar fuel cells, is generated when an oxidant gas such as air and a fuel such as hydrogen gas is directed through the fuel cell at high temperatures, e.g., between 700 C and 1,100 C. 8 figs.

State Accountability in the Transition to Common Core. Updated

ERIC Educational Resources Information Center

Sears, Victoria

2014-01-01

The Common Core is at a critical juncture. While many surveys show that support for the standards themselves remains strong, implementation has not been without major challenges. "State Accountability in the Transition to Common Core," a new policy brief from the Thomas B. Fordham Institute, provides cautionary advice about what key…

Higher-order symmetry energy and neutron star core-crust transition with Gogny forces

NASA Astrophysics Data System (ADS)

Gonzalez-Boquera, C.; Centelles, M.; Viñas, X.; Rios, A.

2017-12-01

Background: An accurate determination of the core-crust transition is necessary in the modeling of neutron stars for astrophysical purposes. The transition is intimately related to the isospin dependence of the nuclear force at low baryon densities. Purpose: To study the symmetry energy and the core-crust transition in neutron stars using the finite-range Gogny nuclear interaction and to examine the deduced crustal thickness and crustal moment of inertia. Methods: The second-, fourth-, and sixth-order coefficients of the Taylor expansion of the energy per particle in powers of the isospin asymmetry are analyzed for Gogny forces. These coefficients provide information about the departure of the symmetry energy from the widely used parabolic law. The neutron star core-crust transition is evaluated by looking at the onset of thermodynamical instability of the liquid core. The calculation is performed with the exact Gogny equation of state (EoS) (i.e., the Gogny EoS with the full isospin dependence) for the β -equilibrated matter of the core, and also with the Taylor expansion of the Gogny EoS in order to assess the influence of isospin expansions on locating the inner edge of neutron star crusts. Results: The properties of the core-crust transition derived from the exact EoS differ from the predictions of the Taylor expansion even when the expansion is carried through sixth order in the isospin asymmetry. Gogny forces, using the exact EoS, predict the ranges 0.094 fm-3≲ρt≲0.118 fm-3 for the transition density and 0.339 MeVfm-3≲Pt≲0.665 MeVfm-3 for the transition pressure. The transition densities show an anticorrelation with the slope parameter L of the symmetry energy. The transition pressures are not found to correlate with L . Neutron stars obtained with Gogny forces have maximum masses below 1.74 M⊙ and relatively small moments of inertia. The crustal mass and moment of inertia are evaluated and comparisons are made with the constraints from observed

Effectively managing consumer fuel price driven transit demand.

DOT National Transportation Integrated Search

2013-05-01

This study presents a literature review of transit demand elasticities with respect to gas prices, describes features of a transit service area population that may be more sensitive to fuel prices, identifies where stress points in the family of tran...

Core-shell Au-Pd nanoparticles as cathode catalysts for microbial fuel cell applications

PubMed Central

Yang, Gaixiu; Chen, Dong; Lv, Pengmei; Kong, Xiaoying; Sun, Yongming; Wang, Zhongming; Yuan, Zhenhong; Liu, Hui; Yang, Jun

2016-01-01

Bimetallic nanoparticles with core-shell structures usually display enhanced catalytic properties due to the lattice strain created between the core and shell regions. In this study, we demonstrate the application of bimetallic Au-Pd nanoparticles with an Au core and a thin Pd shell as cathode catalysts in microbial fuel cells, which represent a promising technology for wastewater treatment, while directly generating electrical energy. In specific, in comparison with the hollow structured Pt nanoparticles, a benchmark for the electrocatalysis, the bimetallic core-shell Au-Pd nanoparticles are found to have superior activity and stability for oxygen reduction reaction in a neutral condition due to the strong electronic interaction and lattice strain effect between the Au core and the Pd shell domains. The maximum power density generated in a membraneless single-chamber microbial fuel cell running on wastewater with core-shell Au-Pd as cathode catalysts is ca. 16.0 W m−3 and remains stable over 150 days, clearly illustrating the potential of core-shell nanostructures in the applications of microbial fuel cells. PMID:27734945

Fuel Cell Demonstration Project at a Sunline Transit Agency

NASA Astrophysics Data System (ADS)

Hsiung, S.

2001-09-01

This is the final report summarizing the Fuel Cell Demonstration Project activities of the XCELLSIS Zebus (zero emissions bus) performance at the SunLine Transit Agency in Thousand Palms, California. Under this demonstration project, SunLine participated with XCELLSIS in the fueling, training, operating, and testing of this prototype fuel cell bus. The report presents a summary of project activities, including the results of the 13-month test of the XCELLSIS Zebus performance at SunLine Transit. This final report includes data relating to Zebus performance, along with the successes achieved beyond the technical realm. The study concludes that the project was very useful in establishing operating parameters and environmental testing in extreme heat conditions and in transferring technology to a transit agency. At the end of the 13-month test period, the Zebus ran flawlessly in the Michelin Challenge Bibendum from Los Angeles to Las Vegas, a 275-mile trek. SunLine refueled the Zebus in transit to Baker, California, 150 miles from its home base. Everyone who encountered or rode the Zebus was impressed with its smoothness, low engine noise, and absence of emissions. The study states that the future for the Zebus looks very bright. Fuel cell projects are anticipated to continue in California and Europe with the introduction new buses equipped with Ballard P5 and other fuel cell engines as early as the first half of 2003.

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2017

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eudy, Leslie; Post, Matthew B

This report, published annually, summarizes the progress of fuel cell electric bus (FCEB) development in the United States and discusses the achievements and challenges of introducing fuel cell propulsion in transit. The report provides a summary of results from evaluations performed by the National Renewable Energy Laboratory. This annual status report combines results from all FCEB demonstrations, tracks the progress of the FCEB industry toward meeting technical targets, documents the lessons learned, and discusses the path forward for commercial viability of fuel cell technology for transit buses. These data and analyses help provide needed information to guide future early-stage researchmore » and development. The 2017 summary results primarily focus on the most recent year for each demonstration, from August 2016 through July 2017. The primary results presented in the report are from five demonstrations of two different fuel-cell-dominant bus designs: Zero Emission Bay Area Demonstration Group led by Alameda-Contra Costa Transit District (AC Transit) in California; American Fuel Cell Bus (AFCB) Project at SunLine Transit Agency in California; AFCB Project at the University of California at Irvine; AFCB Project at Orange County Transportation Authority; and AFCB Project at Massachusetts Bay Transportation Authority.« less

Control of the Speed of a Light-Induced Spin Transition through Mesoscale Core-Shell Architecture.

PubMed

Felts, Ashley C; Slimani, Ahmed; Cain, John M; Andrus, Matthew J; Ahir, Akhil R; Abboud, Khalil A; Meisel, Mark W; Boukheddaden, Kamel; Talham, Daniel R

2018-05-02

The rate of the light-induced spin transition in a coordination polymer network solid dramatically increases when included as the core in mesoscale core-shell particles. A series of photomagnetic coordination polymer core-shell heterostructures, based on the light-switchable Rb a Co b [Fe(CN) 6 ] c · mH 2 O (RbCoFe-PBA) as core with the isostructural K j Ni k [Cr(CN) 6 ] l · nH 2 O (KNiCr-PBA) as shell, are studied using temperature-dependent powder X-ray diffraction and SQUID magnetometry. The core RbCoFe-PBA exhibits a charge transfer-induced spin transition (CTIST), which can be thermally and optically induced. When coupled to the shell, the rate of the optically induced transition from low spin to high spin increases. Isothermal relaxation from the optically induced high spin state of the core back to the low spin state and activation energies associated with the transition between these states were measured. The presence of a shell decreases the activation energy, which is associated with the elastic properties of the core. Numerical simulations using an electro-elastic model for the spin transition in core-shell particles supports the findings, demonstrating how coupling of the core to the shell changes the elastic properties of the system. The ability to tune the rate of optically induced magnetic and structural phase transitions through control of mesoscale architecture presents a new approach to the development of photoswitchable materials with tailored properties.

Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period in Japan

NASA Astrophysics Data System (ADS)

Fujimura, Koji; Sasahira, Akira; Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi

According to the Japan's Framework for Nuclear Energy Policy(1), a basic scenario for fast breeder reactors (FBRs) is that they will be introduced on a commercial basis starting around 2050 replacing light water reactors (LWRs). During the FBR introduction period, the Pu from LWR spent fuel is used for FBR startup. Howerver, the FBR core loaded with this Pu has a larger burnup reactivity due to its larger isotopic content of Pu-241 than a core loaded with Pu from an FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of an FBR. We investigated, an FBR transitional core concept to confront the issues of the FBR introductory period in Japan. Core specifications are based on the compact-type sodium-cooled mixed oxide (MOX)-fueled core designed from the Japanese FBR cycle feasibility studies, because lower Pu inventory should be better for the FBR introductory period in view of its flexibility for the required reprocessing amount of LWR spent fuel to start up FBRs. The reference specifications were selected as follows. Output of 1500MWe and average discharge fuel burnup of about 150GWd/t. Minor Actinides (MAs) recovered from LWR spent fuels which provide Pu to startup FBRs are loaded to the initial loading fuels and exchanged fuels during few cycles until equilibrium. We made the MA content of the initial loading fuel four kinds like 0%, 3%, 4%, 5%. The average of the initial loading fuel is assumed to be 3%, and that of the exchange fuel is set as 5%. This 5% maximum of the MA content is based on the irradiation results of the experimental fast reactor Joyo. We evaluated the core performances including burnup characteristics and the reactivity coefficient and confirmed that transitional core from initial loading until equilibrium cycle with loaded Pu from LWR spent fuel performs similary to an FBR multi-recycling core.

BC Transit Fuel Cell Bus Project : Evaluation Results Report

DOT National Transportation Integrated Search

2014-02-02

British Columbia Transit (BC Transit) has been leading a demonstration of fuel cell electric buses (FCEB) in Whistler, Canada, since early 2010. This 20-bus demonstration was introduced during the 2010 Winter Olympic Games and is the worlds larges...

Clean Air Program : Design Guidelines for Bus Transit Systems Using Alcohol Fuel (Methanol and Ethanol) as an Alternative Fuel

DOT National Transportation Integrated Search

1996-08-01

Although there are over one thousand transit buses in revenue service in the U.S. that are powered by alternative fuels, there are no comprehensive guidelines for the safe design and operation of alternative fuel facilities and vehicles for transit s...

Leading Change: Transitioning to the Common Core

ERIC Educational Resources Information Center

Brandt, Brian

2012-01-01

For many school districts in the United States, making the transition to the Common Core State Standards (CCSS) is proving to be a challenge at a time when they are already facing many obstacles, not the least of which is dwindling financial resources. A change of this magnitude cannot be made haphazardly; to be successful in its goal of helping…

Using Checklists to Assess Your Transition to Alternative Fuels: A Technical Reference

DOE Office of Scientific and Technical Information (OSTI.GOV)

Risch, C. E.; Santini, D. J.; Johnson, L. R.

The Checklist for Transition to New Alternative Fuel(s) was published in September 2011 by Chuck Risch and Dan Santini. Many improvements, described below, have been incorporated into this current document, Checklists for Assessing the Transitions to New Highway Fuels.2 Further, the original authors and Larry Johnson, co-author of the current report, identified a need for a succinct version of the full report and prepared a brochure based on it to aid busy decisionmakers: Check It Out: Using Checklists to Assess Your Transition to Alternative Fuels.2 These checklists are tools for those stakeholders charged with determining a feasible alternative fuel ormore » fuels for highway transportation systems of the future. The original had four major players whose needs had to be satisfied for a successful transition. The term “activist,” intended to encompass environmental and other special interests, was included in the “customers” category. Activists are customers of the government in the sense that they organize citizens to exert political pressure to regulate the design of vehicles, fuel infrastructure, and roadway networks. Many who evaluate alternative fuels view activists, particularly environmental activists, as a separate category. Further, “activist” has become a pejorative term to many people. Therefore, we have used the word “advocate” or “activist/advocate” instead. Thus, in this update we recognize that environmental and other activists/advocates have been--and will continue to be--a powerful force promoting change in the nature of the fuels that are used in transportation.« less

Getting to the Core of Transition: A Re-Assessment of Old Wine in New Bottles.

ERIC Educational Resources Information Center

Stodden, Robert A.; Leake, David W.

1994-01-01

The core of the education system must be transformed to make it supportive of exceptional students in transition. Attempts to improve the system by adding on programs rather than instituting core changes have not been effective. A "teaming" process that promotes school-community and interagency collaboration and infuses transition into the core of…

Transition from HEU to LEU fuel in Romania`s 14-MW TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M.M.; Snelgrove, J.L.

1991-12-31

The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each ofmore » the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.« less

Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

DOEpatents

Peterson, Per F.

2013-05-14

A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2013

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eudy, Leslie; Gikakis, Christina

2013-12-01

This report is the seventh in an annual series of reports that summarize the progress of fuel cell electric bus (FCEB) development in the United States and discuss the achievements and challenges of introducing fuel cell propulsion in transit. The report also provides a snapshot of current FCEB performance results from August 2012 through July 2013 for five FCEB demonstrations at four transit agencies.

Performance of the MTR core with MOX fuel using the MCNP4C2 code.

PubMed

Shaaban, Ismail; Albarhoum, Mohamad

2016-08-01

The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. Copyright © 2016 Elsevier Ltd. All rights reserved.

BC Transit Fuel Cell Bus Project Evaluation Results : Second Report

DOT National Transportation Integrated Search

2014-09-01

Beginning in 2009, British Columbia Transit (BC Transit) led a project to conduct a 5-year demonstration of 20 fuel cell electric buses (FCEB) in Whistler, Canada. The FCEB fleet was introduced during the 2010 Winter Olympic Games and operated throug...

An evaluation of alternative fuels usage by public transit agencies.

DOT National Transportation Integrated Search

2009-12-01

The oil crisis of the 1970s forced Americans to reconsider using fossil fuels as a primary energy source. In : the public transit arena, private transit companies found themselves unable to compete in the urban : environment as rapidly rising oil pri...

National Fuel Cell Bus Program : Accelerated Testing Report, AC Transit

DOT National Transportation Integrated Search

2009-01-01

This is an evaluation of hydrogen fuel cell transit buses operating at AC Transit in revenue service since March 20, 2006 compared to similar diesel buses operating from the same depot. This evaluation report includes results from November 2007 throu...

HTGR fuels and core development program. Quarterly progress report for the period ending August 31, 1975

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1975-09-30

Studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies are described. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and, where appropriate, the data are presented in tables, graphs, and photographs. (auth)

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh

2010-09-01

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 physicsmore » 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

Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

NASA Astrophysics Data System (ADS)

Permana, Sidik; Saputra, Geby; Suzuki, Mitsutoshi; Saito, Masaki

2017-01-01

Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

Alternative Fuels Data Center: Greater Portland Transit District Looks

Science.gov Websites

Forward with Natural GasA> Greater Portland Transit District Looks Forward with Natural Gas to , Maine, powers its transit vehicles with compressed natural gas. For information about this project Related Videos Photo of a car Hydrogen Powers Fuel Cell Vehicles in California Nov. 18, 2017 Photo of a

3D Field Modifications of Core Neutral Fueling In the EMC3-EIRENE Code

NASA Astrophysics Data System (ADS)

Waters, Ian; Frerichs, Heinke; Schmitz, Oliver; Ahn, Joon-Wook; Canal, Gustavo; Evans, Todd; Feng, Yuehe; Kaye, Stanley; Maingi, Rajesh; Soukhanovskii, Vsevolod

2017-10-01

The application of 3-D magnetic field perturbations to the edge plasmas of tokamaks has long been seen as a viable way to control damaging Edge Localized Modes (ELMs). These 3-D fields have also been correlated with a density drop in the core plasmas of tokamaks; known as `pump-out'. While pump-out is typically explained as the result of enhanced outward transport, degraded fueling of the core may also play a role. By altering the temperature and density of the plasma edge, 3-D fields will impact the distribution function of high energy neutral particles produced through ion-neutral energy exchange processes. Starved of the deeply penetrating neutral source, the core density will decrease. Numerical studies carried out with the EMC3-EIRENE code on National Spherical Tokamak eXperiment-Upgrade (NSTX-U) equilibria show that this change to core fueling by high energy neutrals may be a significant contributor to the overall particle balance in the NSTX-U tokamak: deep core (Ψ < 0.5) fueling from neutral ionization sources is decreased by 40-60% with RMPs. This work was funded by the US Department of Energy under Grant DE-SC0012315.

Fuel cell performance of palladium-platinum core-shell electrocatalysts synthesized in gram-scale batches

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khateeb, Siddique; Su, Dong; Guerreo, Sandra

This article presents the performance of palladium-platinum core-shell catalysts (Pt/Pd/C) for oxygen reduction synthesized in gram-scale batches in both liquid cells and polymer-electrolyte membrane fuel cells. Core-shell catalyst synthesis and characterization, ink fabrication, and cell assembly details are discussed. The Pt mass activity of the Pt/Pd core-shell catalyst was 0.95 A mg –1 at 0.9 V measured in liquid cells (0.1 M HClO4), which was 4.8 times higher than a commercial Pt/C catalyst. The performances of Pt/Pd/C and Pt/C in large single cells (315 cm 2) were assessed under various operating conditions. The core-shell catalyst showed consistently higher performance thanmore » commercial Pt/C in fuel cell testing. A 20–60 mV improvement across the whole current density range was observed on air. Sensitivities to temperature, humidity, and gas composition were also investigated and the core-shell catalyst showed a consistent benefit over Pt under all conditions. However, the 4.8 times activity enhancement predicated by liquid cell measurements was not fully realized in fuel cells.« less

Fuel cell performance of palladium-platinum core-shell electrocatalysts synthesized in gram-scale batches

DOE PAGES

Khateeb, Siddique; Su, Dong; Guerreo, Sandra; ...

2016-05-03

This article presents the performance of palladium-platinum core-shell catalysts (Pt/Pd/C) for oxygen reduction synthesized in gram-scale batches in both liquid cells and polymer-electrolyte membrane fuel cells. Core-shell catalyst synthesis and characterization, ink fabrication, and cell assembly details are discussed. The Pt mass activity of the Pt/Pd core-shell catalyst was 0.95 A mg –1 at 0.9 V measured in liquid cells (0.1 M HClO4), which was 4.8 times higher than a commercial Pt/C catalyst. The performances of Pt/Pd/C and Pt/C in large single cells (315 cm 2) were assessed under various operating conditions. The core-shell catalyst showed consistently higher performance thanmore » commercial Pt/C in fuel cell testing. A 20–60 mV improvement across the whole current density range was observed on air. Sensitivities to temperature, humidity, and gas composition were also investigated and the core-shell catalyst showed a consistent benefit over Pt under all conditions. However, the 4.8 times activity enhancement predicated by liquid cell measurements was not fully realized in fuel cells.« less

Santa Clara Valley Transportation Authority and San Mateo County Transit District Fuel Cell Transit Buses : Evaluation Results

DOT National Transportation Integrated Search

2006-11-23

This report provides evaluation results for prototype fuel cell transit buses operating at Santa Clara Valley Transportation Authority (VTA) in San Jose, California. San Mateo County Transit District (SamTrans) in San Carlos, California, is a partner...

Fuel Cell Buses in U.S. Transit Fleets : Current Status 2014

DOT National Transportation Integrated Search

2014-12-03

This report, published annually, summarizes the progress of fuel cell electric bus (FCEB) development in the United States and discusses the achievements and challenges of introducing fuel cell propulsion in transit. Various stakeholders, including d...

Fuel Cell Transit Buses : ThunderPower Bus Evaluation at SunLine Transit Agency

DOT National Transportation Integrated Search

2003-11-01

This report provides an overview of the ThunderPower fuel cell bus demonstration at SunLine Transit Agency in Thousand Palms, California. Under contract with the U.S. Department of Energy (DOE), the National Renewable Energy Laboratory (NREL) evaluat...

Fuel Cell Buses in U.S. Transit Fleets : Current Status 2015

DOT National Transportation Integrated Search

2015-12-01

This report, published annually, summarizes the progress of fuel cell electric bus (FCEB) development in the United States and discusses the achievements and challenges of introducing fuel cell propulsion in transit. The report provides a summary of ...

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2017

DOT National Transportation Integrated Search

2017-11-01

This report, published annually, summarizes the progress of fuel cell electric bus (FCEB) development in the United States and discusses the achievements and challenges of introducing fuel cell propulsion in transit. The report provides a summary of ...

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2016

DOT National Transportation Integrated Search

2016-11-01

This report, published annually, summarizes the progress of fuel cell electric bus (FCEB) development in the United States and discusses the achievements and challenges of introducing fuel cell propulsion in transit. The report provides a summary of ...

Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

DOE Office of Scientific and Technical Information (OSTI.GOV)

Permana, Sidik; Novitrian,; Waris, Abdul

Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by conversion ratio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissilemore » material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loading scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.« less

Fuel Cell Buses in U.S. Transit Fleets : Current Status 2012

DOT National Transportation Integrated Search

2012-11-12

This report is the sixth in an annual series of reports that summarize the progress of fuel cell electric bus (FCEB) development in the United States and discuss the achievements and challenges of introducing fuel cell propulsion in transit. The repo...

Fuel Cell Buses in U.S. Transit Fleets : Current Status 2013

DOT National Transportation Integrated Search

2013-12-01

This report is the seventh in an annual series of reports that summarize the progress of fuel cell electric bus (FCEB) development in the United States and discuss the achievements and challenges of introducing fuel cell propulsion in transit. This r...

Analysis of fuel cycle strategies and U.S. transition scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wigeland, Roald; Taiwo, Temitope A.

2016-10-17

The nuclear fuel cycle Evaluation and Screening (E&S) study that was completed in October 2014 [1] enabled the identification of four fuel cycle groups that are considered most promising based on a set of nine evaluation criteria: (a) six benefit criteria of Nuclear Waste Management, Proliferation Risk, Nuclear Material Security Risk, Safety, Environmental Impact, Resource Utilization, and (b) three challenge criteria of Development and Deployment Risk, Institutional Issues, Financial Risk and Economics. The E&S study was conducted at a level of analysis that is "technology- neutral," that is, without consideration of specific technologies, but using the fundamental physics characteristics ofmore » each part of the fuel cycle. The study focused on the fuel cycle performance benefits at the fuel cycle equilibrium state, with only limited consideration of transition and deployment impacts. Common characteristics of the four most promising fuel cycle options include continuous recycle of all U/Pu or U/TRU, the use of fast-spectrum reactors, and no use of uranium enrichment once fuel cycle equilibrium has been established. The high-level wastes are mainly from processing of irradiated fuel, and there would be no disposal of any spent fuel. Building on the findings of the E&S study, additional studies have been conducted in the last two years following the information exchange meeting, the 13th IEMPT, which was held in Seoul, the Republic of Korea in 2014. Insights are presented from the recent studies on the benefits and challenges of recycling minor actinides, and transition considerations to some of the most promising fuel cycle options.« less

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2011

DOT National Transportation Integrated Search

2011-11-11

his report is the fifth in a series of annual status reports that summarize the progress resulting from fuel cell transit bus demonstrations in the United States and provide a discussion of the achievements and challenges of fuel cell propulsion in t...

Santa Clara Valley Transportation Authority and San Mateo County Transit District Fuel Cell Transit Buses : Preliminary Evaluation Results

DOT National Transportation Integrated Search

2006-03-23

This report provides preliminary results from an evaluation of prototype fuel cell transit buses operating at Santa Clara Valley Transportation Authority (VTA) in San Jose, California. San Mateo County Transit District (SamTrans) in San Carlos, Calif...

Impact of Americium-241 (n,γ) Branching Ratio on SFR Core Reactivity and Spent Fuel Characteristics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hiruta, Hikaru; Youinou, Gilles J.; Dixon, Brent W.

An accurate prediction of core physics and fuel cycle parameters largely depends on the order of details and accuracy in nuclear data taken into account for actual calculations. 241Am is a major gateway nuclide for most of minor actinides and thus important nuclide for core physics and fuel-cycle calculations. The 241Am(n,?) branching ratio (BR) is in fact the energy dependent (see Fig. 1), therefore, it is necessary to taken into account the spectrum effect on the calculation of the average BR for the full-core depletion calculations. Moreover, the accuracy of the BR used in the depletion calculations could significantly influencemore » the core physics performance and post irradiated fuel compositions. The BR of 241Am(n,?) in ENDF/B-VII.0 library is relatively small and flat in thermal energy range, gradually increases within the intermediate energy range, and even becomes larger at the fast energy range. This indicates that the properly collapsed BR for fast reactors could be significantly different from that of thermal reactors. The evaluated BRs are also differ from one evaluation to another. As seen in Table I, average BRs for several evaluated libraries calculated by means of a fast spectrum are similar but have some differences. Most of currently available depletion codes use a pre-determined single value BR for each library. However, ideally it should be determined on-the-fly basis like that of one-group cross sections. These issues provide a strong incentive to investigate the effect of different 241Am(n,?) BRs on core and spent fuel parameters. This paper investigates the impact of the 241Am(n,?) BR on the results of SFR full-core based fuel-cycle calculations. The analysis is performed by gradually increasing the value of BR from 0.15 to 0.25 and studying its impact on the core reactivity and characteristics of SFR spent fuels over extended storage times (~10,000 years).« less

Method of fabricating a monolithic core for a solid oxide fuel cell

DOEpatents

Zwick, Stanley A.; Ackerman, John P.

1985-01-01

A method is disclosed for forming a core for use in a solid oxide fuel cell that electrochemically combines fuel and oxidant for generating galvanic output. The core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support consisting instead only of the active anode, cathode, electrolyte and interconnect materials. Each electrolyte wall consists of cathode and anode materials sandwiching electrolyte material therebetween, and each interconnect wall consists of the cathode and anode materials sandwiching interconnect material therebetween. The electrolyte and interconnect walls define a plurality of substantially parallel core passageways alternately having respectively the inside faces thereof with only the anode material or with only the cathode material exposed. In the wall structure, the electrolyte and interconnect materials are only 0.002-0.01 cm thick; and the cathode and anode materials are only 0.002-0.05 cm thick. The method consists of building up the electrolyte and interconnect walls by depositing each material on individually and endwise of the wall itself, where each material deposit is sequentially applied for one cycle; and where the depositing cycle is repeated many times until the material buildup is sufficient to formulate the core. The core is heat cured to become dimensionally and structurally stable.

Density Affects the Nature of the Hexatic-Liquid Transition in Two-Dimensional Melting of Soft-Core Systems

NASA Astrophysics Data System (ADS)

Zu, Mengjie; Liu, Jun; Tong, Hua; Xu, Ning

2016-08-01

We find that both continuous and discontinuous hexatic-liquid transitions can happen in the melting of two-dimensional solids of soft-core disks. For three typical model systems, Hertzian, harmonic, and Gaussian-core models, we observe the same scenarios. These systems exhibit reentrant crystallization (melting) with a maximum melting temperature Tm happening at a crossover density ρm. The hexatic-liquid transition at a density smaller than ρm is discontinuous. Liquid and hexatic phases coexist in a density interval, which becomes narrower with increasing temperature and tends to vanish approximately at Tm. Above ρm, the transition is continuous, in agreement with the Kosterlitz-Thouless-Halperin-Nelson-Young theory. For these soft-core systems, the nature of the hexatic-liquid transition depends on density (pressure), with the melting at ρm being a plausible transition point from discontinuous to continuous hexatic-liquid transition.

Core-shell rhodium sulfide catalyst for hydrogen evolution reaction / hydrogen oxidation reaction in hydrogen-bromine reversible fuel cell

NASA Astrophysics Data System (ADS)

Li, Yuanchao; Nguyen, Trung Van

2018-04-01

Synthesis and characterization of high electrochemical active surface area (ECSA) core-shell RhxSy catalysts for hydrogen evolution oxidation (HER)/hydrogen oxidation reaction (HOR) in H2-Br2 fuel cell are discussed. Catalysts with RhxSy as shell and different percentages (5%, 10%, and 20%) of platinum on carbon as core materials are synthesized. Cyclic voltammetry is used to evaluate the Pt-equivalent mass specific ECSA and durability of these catalysts. Transmission electron microscopy (TEM), X-ray Photoelectron spectroscopy (XPS) and Energy-dispersive X-ray spectroscopy (EDX) techniques are utilized to characterize the bulk and surface compositions and to confirm the core-shell structure of the catalysts, respectively. Cycling test and polarization curve measurements in the H2-Br2 fuel cell are used to assess the catalyst stability and performance in a fuel cell. The results show that the catalysts with core-shell structure have higher mass specific ECSA (50 m2 gm-Rh-1) compared to a commercial catalyst (RhxSy/C catalyst from BASF, 6.9 m2 gm-Rh-1). It also shows better HOR/HER performance in the fuel cell. Compared to the platinum catalyst, the core-shell catalysts show more stable performance in the fuel cell cycling test.

Analysis, operation and maintenance of a fuel cell/battery series-hybrid bus for urban transit applications

NASA Astrophysics Data System (ADS)

Bubna, Piyush; Brunner, Doug; Gangloff, John J.; Advani, Suresh G.; Prasad, Ajay K.

The fuel cell hybrid bus (FCHB) program was initiated at the University of Delaware in 2005 to demonstrate the viability of fuel cell vehicles for transit applications and to conduct research and development to facilitate the path towards their eventual commercialization. Unlike other fuel cell bus programs, the University of Delaware's FCHB design features a battery-heavy hybrid which offers multiple advantages in terms of cost, performance and durability. The current fuel cell hybrid bus is driven on a regular transit route at the University of Delaware. The paper describes the baseline specifications of the bus with a focus on the fuel cell and the balance of plant. The fuel cell/battery series-hybrid design is well suited for urban transit routes and provides key operational advantages such as hydrogen fuel economy, efficient use of the fuel cell for battery recharging, and regenerative braking. The bus is equipped with a variety of sensors including a custom-designed cell voltage monitoring system which provide a good understanding of bus performance under normal operation. Real-time data collection and analysis have yielded key insights for fuel cell bus design optimization. Results presented here illustrate the complex flow of energy within the various subsystems of the fuel cell hybrid bus. A description of maintenance events has been included to highlight the issues that arise during general operation. The paper also describes several modifications that will facilitate design improvements in future versions of the bus. Overall, the fuel cell hybrid bus demonstrates the viability of fuel cells for urban transit applications in real world conditions.

20 CFR 641.730 - How will the Department assist grantees in the transition to the new core performance indicators?

Code of Federal Regulations, 2011 CFR

2011-04-01

... the transition to the new core performance indicators? 641.730 Section 641.730 Employees' Benefits... transition to the new core performance indicators? (a) General transition provision. As soon as practicable...-need indicator so that the grantees and the Department may collect sufficient data to set a meaningful...

20 CFR 641.730 - How will the Department assist grantees in the transition to the new core performance indicators?

Code of Federal Regulations, 2010 CFR

2010-04-01

... 20 Employees' Benefits 3 2010-04-01 2010-04-01 false How will the Department assist grantees in the transition to the new core performance indicators? 641.730 Section 641.730 Employees' Benefits... transition to the new core performance indicators? (a) General transition provision. As soon as practicable...

Measuring air core characteristics of a pressure-swirl atomizer via a transparent acrylic nozzle at various Reynolds numbers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Eun J.; Oh, Sang Youp; Kim, Ho Y.

2010-11-15

Because of thermal fluid-property dependence, atomization stability (or flow regime) can change even at fixed operating conditions when subject to temperature change. Particularly at low temperatures, fuel's high viscosity can prevent a pressure-swirl (or simplex) atomizer from sustaining a centrifugal-driven air core within the fuel injector. During disruption of the air core inside an injector, spray characteristics outside the nozzle reflect a highly unstable, nonlinear mode where air core length, Sauter mean diameter (SMD), cone angle, and discharge coefficient variability. To better understand injector performance, these characteristics of the pressure-swirl atomizer were experimentally investigated and data were correlated to Reynoldsmore » numbers (Re). Using a transparent acrylic nozzle, the air core length, SMD, cone angle, and discharge coefficient are observed as a function of Re. The critical Reynolds numbers that distinguish the transition from unstable mode to transitional mode and eventually to a stable mode are reported. The working fluids are diesel and a kerosene-based fuel, referred to as bunker-A. (author)« less

SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: Fourth Results Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eudy, L.; Chandler, K.

2013-01-01

SunLine Transit Agency, which provides public transit services to the Coachella Valley area of California, has demonstrated hydrogen and fuel cell bus technologies for more than 10 years. In May 2010, SunLine began demonstrating the advanced technology (AT) fuel cell bus with a hybrid electric propulsion system, fuel cell power system, and lithium-based hybrid batteries. This report describes operations at SunLine for the AT fuel cell bus and five compressed natural gas buses. The U.S. Department of Energy's National Renewable Energy Laboratory (NREL) is working with SunLine to evaluate the bus in real-world service to document the results and helpmore » determine the progress toward technology readiness. NREL has previously published three reports documenting the operation of the fuel cell bus in service. This report provides a summary of the results with a focus on the bus operation from February 2012 through November 2012.« less

Direct measurement of the transition from edge to core power coupling in a light-ion helicon source

NASA Astrophysics Data System (ADS)

Piotrowicz, P. A.; Caneses, J. F.; Showers, M. A.; Green, D. L.; Goulding, R. H.; Caughman, J. B. O.; Biewer, T. M.; Rapp, J.; Ruzic, D. N.

2018-05-01

We present time-resolved measurements of an edge-to-core power transition in a light-ion (deuterium) helicon discharge in the form of infra-red camera imaging of a thin stainless steel target plate on the Proto-Material Exposure eXperiment device. The time-resolved images measure the two-dimensional distribution of power deposition in the helicon discharge. The discharge displays a mode transition characterized by a significant increase in the on-axis electron density and core power coupling, suppression of edge power coupling, and the formation of a fast-wave radial eigenmode. Although the self-consistent mechanism that drives this transition is not yet understood, the edge-to-core power transition displays characteristics that are consistent with the discharge entering a slow-wave anti-resonant regime. RF magnetic field measurements made across the plasma column, together with the power deposition results, provide direct evidence to support the suppression of the slow-wave in favor of core plasma production by the fast-wave in a light-ion helicon source.

Direct measurement of the transition from edge to core power coupling in a light-ion helicon source

DOE PAGES

Piotrowicz, Pawel A.; Caneses, Juan F.; Showers, Melissa A.; ...

2018-05-02

Here, we present time-resolved measurements of an edge-to-core power transition in a light-ion (deuterium) helicon discharge in the form of infra-red camera imaging of a thin stainless steel target plate on the Proto-Material Exposure eXperiment device. The time-resolved images measure the two-dimensional distribution of power deposition in the helicon discharge. The discharge displays a mode transition characterized by a significant increase in the on-axis electron density and core power coupling, suppression of edge power coupling, and the formation of a fast-wave radial eigenmode. Although the self-consistent mechanism that drives this transition is not yet understood, the edge-to-core power transition displaysmore » characteristics that are consistent with the discharge entering a slow-wave anti-resonant regime. RF magnetic field measurements made across the plasma column, together with the power deposition results, provide direct evidence to support the suppression of the slow-wave in favor of core plasma production by the fast-wave in a light-ion helicon source.« less

Direct measurement of the transition from edge to core power coupling in a light-ion helicon source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piotrowicz, Pawel A.; Caneses, Juan F.; Showers, Melissa A.

Here, we present time-resolved measurements of an edge-to-core power transition in a light-ion (deuterium) helicon discharge in the form of infra-red camera imaging of a thin stainless steel target plate on the Proto-Material Exposure eXperiment device. The time-resolved images measure the two-dimensional distribution of power deposition in the helicon discharge. The discharge displays a mode transition characterized by a significant increase in the on-axis electron density and core power coupling, suppression of edge power coupling, and the formation of a fast-wave radial eigenmode. Although the self-consistent mechanism that drives this transition is not yet understood, the edge-to-core power transition displaysmore » characteristics that are consistent with the discharge entering a slow-wave anti-resonant regime. RF magnetic field measurements made across the plasma column, together with the power deposition results, provide direct evidence to support the suppression of the slow-wave in favor of core plasma production by the fast-wave in a light-ion helicon source.« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE PAGES

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig; ...

2017-07-10

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Eemian and penultimate transition reflected in the chemical ice core record from Dome C

NASA Astrophysics Data System (ADS)

Bigler, M.; Lambert, F.; Stauffer, B.; Röthlisberger, R.; Wolff, E. W.

2003-04-01

Within the scope of the European Project for Ice Coring in Antarctica (EPICA) chemical analyses have been done along the Dome C ice core. Among other substances, Ca2+, dust, Na+, NH_4{}+, NO_3{}- and electrolytical melt water conductivity have been measured at 1 cm resolution with the Bern Continuous Flow Analysis (CFA) system. Here we present new data from the Eemian and the preceding transition covering an age interval from approximately 180 kyr to 110 kyr before present. This sequence is compared with the Holocene and the last transition, mainly with emphasis on terrestrial and marine tracers. Concentration levels for the two periods compare quite well, but the general shape differs considerably. The changes in dust input to Dome C seemed to have been much more abrupt during the penultimate transition than during the last transition (18 to 15 kyr BP). This may reflect different conditions and/or processes in the dust source region.

The radiative decays of excited states of transition elements located inside and near core-shell nanoparticles

NASA Astrophysics Data System (ADS)

Pukhov, Konstantin K.

2017-12-01

Here we discuss the radiative decays of excited states of transition elements located inside and outside of the subwavelength core-shell nanoparticles embedded in dielectric medium. Based on the quantum mechanics and quantum electrodynamics, the general analytical expressions are derived for the probability of the spontaneous transitions in the luminescent centers (emitter) inside and outside the subwavelength core-shell nanoparticle. Obtained expressions holds for arbitrary orientation of the dipole moment and the principal axes of the quadrupole moment of the emitter with respect to the radius-vector r connecting the center of the emitter with the center of the nanoparticle. They have simple form and show how the spontaneous emission in core-shell NPs can be controlled and engineered due to the dependence of the emission rates on core-shell sizes, radius-vector r and permittivities of the surrounding medium, shell, and core.

10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive of...

10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive of...

10 CFR 73.35 - Requirements for physical protection of irradiated reactor fuel (100 grams or less) in transit.

Code of Federal Regulations, 2014 CFR

2014-01-01

... fuel (100 grams or less) in transit. 73.35 Section 73.35 Energy NUCLEAR REGULATORY COMMISSION... Transit § 73.35 Requirements for physical protection of irradiated reactor fuel (100 grams or less) in... quantity of irradiated reactor fuel weighing 100 grams (0.22 pounds) or less in net weight of irradiated...

SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation : Third Results Report

DOT National Transportation Integrated Search

2012-05-01

SunLine Transit Agency provides public transit services to the Coachella Valley area of California. SunLine has demonstrated hydrogen and fuel cell bus technologies for more than 10 years. This report describes operations at SunLine for a prototype f...

SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: Fourth Results Report

DOT National Transportation Integrated Search

2013-01-01

SunLine Transit Agency provides public transit services to the Coachella Valley area of California. SunLine has demonstrated hydrogen and fuel cell bus technologies for more than 10 years. In May 2010, SunLine began demonstrating the advanced technol...

Phase transitions in core-collapse supernova matter at sub-saturation densities

NASA Astrophysics Data System (ADS)

Pais, Helena; Newton, William G.; Stone, Jirina R.

2014-12-01

Phase transitions in hot, dense matter in the collapsing cores of massive stars have an important impact on the core-collapse supernova mechanism as they absorb heat, disrupt homology, and so weaken the developing shock. We perform a three-dimensional, finite temperature Skyrme-Hartree-Fock (SHF) study of inhomogeneous nuclear matter to determine the critical density and temperature for the phase transition between the pasta phase and homogeneous matter and its properties. We employ four different parametrizations of the Skyrme nuclear energy-density functional, SkM*, SLy4, NRAPR, and SQMC700, which span a range of saturation-density symmetry energy behaviors constrained by a variety of nuclear experimental probes. For each of these interactions we calculate free energy, pressure, entropy, and chemical potentials in the range of particle number densities where the nuclear pasta phases are expected to exist, 0.02-0.12 fm-3, temperatures 2-8 MeV, and a proton fraction of 0.3. We find unambiguous evidence for a first-order phase transition to uniform matter, unsoftened by the presence of the pasta phases. No conclusive signs of a first-order phase transition between the pasta phases is observed, and it is argued that the thermodynamic quantities vary continuously right up to the first-order phase transition to uniform matter. We compare our results with thermodynamic spinodals calculated using the same Skyrme parametrizations, finding that the effect of short-range Coulomb correlations and quantum shell effects included in our model leads to the pasta phases existing at densities up to 0.01 fm-3 above the spinodal boundaries, thus increasing the transition density to uniform matter by the same amount. The transition density is otherwise shown to be insensitive to the symmetry energy at saturation density within the range constrained by the concordance of a variety of experimental constraints, and can be taken to be a well determined quantity.

Heuristic rules embedded genetic algorithm for in-core fuel management optimization

NASA Astrophysics Data System (ADS)

Alim, Fatih

The objective of this study was to develop a unique methodology and a practical tool for designing loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and BP placement in order to achieve maximum cycle length while satisfying the safety constraints. Genetic Algorithms (GA) have been already used to solve this problem for LP optimization for both PWR and Boiling Water Reactor (BWR). The GA, which is a stochastic method works with a group of solutions and uses random variables to make decisions. Based on the theories of evaluation, the GA involves natural selection and reproduction of the individuals in the population for the next generation. The GA works by creating an initial population, evaluating it, and then improving the population by using the evaluation operators. To solve this optimization problem, a LP optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed in the framework of this thesis. This code is applicable for all types of PWR cores having different geometries and structures with an unlimited number of FA types in the inventory. To reach this goal, an innovative GA is developed by modifying the classical representation of the genotype. To obtain the best result in a shorter time, not only the representation is changed but also the algorithm is changed to use in-core fuel management heuristics rules. The improved GA code was tested to demonstrate and verify the advantages of the new enhancements. The developed methodology is explained in this thesis and preliminary results are shown for the VVER-1000 reactor hexagonal geometry core and the TMI-1 PWR. The improved GA code was tested to verify the advantages of new enhancements. The core physics code

Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

DOE Office of Scientific and Technical Information (OSTI.GOV)

Katoh, Yutai; Terrani, Kurt A.

2015-08-01

Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

Incorporating the Six Core Elements of Health Care Transition Into a Medicaid Managed Care Plan: Lessons Learned From a Pilot Project.

PubMed

McManus, Margaret; White, Patience; Pirtle, Robin; Hancock, Catina; Ablan, Michael; Corona-Parra, Raquel

2015-01-01

This pediatric-to-adult health care transition pilot project describes the process and results of incorporating the "Six Core Elements of Health Care Transition (2.0)" into a Medicaid managed care plan with a group of 35 18-23 year olds who have chronic mental health, developmental, and complex medical conditions. The pilot project demonstrated an effective approach for customizing and delivering recommended transition services. At the start of the 18-month project, the Medicaid plan was at the basic level (1) of transition implementation of the Six Core Elements with no transition policy, member transition readiness assessment results, health care transition plans of care, updated medical summaries, transfer package for the adult-focused provider, and assurance of transfer completion and consumer feedback. At the conclusion of the pilot project, the plan scored at level 3 on each core element. The primary reason for not scoring at the highest level (4) was because the transition elements have not been incorporated into services for all enrollees within the plan. Future efforts in managed care will benefit from starting the transition process much earlier (ages 12-14), expanding the role of nurse care managers and participating pediatric and adult-focused clinicians in transition, and offering payment incentives to clinicians to implement the Six Core Elements of Health Care Transition. Copyright © 2015 Elsevier Inc. All rights reserved.

Measurement of the complete core plasma flow across the LOC-SOC transition at ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Lebschy, A.; McDermott, R. M.; Angioni, C.; Geiger, B.; Prisiazhniuk, D.; Cavedon, M.; Conway, G. D.; Dux, R.; Dunne, M. G.; Kappatou, A.; Pütterich, T.; Stroth, U.; Viezzer, E.; the ASDEX Upgrade Team

2018-02-01

A newly installed core charge exchange recombination spectroscopy (CXRS) diagnostic at ASDEX Upgrade (AUG) enables the evaluation of the core poloidal rotation (upol ) through the inboard-outboard asymmetry of the toroidal rotation with an accuracy of 0.5 to 1 km s-1 . Using this technique, the total plasma flow has been measured in Ohmic L-mode plasmas across the transition from the linear to saturated ohmic confinement (LOC-SOC) regimes. The core poloidal rotation of the plasma around mid-radius is found to be always in the ion diamagnetic direction, in disagreement with neoclassical (NC) predictions. The edge rotation is found to be electron-directed and consistent with NC codes. This measurement provides as well the missing ingredient to evaluate the core E×B velocity (uE×B ) from data only, which can then be compared to measurements of the perpendicular velocity of the turbulent fluctuations (u\\perp ) to gain information on the turbulent phase velocity (vph ). The non neoclassical upol from CXRS leads to good agreement between uE×B and u\\perp indicating that vph is small and at similar values as found with gyrokinetic simulations. Moreover, the data shows a shift of vph in the ion-diamagnetic direction at the edge after the transition from LOC to SOC consistent with a change in the dominant turbulence regime. The upgrade of the core CXRS system provides as well a deeper insight into the intrinsic rotation. This paper shows that the reversal of the core toroidal rotation occurs clearly after the LOC-SOC transition and concomitant with the peaking of the electron density.

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

Core excitation effects on oscillator strengths for transitions in four electron atomic systems

NASA Astrophysics Data System (ADS)

Chang, T. N.; Luo, Yuxiang

2007-06-01

By including explicitly the electronic configurations with two and three simultaneously excited electronic orbital, we have extended the BSCI (B-spline based configuration interaction) method [1] to estimate directly the effect of inner shell core excitation to oscillator strengths for transitions in four-electron atomic systems. We will present explicitly the change in oscillator strengths due to core excitations, especially for transitions involving doubly excited states and those with very small oscillator strengths. The length and velocity results are typically in agreement better than 1% or less. [1] Tu-nan Chang, in Many-body Theory of Atomic Structure and Photoionization, edited by T. N. Chang (World Scientific, Singapore, 1993), p. 213-47; and T. N. Chang and T. K. Fang, Elsevier Radiation Physics and Chemistry 70, 173-190 (2004).

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bokhari, Ishtiaq H.

2004-12-15

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

Utilization of thorium and U-ZrH1.6 fuels in various heterogeneous cores for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

The use of thorium as nuclear fuel has been an appealing prospect for many years and will be great significance to nuclear power generation. There is an increasing need for more research on thorium as Malaysian government is currently active in the national Thorium Flagship Project, which was launched in 2014. The thorium project, which is still in phase 1, focuses on the research and development of the thorium extraction from mineral processing ore. Thus, the aim of the study is to investigate other alternative TRIGA PUSPATI Reactor (RTP) core designs that can fully utilize thorium. Currently, the RTP reactor has an average neutron flux of 2.797 x 1012 cm-2/s-1 and an effective multiplication factor, k eff, of 1.001. The RTP core has a circular array core configuration with six circular rings. Each ring consists of 6, 12, 18, 24, 30 or 36 U-ZrH1.6 fuel rods. There are three main type of uranium weight, namely 8.5, 12 and 20 wt.%. For this research, uranium zirconium hydride (U-ZrH1.6) fuel rods in the RTP core were replaced by thorium (ThO2) fuel rods. Seven core configurations with different thorium fuel rods placements were modelled in a 2D structure and simulated using Monte Carlo n-particle (MCNPX) code. Results show that the highest initial criticality obtained is around 1.35101. Additionally there is a significant discrepancy between results from previous study and the work because of the large estimated leakage probability of approximately 21.7% and 2D model simplification.

Star-forming brightest cluster galaxies at 0.25 < z < 1.25: A transitioning fuel supply

DOE PAGES

McDonald, M.; Stalder, B.; Bayliss, M.; ...

2016-01-22

In this paper, we present a multiwavelength study of the 90 brightest cluster galaxies (BCGs) in a sample of galaxy clusters selected via the Sunyaev Zel'dovich effect by the South Pole Telescope, utilizing data from various ground- and space-based facilities. We infer the star-formation rate (SFR) for the BCG in each cluster—based on the UV and IR continuum luminosity, as well as the [O ii]λλ3726,3729 emission line luminosity in cases where spectroscopy is available—and find seven systems with SFR > 100 M ⊙ yr -1. We find that the BCG SFR exceeds 10 M ⊙ yr -1 in 31 of 90 (34%) cases at 0.25 < z < 1.25, compared to ~1%–5% at z ~ 0 from the literature. At z ≳ 1, this fraction increases tomore » $${92}_{-31}^{+6}$$%, implying a steady decrease in the BCG SFR over the past ~9 Gyr. At low-z, we find that the specific SFR in BCGs is declining more slowly with time than for field or cluster galaxies, which is most likely due to the replenishing fuel from the cooling ICM in relaxed, cool core clusters. At z ≳ 0.6, the correlation between the cluster central entropy and BCG star formation—which is well established at z ~ 0—is not present. Instead, we find that the most star-forming BCGs at high-z are found in the cores of dynamically unrelaxed clusters. We use data from the Hubble Space Telescope to investigate the rest-frame near-UV morphology of a subsample of the most star-forming BCGs, and find complex, highly asymmetric UV morphologies on scales as large as ~50–60 kpc. Finally, the high fraction of star-forming BCGs hosted in unrelaxed, non-cool core clusters at early times suggests that the dominant mode of fueling star formation in BCGs may have recently transitioned from galaxy–galaxy interactions to ICM cooling.« less

Spent Fuel Test-Climax: core logging for site investigation and instrumentation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilder, D.G.; Yow, J.L. Jr.; Thorpe, R.K.

1982-05-28

As an integral part of the Spent Fuel Test-Climax 5150 ft (1570 m) of granite core was obtained. This core was diamond drilled in various sizes, mainly 38-mm and 76-mm diameters. The core was teken with single tube core barrels and was unoriented. Techniques used to drill and log this core are discussed, as well as techniques to orient the core. Of the 5150 ft (1570 m) of core more than 3645 ft (1111 m) was retained and logged in some detail. As a result of the core logging, geologic discontinuities were identified, joint frequency and spacing characterized. Discontinuities identifiedmore » included several joint sets, shear zones and faults. Correlations based on coring along were generally found to be impossible, even for the more prominent features. The only feature properly correlated from the exploratory drilling was the fault system at the end of the facility, but it was not identified from the exploratory core as a fault. Identification of discontinuities was later helped by underground mapping that identified several different joint sets with different characteristics. It was found that joint frequency varied from 0.3 to 1.1 joint per foot of core for open fractures and from 0.3 to 3.3/ft for closed or healed fractures. Histograms of fracture spacing indicate that there is likely a random distribution of spacing superimposed upon uniformly spaced fractures. It was found that a low angle joint set had a persistent mean orientation. These joints were healed and had pervasive wall rock alteration which made identification of joints in this set possible. The recognition of a joint set with known attitude allowed orientation of much of the core. This orientation technique was found to be effective. 10 references, 25 figures, 4 tables.« less

Clean air program : design guidelines for bus transit systems using hydrogen as an alternative fuel

DOT National Transportation Integrated Search

1999-04-01

Alternative fuels such as Compressed Natural Gas (CNG), Liquefied Natural Gas (LNG), Liquefied Petroleum Gas (LPG), and alcohol fuels (methanol, and ethanol) are already being used in commercial vehicles and transit buses in revenue service. Hydrogen...

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope

2011-10-01

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 physicsmore » 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

Out-of-core Evaluations of Uranium Nitride-fueled Converters

NASA Technical Reports Server (NTRS)

Shimada, K.

1972-01-01

Two uranium nitride fueled converters were tested parametrically for their initial characterization and are currently being life-tested out of core. Test method being employed for the parametric and the diagnostic measurements during the life tests, and test results are presented. One converter with a rhenium emitter had an initial output power density of 6.9 W/ sq cm at the black body emitter temperature of 1900 K. The power density remained unchanged for the first 1000 hr of life test but degraded nearly 50% percent during the following 1000 hr. Electrode work function measurements indicated that the uranium fuel was diffusing out of the emitter clad of 0.635 mm. The other converter with a tungsten emitter had an initial output power density of 2.2 W/ sq cm at 1900 K with a power density of 3.9 W/sq cm at 4300 h. The power density suddenly degraded within 20 hr to practically zero output at 4735 hr.

Design Guidelines for Bus Transit Systems Using Liquefied Petroleum Gas (LPG) as an Alternative Fuel.

DOT National Transportation Integrated Search

1996-09-01

The use of alternative fuels to power transit buses is steadily increasing. Several fuels, including Liquefied Petroleum Gas (LPG), Compressed Natural Gas (CNG), and Methanol/Ethanol, are already being used in buses. At present, there do not exist co...

Analysis of electric drive technologies for transit applications : battery-electric, hybrid-electric, and fuel cells.

DOT National Transportation Integrated Search

2005-08-01

This report provides an overview of the current status of electric drive technologies for transit applications, covering battery-electric, hybrid-electric : and fuel cell buses. Based on input from the transit and electric drive industries, the analy...

A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Senor, David J.; Painter, Chad L.; Geelhood, Ken J.

2007-12-01

Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling,more » core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.« less

From time-based to competency-based standards: core transitional competencies in plastic surgery.

PubMed

Lutz, Kristina; Yazdani, Arjang; Ross, Douglas

2015-01-01

Competency-based medical education is becoming increasingly prevalent and is likely to be mandated by the Royal College in the near future. The objective of this study was to define the core technical competencies that should be possessed by plastic surgery residents as they transition into their senior (presently postgraduate year 3) years of training. A list of potential core competencies was generated using a modified Delphi method that included the investigators and 6 experienced, academic plastic surgeons from across Canada and the United States. Generated items were divided into 7 domains: basic surgical skills, anesthesia, hand surgery, cutaneous surgery, esthetic surgery, breast surgery, and craniofacial surgery. Members of the Delphi group were asked to rank particular skills on a 4-point scale with anchored descriptors. Item reduction resulted in a survey consisting of 48 skills grouped into the aforementioned domains. This self-administered survey was distributed to all Canadian program directors (n = 11) via e-mail for validation and further item reduction. The response rate was 100% (11/11). Using the average rankings of program directors, 26 "core" skills were identified. There was agreement of core skills across all domains except for breast surgery and esthetic surgery. Of them, 7 skills were determined to be above the level of a trainee at this stage; a further 15 skills were agreed to be important, but not core, competencies. Overall, 26 competencies have been identified as "core" for plastic surgery residents to possess as they begin their senior, on-service years. The nature of these skills makes them suitable for teaching in a formal, simulated environment, which would ensure that all plastic surgery trainees are competent in these tasks as they transition to their senior years of residency. Copyright © 2014 Association of Program Directors in Surgery. Published by Elsevier Inc. All rights reserved.

Getting College-Ready during State Transition toward the Common Core State Standards

ERIC Educational Resources Information Center

Xu, Zeyu; Cepa, Kennan

2018-01-01

Background: As of 2016, 42 states and the District of Columbia have adopted the Common Core State Standards (CCSS). Tens of millions of students across the country completed high school before their schools were able to fully implement the CCSS. As with previous standards-based reforms, the transition to the CCSS-aligned state education standards…

Effect of buoyancy on fuel containment in an open-cycle gas-core nuclear rocket engine.

NASA Technical Reports Server (NTRS)

Putre, H. A.

1971-01-01

Analysis aimed at determining the scaling laws for the buoyancy effect on fuel containment in an open-cycle gas-core nuclear rocket engine, so conducted that experimental conditions can be related to engine conditions. The fuel volume fraction in a short coaxial flow cavity is calculated with a programmed numerical solution of the steady Navier-Stokes equations for isothermal, variable density fluid mixing. A dimensionless parameter B, called the Buoyancy number, was found to correlate the fuel volume fraction for large accelerations and various density ratios. This parameter has the value B = 0 for zero acceleration, and B = 350 for typical engine conditions.

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; Leland M. Montierth

2013-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters.more » One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage

HTR-proteus pebble bed experimental program core 4: random packing with a 1:1 moderator-to-fuel pebble ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Montierth, Leland M.; Sterbentz, James W.

2014-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters.more » One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage

FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Abbott, W.E.; Balent, R.

1958-09-16

A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

Isostructural solid-solid phase transition in monolayers of soft core-shell particles at fluid interfaces: structure and mechanics.

PubMed

Rey, Marcel; Fernández-Rodríguez, Miguel Ángel; Steinacher, Mathias; Scheidegger, Laura; Geisel, Karen; Richtering, Walter; Squires, Todd M; Isa, Lucio

2016-04-21

We have studied the complete two-dimensional phase diagram of a core-shell microgel-laden fluid interface by synchronizing its compression with the deposition of the interfacial monolayer. Applying a new protocol, different positions on the substrate correspond to different values of the monolayer surface pressure and specific area. Analyzing the microstructure of the deposited monolayers, we discovered an isostructural solid-solid phase transition between two crystalline phases with the same hexagonal symmetry, but with two different lattice constants. The two phases corresponded to shell-shell and core-core inter-particle contacts, respectively; with increasing surface pressure the former mechanically failed enabling the particle cores to come into contact. In the phase-transition region, clusters of particles in core-core contacts nucleate, melting the surrounding shell-shell crystal, until the whole monolayer moves into the second phase. We furthermore measured the interfacial rheology of the monolayers as a function of the surface pressure using an interfacial microdisk rheometer. The interfaces always showed a strong elastic response, with a dip in the shear elastic modulus in correspondence with the melting of the shell-shell phase, followed by a steep increase upon the formation of a percolating network of the core-core contacts. These results demonstrate that the core-shell nature of the particles leads to a rich mechanical and structural behavior that can be externally tuned by compressing the interface, indicating new routes for applications, e.g. in surface patterning or emulsion stabilization.

ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lell, R. M.; Morman, J. A.; Schaefer, R.W.

ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide,more » U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

Two-component Gaussian core model: Strong-coupling limit, Bjerrum pairs, and gas-liquid phase transition.

PubMed

Frydel, Derek; Levin, Yan

2018-01-14

In the present work, we investigate a gas-liquid transition in a two-component Gaussian core model, where particles of the same species repel and those of different species attract. Unlike a similar transition in a one-component system with particles having attractive interactions at long separations and repulsive interactions at short separations, a transition in the two-component system is not driven solely by interactions but by a specific feature of the interactions, the correlations. This leads to extremely low critical temperature, as correlations are dominant in the strong-coupling limit. By carrying out various approximations based on standard liquid-state methods, we show that a gas-liquid transition of the two-component system poses a challenging theoretical problem.

Two-component Gaussian core model: Strong-coupling limit, Bjerrum pairs, and gas-liquid phase transition

NASA Astrophysics Data System (ADS)

Frydel, Derek; Levin, Yan

2018-01-01

In the present work, we investigate a gas-liquid transition in a two-component Gaussian core model, where particles of the same species repel and those of different species attract. Unlike a similar transition in a one-component system with particles having attractive interactions at long separations and repulsive interactions at short separations, a transition in the two-component system is not driven solely by interactions but by a specific feature of the interactions, the correlations. This leads to extremely low critical temperature, as correlations are dominant in the strong-coupling limit. By carrying out various approximations based on standard liquid-state methods, we show that a gas-liquid transition of the two-component system poses a challenging theoretical problem.

Core Fueling of DEMO by Direct Line Injection of High-Speed Pellets From the HFS

DOE PAGES

Frattolillo, Antonio; Baylor, Larry R.; Bombarda, Francesca; ...

2018-04-17

Pellet injection represents to date the most realistic candidate technology for core fueling of a demonstration fusion power reactor tokamak fusion reactor. Modeling of both pellet penetration and fuel deposition profiles, for different injection locations, indicates that effective core fuelling can be achieved launching pellets from the inboard high field side at speeds not less than ~ 1 km/s. Inboard pellet fueling is commonly achieved in present tokamaks, using curved guide tubes; however, this technology might be hampered at velocities ≥ 1 km/s. An innovative approach, aimed at identifying suitable inboard "direct line'' paths, to inject high-speed pellets (in themore » 3 to 4 km/s range), has recently been proposed as a potential complementary solution. The fuel deposition profiles achievable by this approach have been explored using the HPI2 simulation code. The results presented here show that there are possible geometrical schemes providing good fueling performance. The problem of neutron flux in a direct line-of-sight injection path is being investigated, though preliminary analyses indicate that, perhaps, this is not a serious problem. The identification and integration of straight injection paths suitably tilted may be a rather difficult task due to the many constraints and to interference with existing structures. The suitability of straight guide tubes to reduce the scatter cone of high-speed pellets is, therefore, of main interest. A preliminary investigation, aimed at addressing these technological issues, has recently been started. As a result, a possible implementation plan, using an existing Italian National Agency for New Technologies, Energy and Sustainable Economic Development-Oak Ridge National Laboratory facility is shortly outlined.« less

Core Fueling of DEMO by Direct Line Injection of High-Speed Pellets From the HFS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frattolillo, Antonio; Baylor, Larry R.; Bombarda, Francesca

Pellet injection represents to date the most realistic candidate technology for core fueling of a demonstration fusion power reactor tokamak fusion reactor. Modeling of both pellet penetration and fuel deposition profiles, for different injection locations, indicates that effective core fuelling can be achieved launching pellets from the inboard high field side at speeds not less than ~ 1 km/s. Inboard pellet fueling is commonly achieved in present tokamaks, using curved guide tubes; however, this technology might be hampered at velocities ≥ 1 km/s. An innovative approach, aimed at identifying suitable inboard "direct line'' paths, to inject high-speed pellets (in themore » 3 to 4 km/s range), has recently been proposed as a potential complementary solution. The fuel deposition profiles achievable by this approach have been explored using the HPI2 simulation code. The results presented here show that there are possible geometrical schemes providing good fueling performance. The problem of neutron flux in a direct line-of-sight injection path is being investigated, though preliminary analyses indicate that, perhaps, this is not a serious problem. The identification and integration of straight injection paths suitably tilted may be a rather difficult task due to the many constraints and to interference with existing structures. The suitability of straight guide tubes to reduce the scatter cone of high-speed pellets is, therefore, of main interest. A preliminary investigation, aimed at addressing these technological issues, has recently been started. As a result, a possible implementation plan, using an existing Italian National Agency for New Technologies, Energy and Sustainable Economic Development-Oak Ridge National Laboratory facility is shortly outlined.« less

Fuel transfer system

DOEpatents

Townsend, Harold E.; Barbanti, Giancarlo

1994-01-01

A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool.

Reducing transit bus emissions: Alternative fuels or traffic operations?

NASA Astrophysics Data System (ADS)

Alam, Ahsan; Hatzopoulou, Marianne

2014-06-01

In this study, we simulated the operations and greenhouse gas (GHG) emissions of transit buses along a busy corridor and quantified the effects of two different fuels (conventional diesel and compressed natural gas) as well as a set of driving conditions on emissions. Results indicate that compressed natural gas (CNG) reduces GHG emissions by 8-12% compared to conventional diesel, this reduction could increase to 16% with high levels of traffic congestion. However, the benefits of switching from conventional diesel to CNG are less apparent when the road network is uncongested. We also investigated the effects of bus operations on emissions by applying several strategies such as transit signal priority (TSP), queue jumper lanes, and relocation of bus stops. Results show that in congested conditions, TSP alone can reduce GHG emissions by 14% and when combined with improved technology; a reduction of 23% is achieved. The reduction benefits are even more apparent when other transit operational improvements are combined with TSP. Finally a sensitivity analysis was performed to investigate the effect of operational improvements on emissions under varying levels of network congestion. We observe that under “extreme congestion”, the benefits of TSP decrease.

Predicted trends of core-shell preferences for 132 late transition-metal binary-alloy nanoparticles.

PubMed

Wang, Lin-Lin; Johnson, Duane D

2009-10-07

Transition-metal alloyed nanoparticles with core-shell features (shell enrichment by one of the metals) are becoming ubiquitous, from (electro-)catalysis to biomedical applications, due to their size control, performance, biocompatibility, and cost. We investigate 132 binary-alloyed nanoparticle systems (groups 8 to 11 in the Periodic Table) using density functional theory (DFT) and systematically explore their segregation energies to determine core-shell preferences. We find that core-shell preferences are generally described by two independent factors: (1) cohesive energy (related to vapor pressure) and (2) atomic size (quantified by the Wigner-Seitz radius), and the interplay between them. These independent factors are shown to provide general trends for the surface segregation preference for atoms in nanoparticles, as well as semi-infinite surfaces, and give a simple correlation (a "design map") for the alloying and catalytic behavior. Finally, we provide a universal description of core-shell preference via tight-binding theory (band-energy differences) that (i) quantitatively reproduces the DFT segregation energies and (ii) confirms the electronic origins and correlations for core-shell behavior.

Core-core and core-valence correlation

NASA Technical Reports Server (NTRS)

Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.

1988-01-01

The effect of (1s) core correlation on properties and energy separations was analyzed using full configuration-interaction (FCI) calculations. The Be 1 S - 1 P, the C 3 P - 5 S and CH+ 1 Sigma + or - 1 Pi separations, and CH+ spectroscopic constants, dipole moment and 1 Sigma + - 1 Pi transition dipole moment were studied. The results of the FCI calculations are compared to those obtained using approximate methods. In addition, the generation of atomic natural orbital (ANO) basis sets, as a method for contracting a primitive basis set for both valence and core correlation, is discussed. When both core-core and core-valence correlation are included in the calculation, no suitable truncated CI approach consistently reproduces the FCI, and contraction of the basis set is very difficult. If the (nearly constant) core-core correlation is eliminated, and only the core-valence correlation is included, CASSCF/MRCI approached reproduce the FCI results and basis set contraction is significantly easier.

Clean Air Program : Design Guidelines for Bus Transit Systems Using Liquefied Natural Gas (LNG) as an Alternative Fuel

DOT National Transportation Integrated Search

1997-03-01

The use of alternative fuels to power transit buses is steadily increasing. Several fuels, including Liquefied Natural Gas (LNG), Compressed Natural Gas (CNG), Liquefied Petroleum Gas (LPG), and Methanol/Ethanol, are already being used. At present, t...

Fuel transfer system

DOEpatents

Townsend, H.E.; Barbanti, G.

1994-03-01

A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool. 6 figures.

Effective core potential calculations on small molecules containing transition metal atoms

NASA Astrophysics Data System (ADS)

Gropen, O.; Wahlgren, U.; Pettersson, L.

1982-04-01

A series of test calculations on diatomic oxides and hydrides of Sc, Ti, Cr, Ni and Zn have been carried out in order to test the reliability of some pseudopotential methods. Several different forms of some pseudopotential operators were used. Only the highest valence orbitals of each atomic symmetry were explicitly included in the calculations. The results indicate that there are problems associated with all the investigated operators particularly for the lighter transition elements. It is suggested that more reliable results may be obtained with pseudopotential methods using smaller cores.

TEMperature Pressure ESTimation of a homogeneous boiling fuel-steel mixture in an LMFBR core. [TEMPEST code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pyun, J.J.; Majumdar, D.

The paper describes TEMPEST, a simple computer program for the temperature and pressure estimation of a boiling fuel-steel pool in an LMFBR core. The time scale of interest of this program is large, of the order of ten seconds. Further, the vigorous boiling in the pool will generate a large contact, and hence a large heat transfer between fuel and steel. The pool is assumed to be a uniform mixture of fuel and steel, and consequently vapor production is also assumed to be uniform throughout the pool. The pool is allowed to expand in volume if there is steel meltingmore » at the walls. In this program, the total mass of liquid and vapor fuel is always kept constant, but the total steel mass in the pool may change by steel wall melting. Because of a lack of clear understanding of the physical phenomena associated with the progression of a fuel-steel mixture at high temperature, various input options have been built-in to enable one to perform parametric studies. For example, the heat transfer from the pool to the surrounding steel structure may be controlled by input values for the heat transfer coefficients, or, the heat transfer may be calculated by a correlation obtained from the literature. Similarly, condensation of vapor on the top wall can be specified by input values of the condensation coefficient; the program can otherwise calculate condensation according to the non-equilibrium model predictions. Meltthrough rates of the surrounding steel walls can be specified by a fixed melt-rate or can be determined by a fraction of the heat loss that goes to steel-melting. The melted steel is raised to the pool temperature before it is joined with the pool material. Several applications of this program to various fuel-steel pools in the FFTF and the CRBR cores are discussed.« less

Clean air program : design guidelines for bus transit systems using electric and hybrid electric propulsion as an alternative fuel

DOT National Transportation Integrated Search

2003-03-01

The use of alternative fuels to power transit buses is steadily increasing. Several fuels, including : Compressed Natural Gas (CNG), Liquefied Natural Gas (LNG), Liquefied Petroleum Gas (LPG), and : Methanol/Ethanol, are already being used. At presen...

Gravitational-Wave and Neutrino Signals from Core-Collapse Supernovae with QCD Phase Transition

NASA Astrophysics Data System (ADS)

Zha, Shuai; Leung, Shing Chi; Lin, Lap Ming; Chu, Ming-Chung

Core-collapse supernovae (CCSNe) mark the catastrophic death of massive stars. We simulate CCSNe with a hybrid equations of state (EOS) containing a QCD (quantum chromodynamics) phase transition. The hybrid EOS incorporates the pure hadronic HShen EOS and the MIT Bag Model, with a Gibbs construction. Our two-dimensional hydrodynamics code includes a fifth-order shock capturing scheme WENO and models neutrino transport with the isotropic diffusion source approximation (IDSA). As the proto-neutron-star accretes matter and the core enters the mixed phase, a second collapse takes place due to softening of the EOS. We calculate the gravitational-wave (GW) and neutrino signals for this kind of CCSNe model. Future detection of these signals from CCSNe may help to constrain this scenario and the hybrid EOS.

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

Evaluation of in-use fuel economy and on-board emissions for hybrid and regular CyRide transit buses.

DOT National Transportation Integrated Search

2012-10-01

The objective of this project was to evaluate the in-use fuel economy and emission differences between hybrid-electric and : conventional transit buses for the Ames, Iowa transit authority, CyRide. These CyRide buses were deployed in the fall of : 20...

Accidental degeneracy in photonic bands and topological phase transitions in two-dimensional core-shell dielectric photonic crystals.

PubMed

Xu, Lin; Wang, Hai-Xiao; Xu, Ya-Dong; Chen, Huan-Yang; Jiang, Jian-Hua

2016-08-08

A simple core-shell two-dimensional photonic crystal is studied where the triangular lattice symmetry and the C₆ point group symmetry give rich physics in accidental touching points of photonic bands. We systematically evaluate different types of accidental nodal points at the Brillouin zone center for transverse-magnetic harmonic modes when the geometry and permittivity of the core-shell material are continuously tuned. The accidental nodal points can have different dispersions and topological properties (i.e., Berry phases). These accidental nodal points can be the critical states lying between a topological phase and a normal phase of the photonic crystal. They are thus very important for the study of topological photonic states. We show that, without breaking time-reversal symmetry, by tuning the geometry of the core-shell material, a phase transition into the photonic quantum spin Hall insulator can be achieved. Here the "spin" is defined as the orbital angular momentum of a photon. We study the topological phase transition as well as the properties of the edge and bulk states and their application potentials in optics.

Hierarchy effect on electronic structure and core-to-valence transitions in bone tissue: perspectives in medical nanodiagnostics of mineralized bone

NASA Astrophysics Data System (ADS)

Samoilenko, Dmitrii O.; Avrunin, Alexander S.; Pavlychev, Andrey A.

2017-06-01

Electronic structure and core-to-valence transitions in bone tissue are examined in the framework of the morphological 3DSL model that takes into account (i) structural and functional organization of the skeleton in the normal and pathological conditions and (ii) peculiarities of electron wave propagation in a three-dimensional superlattice of "black-nanocrystallites-in-muddy-waters". Our focus is on the HAP-to-bone red shifts of core-to-valence transitions near Ca and P 2p and O 1s edges in single-crystal hydroxyapatite (HAP) Ca10(PO4)6(OH)2. The origin of the HAP-to-bone shift is discussed and the extended comparative analysis of the experimental data is performed. The detected spectral shift is assigned with the effect of hierarchical organization of bone tissue. This hierarchy effect on the core-to-valence transition energies is regarded as a promising tool for medical imaging and perspective pathway for nanodiagnostics of mineralized bone. Contribution to the Topical Issue "Dynamics of Systems at the Nanoscale", edited by Andrey Solov'yov and Andrei Korol.

Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

NASA Astrophysics Data System (ADS)

Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

2017-07-01

Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (Dfuel with fissile contain from Plutonium waste LWR for each geometry. The minimum power density is around 72 Watt/cc, and maximum power density 114 Watt/cc. After we calculate with various geometry core, when we use the balance geometry, the k-eff value flattest and more stable than the others.

LMFBR fuel assembly design for HCDA fuel dispersal

DOEpatents

Lacko, Robert E.; Tilbrook, Roger W.

1984-01-01

A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

Solid oxide fuel cell with transitioned cross-section for improved anode gas management at the open end

DOEpatents

Zafred, Paolo R [Murrysville, PA; Draper, Robert [Pittsburgh, PA

2012-01-17

A solid oxide fuel cell (400) is made having a tubular, elongated, hollow, active section (445) which has a cross-section containing an air electrode (452) a fuel electrode (454) and solid oxide electrolyte (456) between them, where the fuel cell transitions into at least one inactive section (460) with a flattened parallel sided cross-section (462, 468) each cross-section having channels (472, 474, 476) in them which smoothly communicate with each other at an interface section (458).

CERDEC Fuel Cell Team: Military Transitions for Soldier Fuel Cells

DTIC Science & Technology

2008-10-27

Fuel Cell (DMFC) (PEO Soldier) Samsung: 20W DMFC (CRADA) General Atomics & Jadoo: 50W Ammonia Borane Fueled PEMFC Current Fuel Cell Team Efforts...Continued Ardica: 20W Wearable PEMFC operating on Chemical Hydrides Spectrum Brands w/ Rayovac: Hydrogen Generators and Alkaline Fuel Cells for AA...100W Ammonia Borane fueled PEMFC Ultralife: 150W sodium borohydride fueled PEMFC Protonex: 250W RMFC and Power Manager (ARO) NanoDynamics: 250W SOFC

Core-shell fuel cell electrodes

DOEpatents

Adzic, Radoslav; Bliznakov, Stoyan; Vukmirovic, Miomir

2017-07-25

Embodiments of the disclosure relate to electrocatalysts. The electrocatalyst may include at least one gas-diffusion layer having a first side and a second side, and particle cores adhered to at least one of the first and second sides of the at least one gas-diffusion layer. The particle cores includes surfaces adhered to the at least one of the first and second sides of the at least one gas-diffusion layer and surfaces not in contact with the at least one gas-diffusion layer. Furthermore, a thin layer of catalytically atoms may be adhered to the surfaces of the particle cores not in contact with the at least one gas-diffusion layer.

Core-shell fuel cell electrodes

DOEpatents

Adzic, Radoslav; Bliznakov, Stoyan; Vukmirovic, Miomir

2017-12-26

Embodiments of the disclosure relate to membrane electrode assemblies. The membrane electrode assembly may include at least one gas-diffusion layer having a first side and a second side, and particle cores adhered to at least one of the first and second sides of the at least one gas-diffusion layer. The particle cores includes surfaces adhered to the at least one of the first and second sides of the at least one gas-diffusion layer and surfaces not in contact with the at least one gas-diffusion layer. Furthermore, a thin layer of catalytically atoms may be adhered to the surfaces of the particle cores not in contact with the at least one gas-diffusion layer.

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heidet, F.; Kim, T.; Grandy, C.

2012-07-30

Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium ismore » more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to

FUEL ELEMENT CONSTRUCTION

DOEpatents

Zumwalt, L.R.

1961-08-01

Fuel elements having a solid core of fissionable material encased in a cladding material are described. A conversion material is provided within the cladding to react with the fission products to form stable, relatively non- volatile compounds thereby minimizing the migration of the fission products into the coolant. The conversion material is preferably a metallic fluoride, such as lead difluoride, and may be in the form of a coating on the fuel core or interior of the cladding, or dispersed within the fuel core. (AEC)

Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region

NASA Technical Reports Server (NTRS)

Klann, P. G.; Lantz, E.

1973-01-01

A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

California's Transition to the Common Core State Standards: The State's Role in Local Capacity Building

ERIC Educational Resources Information Center

Warren, Paul; Murphy, Patrick

2014-01-01

The Common Core State Standards (CCSS) and the Local Control Funding Formula are introducing major changes to California's K-12 system. Implementation of new curricula and instruction is under way at the district level, but California started its transition relatively late and it has taken a more decentralized approach than most other states.…

High performance of SDC and GDC core shell type composite electrolytes using methane as a fuel for low temperature SOFC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Irshad, Muneeb; Siraj, Khurram, E-mail: razahussaini786@gmail.com, E-mail: khurram.uet@gmail.com; Javed, Fayyaz

Nanocomposites Samarium doped Ceria (SDC), Gadolinium doped Ceria (GDC), core shell SDC amorphous Na{sub 2}CO{sub 3} (SDCC) and GDC amorphous Na{sub 2}CO{sub 3} (GDCC) were synthesized using co-precipitation method and then compared to obtain better solid oxide electrolytes materials for low temperature Solid Oxide Fuel Cell (SOFCs). The comparison is done in terms of structure, crystallanity, thermal stability, conductivity and cell performance. In present work, XRD analysis confirmed proper doping of Sm and Gd in both single phase (SDC, GDC) and dual phase core shell (SDCC, GDCC) electrolyte materials. EDX analysis validated the presence of Sm and Gd in bothmore » single and dual phase electrolyte materials; also confirming the presence of amorphous Na{sub 2}CO{sub 3} in SDCC and GDCC. From TGA analysis a steep weight loss is observed in case of SDCC and GDCC when temperature rises above 725 °C while SDC and GDC do not show any loss. The ionic conductivity and cell performance of single phase SDC and GDC nanocomposite were compared with core shell GDC/amorphous Na{sub 2}CO{sub 3} and SDC/ amorphous Na{sub 2}CO{sub 3} nanocomposites using methane fuel. It is observed that dual phase core shell electrolytes materials (SDCC, GDCC) show better performance in low temperature range than their corresponding single phase electrolyte materials (SDC, GDC) with methane fuel.« less

Extended burnup core management for once-through uranium fuel cycles in LWRS. First annual report for the period 1 July 1979-30 June 1980

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sesonske, A.

1980-08-01

Detailed core management arrangements are developed requiring four operating cycles for the transition from present three-batch loading to an extended burnup four-batch plan for Zion-1. The ARMP code EPRI-NODE-P was used for core modeling. Although this work is preliminary, uranium and economic savings during the transition cycles appear of the order of 6 percent.

Random pinning elucidates the nature of melting transition in two-dimensional core-softened potential system

NASA Astrophysics Data System (ADS)

Tsiok, E. N.; Fomin, Y. D.; Ryzhov, V. N.

2018-01-01

Despite about forty years of investigations, the nature of the melting transition in two dimensions is not completely clear. In the framework of the most popular Berezinskii-Kosterlitz-Thouless-Halperin-Nelson-Young (BKTHNY) theory, 2D systems melt through two continuous Berezinskii-Kosterlitz-Thouless (BKT) transitions with intermediate hexatic phase. The conventional first-order transition is also possible. On the other hand, recently on the basis of computer simulations the new melting scenario was proposed with continuous BKT type solid-hexatic transition and first order hexatic-liquid transition. However, in the simulations the hexatic phase is extremely narrow that makes its study difficult. In the present paper, we propose to apply the random pinning to investigate the hexatic phase in more detail. The results of molecular dynamics simulations of two dimensional system having core-softened potentials with narrow repulsive step which is similar to the soft disk system are outlined. The system has a small fraction of pinned particles giving quenched disorder. Random pinning widens the hexatic phase without changing the melting scenario and gives the possibility to study the behavior of the diffusivity and order parameters in the vicinity of the melting transition and inside the hexatic phase.

Thermo-Physics Technical Note No. 60: thermal analysis of SNAP 10A reactor core during atmospheric reentry and resulting core disintegration and fuel element separation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mouradian, E.M.

1966-02-16

A thermal analysis is carried out to determine the temperature distribution throughout a SNAP 10A reactor core, particularly in the vicinity of the grid plates, during atmospheric reentry. The transient temperatue distribution of the grid plate indicates when sufficient melting occurs so that fuel elements are free to be released and continue their descent individually.

The influence of transition metal solutes on the dislocation core structure and values of the Peierls stress and barrier in tungsten

NASA Astrophysics Data System (ADS)

Samolyuk, G. D.; Osetsky, Y. N.; Stoller, R. E.

2013-01-01

Several transition metals were examined to evaluate their potential for improving the ductility of tungsten. The dislocation core structure and Peierls stress and barrier of 1/2<111> screw dislocations in binary tungsten-transition metal alloys (W1-xTMx) were investigated using density functional theory calculations. The periodic quadrupole approach was applied to model the structure of the 1/2<111> dislocation. Alloying with transition metals was modeled using the virtual crystal approximation and the applicability of this approach was assessed by calculating the equilibrium lattice parameter and elastic constants of the tungsten alloys. Reasonable agreement was obtained with experimental data and with results obtained from the conventional supercell approach. Increasing the concentration of a transition metal from the VIIIA group, i.e. the elements in columns headed by Fe, Co and Ni, leads to reduction of the C‧ elastic constant and increase of the elastic anisotropy A = C44/C‧. Alloying W with a group VIIIA transition metal changes the structure of the dislocation core from symmetric to asymmetric, similarly to results obtained for W1-xRex alloys in the earlier work of Romaner et al (2010 Phys. Rev. Lett. 104 195503). In addition to a change in the core symmetry, the values of the Peierls stress and barrier are reduced. The latter effect could lead to increased ductility in a tungsten-based alloy. Our results demonstrate that alloying with any of the transition metals from the VIIIA group should have a similar effect to alloying with Re.

Iridium-decorated palladium-platinum core-shell catalysts for oxygen reduction reaction in proton exchange membrane fuel cell.

PubMed

Wang, Chen-Hao; Hsu, Hsin-Cheng; Wang, Kai-Ching

2014-08-01

Carbon-supported Pt, Pd, Pd-Pt core-shell (Pt(shell)-Pd(core)/C) and Ir-decorated Pd-Pt core-shell (Ir-decorated Pt(shell)-Pd(core)/C) catalysts were synthesized, and their physical properties, electrochemical behaviors, oxygen reduction reaction (ORR) characteristics and proton exchange membrane fuel cell (PEMFC) performances were investigated herein. From the XRD patterns and TEM images, Ir-decorated Pt(shell)-Pd(core)/C has been confirmed that Pt was deposited on the Pd nanoparticle which had the core-shell structure. Ir-decorated Pt(shell)-Pd(core)/C has more positive OH reduction peak than Pt/C, which is beneficial to weaken the binding energy of Pt-OH during the ORR. Thus, Ir-decorated Pt(shell)-Pd(core)/C has higher ORR activity than Pt/C. The maximum power density of H2-O2 PEMFC using Ir-decorated Pt(shell)-Pd(core)/C is 792.2 mW cm(-2) at 70°C, which is 24% higher than that using Pt/C. The single-cell accelerated degradation test of PEMFC using Ir-decorated Pt(shell)-Pd(core)/C shows good durability by the potential cycling of 40,000 cycles. This study concludes that Ir-decorated Pt(shell)-Pd(core)/C has the low Pt content, but it can facilitate the low-cost and high-efficient PEMFC. Copyright © 2013 Elsevier Inc. All rights reserved.

Apparatus for and method of monitoring for breached fuel elements

DOEpatents

Gross, K.C.; Strain, R.V.

1981-04-28

This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus uses a separate bypass loop for conveying part of the reactor coolant away from the core, and at least three separate delayed-neutron detectors mounted proximate this detector loop. The detectors are spaced apart so that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the delay time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector.

Transitions, Risks, and Actions in Coronary Events--Center for Outcomes Research and Education (TRACE-CORE): design and rationale.

PubMed

Waring, Molly E; McManus, Richard H; Saczynski, Jane S; Anatchkova, Milena D; McManus, David D; Devereaux, Randolph S; Goldberg, Robert J; Allison, Jeroan J; Kiefe, Catarina I

2012-09-01

Cardiovascular disease continues to cause significant morbidity, mortality, and impaired quality of life, with unrealized health gains from the underuse of available evidence. The Transitions, Risks, and Actions in Coronary Events Center for Outcomes Research and Education (TRACE-CORE) aims to advance the science of acute coronary syndromes by examining the determinants and outcomes of the quality of transition from hospital to community and by quantifying the impact of potentially modifiable characteristics associated with decreased quality of life, rehospitalization, and mortality. TRACE-CORE comprises a longitudinal multiracial cohort of patients hospitalized with acute coronary syndromes, 2 research projects, and development of a nucleus of early stage investigators. We are currently enrolling 2500 adults hospitalized for acute coronary syndromes at 6 hospitals in the northeastern and southeastern United States. We will follow these patients for 24 months after hospitalization through medical record abstraction and 5 patient interviews focusing on quality of life, cardiac events, rehospitalizations, mortality, and medical, behavioral, and psychosocial characteristics. The Transitions Project studies determinants of and disparities in outcomes of the quality of patients' transition from hospital to community. Focusing on potentially modifiable factors, the Action Scores Project will develop and validate action scores to predict recurrent cardiac events, death, and quality of life, describe longitudinal variation in these scores, and develop a dashboard for patient and provider action on the basis of these scores. In TRACE-CORE, sound methodologic principles of observational studies converge with outcomes and effectiveness research approaches. We expect that our data, research infrastructure, and research projects will inform the development of novel secondary prevention approaches and underpin the careers of cardiovascular outcomes researchers.

A ``NEW'' Solid-Core Reactor Fuel Form that Maximizes the Performance of Nuclear Thermal and Electric Rockets

NASA Astrophysics Data System (ADS)

Rom, Frank E.; Finnegan, Patrick M.

1994-07-01

The ``NEW'' solid-core fuel form is the old Vapor Transport (VT) fuel pin investigated at NASA about 30 years ago. It is simply a tube sealed at both ends partially filled with UO2. During operation the UO2 forms an annular layer on the inside of the tube by vaporization and condensation. This form is an ideal structure for overall strength and retention of fission products. All of the structural material lies between the fuel (including fission products) and the reactor coolant. The isothermal inside fuel surface temperature that results from the vaporization and condensation of fuel during operation eliminates hotspots, significantly increasing the design fuel pin surface temperature. For NTP, W-UO2 fuel pins yield higher operating temperatures than for other fuel forms, because W has about a ten-fold lower vaporization rate compared to any other known material. The use of perigee propulsion using W-UO2 fuel pins can result in a more than ten-fold reduction in reactor power. Lower reactor power, together with zero fission product release potential, and the simplicity of fabrication of VT fuel pins should greatly simplify and reduce the cost of development of NTP. For NEP, VT fuel pins can increase fast neutron spectrum reactor life with no fission product release. Thermal spectrum NEP reactors using W184 or Mo VT fuel pins, with only small amounts of high neutron absorbing additives, offer benefits because of much lower fissionable fuel requirements. The VT fuel pin has application to commercial power reactors with similar benefits.

Modeling transit bus fuel consumption on the basis of cycle properties.

PubMed

Delgado, Oscar F; Clark, Nigel N; Thompson, Gregory J

2011-04-01

A method exists to predict heavy-duty vehicle fuel economy and emissions over an "unseen" cycle or during unseen on-road activity on the basis of fuel consumption and emissions data from measured chassis dynamometer test cycles and properties (statistical parameters) of those cycles. No regression is required for the method, which relies solely on the linear association of vehicle performance with cycle properties. This method has been advanced and examined using previously published heavy-duty truck data gathered using the West Virginia University heavy-duty chassis dynamometer with the trucks exercised over limited test cycles. In this study, data were available from a Washington Metropolitan Area Transit Authority emission testing program conducted in 2006. Chassis dynamometer data from two conventional diesel buses, two compressed natural gas buses, and one hybrid diesel bus were evaluated using an expanded driving cycle set of 16 or 17 different driving cycles. Cycle properties and vehicle fuel consumption measurements from three baseline cycles were selected to generate a linear model and then to predict unseen fuel consumption over the remaining 13 or 14 cycles. Average velocity, average positive acceleration, and number of stops per distance were found to be the desired cycle properties for use in the model. The methodology allowed for the prediction of fuel consumption with an average error of 8.5% from vehicles operating on a diverse set of chassis dynamometer cycles on the basis of relatively few experimental measurements. It was found that the data used for prediction should be acquired from a set that must include an idle cycle along with a relatively slow transient cycle and a relatively high speed cycle. The method was also applied to oxides of nitrogen prediction and was found to have less predictive capability than for fuel consumption with an average error of 20.4%.

Fuels for urban transit buses: a cost-effectiveness analysis.

PubMed

Cohen, Joshua T; Hammitt, James K; Levy, Jonathan I

2003-04-15

Public transit agencies have begun to adopt alternative propulsion technologies to reduce urban transit bus emissions associated with conventional diesel (CD) engines. Among the most popular alternatives are emission controlled diesel buses (ECD), defined here to be buses with continuously regenerating diesel particle filters burning low-sulfur diesel fuel, and buses burning compressed natural gas (CNG). This study uses a series of simplifying assumptions to arrive at first-order estimates for the incremental cost-effectiveness (CE) of ECD and CNG relative to CD. The CE ratio numerator reflects acquisition and operating costs. The denominator reflects health losses (mortality and morbidity) due to primary particulate matter (PM), secondary PM, and ozone exposure, measured as quality adjusted life years (QALYs). We find that CNG provides larger health benefits than does ECD (nine vs six QALYs annually per 1000 buses) but that ECD is more cost-effective than CNG (dollar 270 000 per QALY for ECD vs dollar 1.7 million to dollar 2.4 million for CNG). These estimates are subject to much uncertainty. We identify assumptions that contribute most to this uncertainty and propose potential research directions to refine our estimates.

Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

C. Fiorina; N. E. Stauff; F. Franceschini

2013-12-01

The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associatedmore » with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.« less

Simulating Valence-to-Core X-ray Emission Spectroscopy of Transition Metal Complexes with Time-Dependent Density Functional Theory.

PubMed

Zhang, Yu; Mukamel, Shaul; Khalil, Munira; Govind, Niranjan

2015-12-08

Valence-to-core (VtC) X-ray emission spectroscopy (XES) has emerged as a powerful technique for the structural characterization of complex organometallic compounds in realistic environments. Since the spectrum represents electronic transitions from the ligand molecular orbitals to the core holes of the metal centers, the approach is more chemically sensitive to the metal-ligand bonding character compared with conventional X-ray absorption techniques. In this paper we study how linear-response time-dependent density functional theory (LR-TDDFT) can be harnessed to simulate K-edge VtC X-ray emission spectra reliably. LR-TDDFT allows one to go beyond the single-particle picture that has been extensively used to simulate VtC-XES. We consider seven low- and high-spin model complexes involving chromium, manganese, and iron transition metal centers. Our results are in good agreement with experiment.

Au/ZnS core/shell nanocrystals as an efficient anode photocatalyst in direct methanol fuel cells.

PubMed

Chen, Wei-Ta; Lin, Yin-Kai; Yang, Ting-Ting; Pu, Ying-Chih; Hsu, Yung-Jung

2013-10-04

Au/ZnS core/shell nanocrystals with controllable shell thicknesses were synthesized using a cysteine-assisted hydrothermal method. Incorporating Au/ZnS nanocrystals into the traditional Pt-catalyzed half-cell reaction led to a 43.3% increase in methanol oxidation current under light illumination, demonstrating their promising potential for metal/semiconductor hybrid nanocrystals as the anode photocatalyst in direct methanol fuel cells.

Simulations of the Fuel Economy and Emissions of Hybrid Transit Buses over Planned Local Routes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gao, Zhiming; LaClair, Tim J; Daw, C Stuart

2014-01-01

We present simulated fuel economy and emissions city transit buses powered by conventional diesel engines and diesel-hybrid electric powertrains of varying size. Six representative city drive cycles were included in the study. In addition, we included previously published aftertreatment device models for control of CO, HC, NOx, and particulate matter (PM) emissions. Our results reveal that bus hybridization can significantly enhance fuel economy by reducing engine idling time, reducing demands for accessory loads, exploiting regenerative braking, and shifting engine operation to speeds and loads with higher fuel efficiency. Increased hybridization also tends to monotonically reduce engine-out emissions, but trends inmore » the tailpipe (post-aftertreatment) emissions involve more complex interactions that significantly depend on motor size and drive cycle details.« less

Transuranic Waste Burning Potential of Thorium Fuel in a Fast Reactor - 12423

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wenner, Michael; Franceschini, Fausto; Ferroni, Paolo

Westinghouse Electric Company (referred to as 'Westinghouse' in the rest of this paper) is proposing a 'back-to-front' approach to overcome the stalemate on nuclear waste management in the US. In this approach, requirements to further the societal acceptance of nuclear waste are such that the ultimate health hazard resulting from the waste package is 'as low as reasonably achievable'. Societal acceptability of nuclear waste can be enhanced by reducing the long-term radiotoxicity of the waste, which is currently driven primarily by the protracted radiotoxicity of the transuranic (TRU) isotopes. Therefore, a transition to a more benign radioactive waste can bemore » accomplished by a fuel cycle capable of consuming the stockpile of TRU 'legacy' waste contained in the LWR Used Nuclear Fuel (UNF) while generating waste which is significantly less radio-toxic than that produced by the current open U-based fuel cycle (once through and variations thereof). Investigation of a fast reactor (FR) operating on a thorium-based fuel cycle, as opposed to the traditional uranium-based is performed. Due to a combination between its neutronic properties and its low position in the actinide chain, thorium not only burns the legacy TRU waste, but it does so with a minimal production of 'new' TRUs. The effectiveness of a thorium-based fast reactor to burn legacy TRU and its flexibility to incorporate various fuels and recycle schemes according to the evolving needs of the transmutation scenario have been investigated. Specifically, the potential for a high TRU burning rate, high U-233 generation rate if so desired and low concurrent production of TRU have been used as metrics for the examined cycles. Core physics simulations of a fast reactor core running on thorium-based fuels and burning an external TRU feed supply have been carried out over multiple cycles of irradiation, separation and reprocessing. The TRU burning capability as well as the core isotopic content have been

Nuclear core positioning system

DOEpatents

Garkisch, Hans D.; Yant, Howard W.; Patterson, John F.

1979-01-01

A structural support system for the core of a nuclear reactor which achieves relatively restricted clearances at operating conditions and yet allows sufficient clearance between fuel assemblies at refueling temperatures. Axially displaced spacer pads having variable between pad spacing and a temperature compensated radial restraint system are utilized to maintain clearances between the fuel elements. The core support plates are constructed of metals specially chosen such that differential thermal expansion produces positive restraint at operating temperatures.

The Core Competencies and MFT Education: Practical Aspects of Transitioning to a Learning-Centered, Outcome-Based Pedagogy

ERIC Educational Resources Information Center

Gehart, Diane

2011-01-01

The MFT core competencies and latest COAMFTE accreditation standards usher in a new paradigm for MFT education. This transition necessitates not only measuring student mastery of competencies but also, more importantly, adopting a contemporary pedagogical model. This article provides an overview of the changes, a review of parallel trends in other…

CS Line Profiles in Hot Cores

NASA Astrophysics Data System (ADS)

Bayet, E.; Yates, J.; Viti, S.

2011-02-01

We present a theoretical study of CS line profiles in archetypal hot cores. We provide estimates of line fluxes from the CS(1-0) to the CS(15-14) transitions and present the temporal variation of these fluxes. We find that (1) the CS(1-0) transition is a better tracer of the Envelope of the hot core whereas the higher-J CS lines trace the ultracompact core (UCC); (2) the peak temperature of the CS transitions is a good indicator of the temperature inside the hot core; (3) in the Envelope, the older the hot core the stronger the self-absorption of CS; (4) the fractional abundance of CS is highest in the innermost parts of the UCC, confirming the CS molecule as one of the best tracers of very dense gas.

HPLC method for rapidly following biodiesel fuel transesterification reaction progress using a core-shell column.

PubMed

Allen, Samuel J; Ott, Lisa S

2012-07-01

There are a wide and growing variety of feedstocks for biodiesel fuel. Most commonly, these feedstocks contain triglycerides which are transesterified into the fatty acid alkyl esters (FAAEs) which comprise biodiesel fuel. While the tranesterification reaction itself is simple, monitoring the reaction progress and reaction products is not. Gas chromatography-mass spectrometry is useful for assessing the FAAE products, but does not directly address either the tri-, di-, or monoglycerides present from incomplete transesterification or the free fatty acids which may also be present. Analysis of the biodiesel reaction mixture is complicated by the solubility and physical property differences among the components of the tranesterification reaction mixture. In this contribution, we present a simple, rapid HPLC method which allows for monitoring all of the main components in a biodiesel fuel transesterification reaction, with specific emphasis on the ability to monitor the reaction as a function of time. The utilization of a relatively new, core-shell stationary phase for the HPLC column allows for efficient separation of peaks with short elution times, saving both time and solvent.

Analysis and Implementation of Accident Tolerant Nuclear Fuels

NASA Astrophysics Data System (ADS)

Prewitt, Benjamin Joseph

To improve the reliability and robustness of LWR, accident tolerant nuclear fuels and cladding materials are being developed to possibly replace the current UO2/zirconium system. This research highlights UN and U3Si 2, two of the most favorable accident tolerant fuels being developed. To evaluate the commercial feasiblilty of these fuels, two areas of research were conducted. Chemical fabrication routes for both fuels were investigated in detail, considering UO2 and UF6 as potential starting materials. Potential pathways for industrial scale fabrication using these methods were discussed. Neutronic performance of 70%UN-30%U3Si2 composite was evaluated in MNCP using PWR assembly and core models. The results showed comparable performance to an identical UO2 fueled simulation with the same configuration. The parameters simulated for composite and oxide fuel include the following: fuel to moderator ratio curves; energy dependent flux spectra; temperature coefficients for fuel and moderator; delayed neutron fractions; power peaking factors; axial and radial flux profiles in 2D and 3D; burnup; critical boron concentration; and shutdown margin. Overall, the neutronic parameters suggest that the transition from UO2 to composite in existing nuclear systems will not require significant changes in operating procedures or modifications to standards and regulations.

NUCLEAR REACTOR CORE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Preece, G.E.; Bell, F.R.; Page, R.W.

1963-03-01

A nuclear reactor core is described. It contains fuel in the form of blocks or pellets that have a grooved, wrinkled, or corrugated surface to provide a greater radiating surface area. The surfaces of spaces in the core are correspondingly corrugated for maximum heat exchange area. (C.E.S.)

Nuclear reactor composite fuel assembly

DOEpatents

Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

1980-01-01

A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

Secondary fuel delivery system

DOEpatents

Parker, David M.; Cai, Weidong; Garan, Daniel W.; Harris, Arthur J.

2010-02-23

A secondary fuel delivery system for delivering a secondary stream of fuel and/or diluent to a secondary combustion zone located in the transition piece of a combustion engine, downstream of the engine primary combustion region is disclosed. The system includes a manifold formed integral to, and surrounding a portion of, the transition piece, a manifold inlet port, and a collection of injection nozzles. A flowsleeve augments fuel/diluent flow velocity and improves the system cooling effectiveness. Passive cooling elements, including effusion cooling holes located within the transition boundary and thermal-stress-dissipating gaps that resist thermal stress accumulation, provide supplemental heat dissipation in key areas. The system delivers a secondary fuel/diluent mixture to a secondary combustion zone located along the length of the transition piece, while reducing the impact of elevated vibration levels found within the transition piece and avoiding the heat dissipation difficulties often associated with traditional vibration reduction methods.

Phase Transition of Poly(acrylic acid-co-N-isopropylacrylamide) Core-shell Nanogels

NASA Astrophysics Data System (ADS)

Liu, Xiao-bing; Zhou, Jian-feng; Ye, Xiao-dong

2012-08-01

A series of poly(acrylic acid) macromolecular chain transfer agents with different molecular weights were synthesized by reversible addition-fragmentation chain transfer (RAFT) polymerization and characterized by 1H NMR and gel permeation chromatography. Multiresponsive core-shell nanogels were prepared by dispersion polymerization of N-isopropylacrylamide in water using these poly(potassium acrylate) macro-RAFT agents as the electrosteric stabilizer. The size of the nanogels decreases with the amount of the macro-RAFT agent, indicating that the surface area occupied by per polyelectrolyte group is a critical parameter for stabilizing the nanogels. The volume phase transition and the zeta potentials of the nanogels in aqueous solutions were studied by dynamic light scattering and zetasizer analyzer, respectively.

Assessing the validity of health impact assessment predictions regarding a Japanese city's transition to core city status: a monitoring review.

PubMed

Hoshiko, M; Hara, K; Ishitake, T

2012-02-01

The validity of health impact assessment (HIA) predictions has not been accurately assessed to date. In recent years, legislative attempts to promote decentralization have been progressing in Japan, and Kurume was designated as a core city in April 2008. An HIA into the transition of Kurume to a core city was conducted before the event, but the recommendations were not accepted by city officials. The aim of this study was to examine the validity of predictions made in the HIA on Kurume by conducting a monitoring review into the accuracy of the predictions. Before Kurume was designated as a core city, the residents completed an online questionnaire and city officials were interviewed. The findings and recommendations were presented to the city administration. One year after the transition, a monitoring review was performed to clarify the accuracy of the HIA predictions by evaluating the correlation between the predictions and reality. Many of the HIA predictions were found to conflict with reality in Kurume. Prediction validity was evaluated for two groups: residents of Kurume and city officials. For the residents, 17% (2/12 items) of the predictions were found to be compatible, 58% (7/12) were incompatible and 25% (3/12) were difficult to evaluate. For city officials, the analysis was divided into those whose department was directly involved in tasks transferred to them (transfer tasks) and those whose department was not. For the city officials in departments responsible for conducting core city transfer tasks, 33% (3/9 items) of the predictions were found to be compatible, 33% (3/9) were incompatible and 33% (3/9) were difficult to evaluate. However, for the city officials whose responsibilities were unrelated to core city transfer tasks, 11% (1/9) of predictions were found to be compatible, 78% (7/9) were incompatible and 11% (1/9) were difficult to evaluate. Although it was possible to validate some of the HIA predictions, the results of this monitoring review

Revised model core potentials for third-row transition-metal atoms from Lu to Hg

NASA Astrophysics Data System (ADS)

Mori, Hirotoshi; Ueno-Noto, Kaori; Osanai, You; Noro, Takeshi; Fujiwara, Takayuki; Klobukowski, Mariusz; Miyoshi, Eisaku

2009-07-01

We have produced new relativistic model core potentials (spdsMCPs) for the third-row transition-metal atoms from Lu to Hg explicitly treating explicitly 5s and 5p electrons in addition to 5d and 6s electrons in the same manner for the first- and second-row transition-metal atoms given in the previous Letters [Y. Osanai, M.S. Mon, T. Noro, H. Mori, H. Nakashima, M. Klobukowski, E. Miyoshi, Chem. Phys. Lett. 452 (2008) 210; Y. Osanai, E. Soejima, T. Noro, H. Mori, M.S. Mon, M. Klobukowski, E. Miyoshi, Chem. Phys. Lett. 463 (2008) 230]. Using suitable correlating functions with the split-valence MCP functions, we demonstrate that the present MCP basis sets show reasonable performance in describing the electronic structures of atoms and molecules, bringing about accurate excitation energies for atoms and proper spectroscopic constants for Au 2, Hg 2, and AuH.

The Transition to a Many-core World

NASA Astrophysics Data System (ADS)

Mattson, T. G.

2012-12-01

The need to increase performance within a fixed energy budget has pushed the computer industry to many core processors. This is grounded in the physics of computing and is not a trend that will just go away. It is hard to overestimate the profound impact of many-core processors on software developers. Virtually every facet of the software development process will need to change to adapt to these new processors. In this talk, we will look at many-core hardware and consider its evolution from a perspective grounded in the CPU. We will show that the number of cores will inevitably increase, but in addition, a quest to maximize performance per watt will push these cores to be heterogeneous. We will show that the inevitable result of these changes is a computing landscape where the distinction between the CPU and the GPU is blurred. We will then consider the much more pressing problem of software in a many core world. Writing software for heterogeneous many core processors is well beyond the ability of current programmers. One solution is to support a software development process where programmer teams are split into two distinct groups: a large group of domain-expert productivity programmers and much smaller team of computer-scientist efficiency programmers. The productivity programmers work in terms of high level frameworks to express the concurrency in their problems while avoiding any details for how that concurrency is exploited. The second group, the efficiency programmers, map applications expressed in terms of these frameworks onto the target many-core system. In other words, we can solve the many-core software problem by creating a software infrastructure that only requires a small subset of programmers to become master parallel programmers. This is different from the discredited dream of automatic parallelism. Note that productivity programmers still need to define the architecture of their software in a way that exposes the concurrency inherent in their

ATR LEU fuel and burnable absorber neutronics performance optimization by fuel meat thickness variation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang, G.S.

2008-07-15

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U-235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core th and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.381 mm and the same U-235 enrichment (19.7 wt%) can be used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.5 mil) to 0.343 mm (13.5 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, 0.8g of a burnable absorber, Boron-10, was added in the inner and outer plates to reduce the initial excess

ATR LEU Fuel and Burnable Absorber Neutronics Performance Optimization by Fuel Meat Thickness Variation

DOE Office of Scientific and Technical Information (OSTI.GOV)

G. S. Chang

2007-09-01

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.508 mm and the same U-235 enrichment (15.5 wt%) can be used to optimize the radial heat flux profile by varying the fuel plate thickness from 0.254 to 0.457 mm at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, a 0.7g of burnable absorber Boron-10 was added in the inner and outer plates to reduce the initial excess reactivity, and the inner

Emergency core cooling system

DOEpatents

Schenewerk, William E.; Glasgow, Lyle E.

1983-01-01

A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.

Emergency deployable core catcher

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosewell, M.P.

An emergency melt down core catcher apparatus for a nuclear reactor having a retrofitable eutectic solute holding vessel connected to a core containment vessel with particle transferring fluid and particles or granules of solid eutectic solute materials contained therein and transferable by automatically operated valve means to transport and position the solid eutectic solute material in a position below the core to catch and react with any partial or complete melt down of the fuel core.

HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; Barbara H. Dolphin; James W. Sterbentz

2013-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters.more » Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.« less

HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; Barbara H. Dolphin; James W. Sterbentz

2012-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters.more » Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.« less

NUCLEAR REACTOR CORE DESIGN

DOEpatents

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

1960-03-22

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

Entropy of vortex cores near the superconductor-insulator transition in an underdoped cuprate.

PubMed

Capan, C; Behnia, K; Hinderer, J; Jansen, A G M; Lang, W; Marcenat, C; Marin, C; Flouquet, J

2002-02-04

We present a study of Nernst effect in underdoped La(2-x)Sr(x)CuO4 in magnetic fields as high as 28 T. At high fields, a sizable Nernst signal was found to persist in the presence of a field-induced nonmetallic resistivity. By simultaneously measuring resistivity and the Nernst coefficient, we extract the entropy of vortex cores in the vicinity of this field-induced superconductor-insulator transition. Moreover, the temperature dependence of the thermoelectric Hall angle provides strong constraints on the possible origins of the finite Nernst signal above T(c), as recently discovered by Xu et al. [Nature (London) 406, 486 (2000)].

Nodal weighting factor method for ex-core fast neutron fluence evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chiang, R. T.

The nodal weighting factor method is developed for evaluating ex-core fast neutron flux in a nuclear reactor by utilizing adjoint neutron flux, a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV, the unit fission source, and relative assembly nodal powers. The method determines each nodal weighting factor for ex-core neutron fast flux evaluation by solving the steady-state adjoint neutron transport equation with a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV as the adjoint source, by integrating the unit fission source with a typical fission spectrum to the solved adjointmore » flux over all energies, all angles and given nodal volume, and by dividing it with the sum of all nodal weighting factors, which is a normalization factor. Then, the fast neutron flux can be obtained by summing the various relative nodal powers times the corresponding nodal weighting factors of the adjacent significantly contributed peripheral assembly nodes and times a proper fast neutron attenuation coefficient over an operating period. A generic set of nodal weighting factors can be used to evaluate neutron fluence at the same location for similar core design and fuel cycles, but the set of nodal weighting factors needs to be re-calibrated for a transition-fuel-cycle. This newly developed nodal weighting factor method should be a useful and simplified tool for evaluating fast neutron fluence at selected locations of interest in ex-core components of contemporary nuclear power reactors. (authors)« less

NUCLEAR REACTOR CORE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bell, F.R.

1963-02-01

A nuclear reactor core composed of a number of identical elements of solid moderator material fitted together was designed. Each moderator element is apertured to provide channels for fuel and coolant. The elements have an external shape which permits them to be stacked in layers with similar elements, with the surfaces of adjacent elements fitting and in contact with each other. The cross section of the element is of a general hexagonal shape with identations and protrusions, so that the elements can be fitted together. The described core should not be liable to fracture under transverse loading. Specific arrangements ofmore » moderator elements and fuel and coolant apertures are described. (M.P.G.)« less

REACTOR FUEL SCAVENGING MEANS

DOEpatents

Coffinberry, A.S.

1962-04-10

A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

Core-core and core-valence correlation

NASA Technical Reports Server (NTRS)

Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.

1988-01-01

The effect of 1s core correlation on properties and energy separations are analyzed using full configuration-interaction (FCI) calculations. The Be1S - 1P, the C 3P - 5S,m and CH(+) 1Sigma(+) - 1Pi separations, and CH(+) spectroscopic constants, dipole moment, and 1Sigma(+) - 1Pi transition dipole moment have been studied. The results of the FCI calculations are compared to those obtained using approximate methods.

Nondestrucive analysis of fuel pins

DOEpatents

Stepan, I.E.; Allard, N.P.; Suter, C.R.

1972-11-03

Disclosure is made of a method and a correspondingly adapted facility for the nondestructive analysis of the concentation of fuel and poison in a nuclear reactor fuel pin. The concentrations of fuel and poison in successive sections along the entire length of the fuel pin are determined by measuring the reactivity of a thermal reactor as each successive small section of the fuel pin is exposed to the neutron flux of the reactor core and comparing the measured reactivity with the reactivities measured for standard fuel pins having various known concentrations. Only a small section of the length of the fuel pin is exposed to the neutron flux at any one time while the remainder of the fuel pin is shielded from the neutron flux. In order to expose only a small section at any one time, a boron-10-lined dry traverse tube is passed through the test region within the core of a low-power thermal nuclear reactor which has a very high fuel sensitivity. A narrow window in the boron-10 lining is positioned at the core center line. The fuel pins are then systematically traversed through the tube past the narrow window such that successive small sections along the length of the fuel pin are exposed to the neutron flux which passes through the narrow window.

Thermal breeder fuel enrichment zoning

DOEpatents

Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

1992-01-01

A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

"Better to Be a Pessimist": A Narrative Inquiry into Mathematics Teachers' Experience of the Transition to the Common Core

ERIC Educational Resources Information Center

Martinie, Sherri L.; Kim, Jeong-Hee; Abernathy, Deborah

2016-01-01

The Common Core State Standards (CCSS) are a focus of state education policy today influencing curriculum implementation and assessment in public schools. The purpose of this narrative inquiry is to understand how high school mathematics teachers experience the transition period. Based on interviews with mathematics teachers in a high school in…

Burnable absorber arrangement for fuel bundle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crowther, R.L.; Townsend, D.B.

1986-12-16

This patent describes a boiling water reactor core whose operation is characterized by a substantial proportion of steam voids with concomitantly reduced moderation toward the top of the core when the reactor is in its hot operating condition. The reduced moderation leads to slower burnup and greater conversion ratio in an upper core region so that when the reactor is in its cold shut down condition the resulting relatively increased moderation in the upper core region is accompanied by a reactivity profile that peaks in the upper core region. A fuel assembly is described comprising; a component of fissile materialmore » distributed over a substantial axial extent of the fuel assembly; and a component of neutron absorbing material having an axial distribution characterized by an enhancement in an axial zone of the fuel assembly, designated the cold shutdown control zone, corresponding to at least a portion of the axial region of the core when the cold shutdown reactivity peaks. The aggregate amount of neutron absorbing material in the cold shutdown zone of the fuel assembly is greater than the aggregate amount of neutron absorbing material in the axial zones of the fuel assembly immediately above and immediately below the cold shutdown control zone whereby the cold shutdown reactivity peak is reduced relative to the cold shutdown reactivity in the zones immediately above and immediately below the cold shutdown control zone. The cold shutdown zone has an axial extent measured from the bottom of the fuel assembly in the range between 68-88 percent of the height of the fissile material in the fuel assembly.« less

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Dickson, J.J.

1963-09-24

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

Magma-Hydrothermal Transition: Basalt Alteration at Supercritical Conditions in Drill Core from Reykjanes, Iceland, Iceland Deep Drilling Project.

NASA Astrophysics Data System (ADS)

Zierenberg, R. A.; Fowler, A. P.; Schiffman, P.; Fridleifsson, G. Ó.; Elders, W. A.

2017-12-01

The Iceland Deep Drilling Project well IDDP-2, drilled to 4,659 m in the Reykjanes geothermal system, the on-land extension of the Mid Atlantic Ridge, SW Iceland. Drill core was recovered, for the first time, from a seawater-recharged, basalt-hosted hydrothermal system at supercritical conditions. The well has not yet been allowed to heat to in situ conditions, but temperature and pressure of 426º C and 340 bar was measured at 4500 m depth prior to the final coring runs. Spot drill cores were recovered between drilling depths of 3648.00 m and 4657.58 m. Analysis of the core is on-going, but we present the following initial observations. The cored material comes from a basaltic sheeted dike complex in the brittle-ductile transition zone. Felsic (plagiogranite) segregation veins are present in minor amounts in dikes recovered below 4300 m. Most core is pervasively altered to hornblende + plagioclase, but shows only minor changes in major and minor element composition. The deepest samples record the transition from the magmatic regime to the presently active hydrothermal system. Diabase near dike margins has been locally recrystallized to granoblastic-textured orthopyroxene-clinopyroxe-plagioclase hornfels. High temperature hydrothermal alteration includes calcic plagioclase (up to An100) and aluminous hornblende (up to 11 Wt. % Al2O3) locally intergrown with hydrothermal biotite, clinopyroxene, orthopyroxene and/or olivine. Hydrothermal olivine is iron-rich (Mg # 59-64) compared to expected values for igneous olivine. Biotite phenocrysts in felsic segregation veins have higher Cl and Fe compared to hydrothermal biotites. Orthopyroxene-clinopyroxene pairs in partially altered quench dike margins give temperature of 955° to 1067° C. Orthopyroxene-clinopyroxene pairs from hornfels and hydrothermal veins and replacements give temperature ranging from 774° to 888° C. Downhole fluid sampling is planned following thermal equilibration of the drill hole. Previous work

Apparatus for and method of monitoring for breached fuel elements

DOEpatents

Gross, Kenny C.; Strain, Robert V.

1983-01-01

This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus uses a separate bypass loop for conveying part of the reactor coolant away from the core, and at least three separate delayed-neutron detectors mounted proximate this detector loop. The detectors are spaced apart so that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the dealy time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector. At least two of these time components are determined during calibrated operation of the reactor. Thereafter during normal reactor operation, repeated comparisons are made by the method of regression approximation of the third time component for the best-fit line correlating measured delayed-neutron activity against activity that is approximated according to specific equations. The equations use these time-delay components and known parameter values of the fuel and of the part and emitting daughter isotopes.

Texas Hydrogen Highway Fuel Cell Hybrid Bus and Fueling Infrastructure Technology Showcase - Final Scientific/Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hitchcock, David

The Texas Hydrogen Highway project has showcased a hydrogen fuel cell transit bus and hydrogen fueling infrastructure that was designed and built through previous support from various public and private sector entities. The aim of this project has been to increase awareness among transit agencies and other public entities on these transportation technologies, and to place such technologies into commercial applications, such as a public transit agency. The initial project concept developed in 2004 was to show that a skid-mounted, fully-integrated, factory-built and tested hydrogen fueling station could be used to simplify the design, and lower the cost of fuelingmore » infrastructure for fuel cell vehicles. The approach was to design, engineer, build, and test the integrated fueling station at the factory then install it at a site that offered educational and technical resources and provide an opportunity to showcase both the fueling station and advanced hydrogen vehicles. The two primary technology components include: Hydrogen Fueling Station: The hydrogen fueling infrastructure was designed and built by Gas Technology Institute primarily through a funding grant from the Texas Commission on Environmental Quality. It includes hydrogen production, clean-up, compression, storage, and dispensing. The station consists of a steam methane reformer, gas clean-up system, gas compressor and 48 kilograms of hydrogen storage capacity for dispensing at 5000 psig. The station is skid-mounted for easy installation and can be relocated if needed. It includes a dispenser that is designed to provide temperaturecompensated fills using a control algorithm. The total station daily capacity is approximately 50 kilograms. Fuel Cell Bus: The transit passenger bus built by Ebus, a company located in Downey, CA, was commissioned and acquired by GTI prior to this project. It is a fuel cell plug-in hybrid electric vehicle which is ADA compliant, has air conditioning sufficient for Texas

Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru; Pinegin, A. A.

The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

Vibrational frequencies of transition metal chloride and oxo compounds using effective core potential analytic second derivatives

NASA Astrophysics Data System (ADS)

Russo, Thomas V.; Martin, Richard L.; Hay, P. Jeffrey; Rappé, Anthony K.

1995-06-01

The application of analytic second derivative techniques to quantum chemical calculations using effective core potentials is discussed. Using a recent implementation of these techniques, the vibrational frequencies of transition metal compounds are calculated including the chlorides TiCl4, ZrCl4, and HfCl4, the oxochlorides CrO2Cl2, MoO2Cl2, WO2Cl2, and VOCl3, and the oxide OsO4. Results are compared to previous calculations and with experimental results.

Climate and Fuel Controls on North American Paleofires: Smoldering to Flaming in the Late-Glacial-Holocene Transition

NASA Technical Reports Server (NTRS)

Han, Y. M.; Peteet, D. M.; Arimoto, R.; Cao, J. J.; An, Z. S.; Sritrairat, S.; Yan, B. Z.

2016-01-01

Smoldering and flaming fires, which emit different proportions of organic (OC) and black carbon (BC, in the form of char and soot), have long been recognized in modern wildfire observations but never in a paleo-record, and little is known about their interactions with climate. Here we show that in the late glacial-early Holocene transition period, when the climate was moist, relatively high quantities of char were deposited in Linsley Pond, Connecticut, USA while soot was more abundant during the warmer and drier early Holocene interval. The highest soot mass accumulation rates (MARs) occurred at the beginning of the Holocene as fuel availability increased through the climatic transition when boreal forests were locally extirpated. These variations with time are related to the different formation pathways of char and soot, which are governed by combustion efficiency. This study provides an approach for differentiating smoldering from flaming combustion in paleo-wildfire reconstructions. Our results suggest that climate and fuel loads control the occurrence of different wildfire types and precipitation may play a key role.

Climate and Fuel Controls on North American Paleofires: Smoldering to Flaming in the Late-glacial-Holocene Transition

NASA Astrophysics Data System (ADS)

Han, Y. M.; Peteet, D. M.; Arimoto, R.; Cao, J. J.; An, Z. S.; Sritrairat, S.; Yan, B. Z.

2016-02-01

Smoldering and flaming fires, which emit different proportions of organic (OC) and black carbon (BC, in the form of char and soot), have long been recognized in modern wildfire observations but never in a paleo-record, and little is known about their interactions with climate. Here we show that in the late glacial-early Holocene transition period, when the climate was moist, relatively high quantities of char were deposited in Linsley Pond, Connecticut, USA while soot was more abundant during the warmer and drier early Holocene interval. The highest soot mass accumulation rates (MARs) occurred at the beginning of the Holocene as fuel availability increased through the climatic transition when boreal forests were locally extirpated. These variations with time are related to the different formation pathways of char and soot, which are governed by combustion efficiency. This study provides an approach for differentiating smoldering from flaming combustion in paleo-wildfire reconstructions. Our results suggest that climate and fuel loads control the occurrence of different wildfire types and precipitation may play a key role.

Climate and Fuel Controls on North American Paleofires: Smoldering to Flaming in the Late-glacial-Holocene Transition.

PubMed

Han, Y M; Peteet, D M; Arimoto, R; Cao, J J; An, Z S; Sritrairat, S; Yan, B Z

2016-02-10

Smoldering and flaming fires, which emit different proportions of organic (OC) and black carbon (BC, in the form of char and soot), have long been recognized in modern wildfire observations but never in a paleo-record, and little is known about their interactions with climate. Here we show that in the late glacial-early Holocene transition period, when the climate was moist, relatively high quantities of char were deposited in Linsley Pond, Connecticut, USA while soot was more abundant during the warmer and drier early Holocene interval. The highest soot mass accumulation rates (MARs) occurred at the beginning of the Holocene as fuel availability increased through the climatic transition when boreal forests were locally extirpated. These variations with time are related to the different formation pathways of char and soot, which are governed by combustion efficiency. This study provides an approach for differentiating smoldering from flaming combustion in paleo-wildfire reconstructions. Our results suggest that climate and fuel loads control the occurrence of different wildfire types and precipitation may play a key role.

Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petrovic, Bojan; Maldonado, Ivan

2016-04-14

The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed onmore » December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.« less

Survey evidence on the importance of fuel availability to choice of alternative fuels and vehicles

DOT National Transportation Integrated Search

1997-11-11

The effect of limited fuel availability on the demand for alternative fuels and : vehicles is a critical factor in the transition to alternative fuels. Because : petroleum fuels have been so dominant for so long, the relationship between fuel : avail...

NEUTRONIC REACTOR CORE

DOEpatents

Thomson, W.B.; Corbin, A. Jr.

1961-07-18

An improved core for a gas-cooled power reactor which admits gas coolant at high temperatures while affording strong integral supporting structure and efficient moderation of neutrons is described. The multiplicities of fuel elements constituting the critical amassment of fissionable material are supported and confined by a matrix of metallic structure which is interspersed therebetween. Thermal insulation is interposed between substantially all of the metallic matrix and the fuel elements; the insulation then defines the principal conduit system for conducting the coolant gas in heat-transfer relationship with the fuel elements. The metallic matrix itseif comprises a system of ducts through which an externally-cooled hydrogeneous liquid, such as water, is circulated to serve as the principal neutron moderant for the core and conjointly as the principal coolant for the insulated metallic structure. In this way, use of substantially neutron transparent metals, such as aluminum, becomes possible for the supporting structure, despite the high temperatures of the proximate gas. The Aircraft Nuclear Propulsion program's "R-1" reactor design is a preferred embodiment.

MODULAR CORE UNITS FOR A NEUTRONIC REACTOR

DOEpatents

Gage, J.F. Jr.; Sherer, D.B.

1964-04-01

A modular core unit for use in a nuclear reactor is described. Many identical core modules can be placed next to each other to make up a complete core. Such a module includes a cylinder of moderator material surrounding a fuel- containing re-entrant coolant channel. The re-entrant channel provides for the circulation of coolant such as liquid sodium from one end of the core unit, through the fuel region, and back out through the same end as it entered. Thermal insulation surrounds the moderator exterior wall inducing heat to travel inwardly to the coolant channel. Spaces between units may be used to accommodate control rods and support structure, which may be cooled by a secondary gas coolant, independently of the main coolant. (AEC)

An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

NASA Astrophysics Data System (ADS)

Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

2013-10-01

In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

FUEL ELEMENTS FOR NUCLEAR REACTORS

DOEpatents

Blainey, A.; Lloyd, H.

1961-07-11

A method of sheathing a tubular fuel element for a nuclear reactor is described. A low melting metal core member is centered in a die, a layer of a powdered sheathing substance is placed on the bottom of the die, the tubular fuel element is inserted in the die, the space between the tubular fuel element and the die walls and core member is filled with the same powdered sheathing substance, a layer of the same substance is placed over the fissile material, and the charge within the die is subjected to pressure in the direction of the axis of the fuel element at the sintering temperature of the protective substance.

Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Worrall, Andrew; Todosow, Michael

2016-01-01

Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include:more » increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle

Monte Carlo simulation of dynamic phase transitions and frequency dispersions of hysteresis curves in core/shell ferrimagnetic cubic nanoparticle

NASA Astrophysics Data System (ADS)

Vatansever, Erol

2017-05-01

By means of Monte Carlo simulation method with Metropolis algorithm, we elucidate the thermal and magnetic phase transition behaviors of a ferrimagnetic core/shell nanocubic system driven by a time dependent magnetic field. The particle core is composed of ferromagnetic spins, and it is surrounded by an antiferromagnetic shell. At the interface of the core/shell particle, we use antiferromagnetic spin-spin coupling. We simulate the nanoparticle using classical Heisenberg spins. After a detailed analysis, our Monte Carlo simulation results suggest that present system exhibits unusual and interesting magnetic behaviors. For example, at the relatively lower temperature regions, an increment in the amplitude of the external field destroys the antiferromagnetism in the shell part of the nanoparticle, leading to a ground state with ferromagnetic character. Moreover, particular attention has been dedicated to the hysteresis behaviors of the system. For the first time, we show that frequency dispersions can be categorized into three groups for a fixed temperature for finite core/shell systems, as in the case of the conventional bulk systems under the influence of an oscillating magnetic field.

PROCESS FOR JACKETING A CORE

DOEpatents

Last, G.A.

1960-07-19

A process is given for enclosing the uranium core of a nuclear fuel element by placing the core in an aluminum cup and closing the open end of the cup over the core. As the metal of the cup is brought together in a weld over the center of the end of the core, it is extruded inwardly as internal projection into a central recess in the core and outwardly as an external projection. Thus oxide inclusions in the weld of the cup are spread out into the internal and external projections and do not interfere with the integrity of the weld.

Gram-level synthesis of core-shell structured catalysts for the oxygen reduction reaction in proton exchange membrane fuel cells

NASA Astrophysics Data System (ADS)

Luo, Mingchuan; Wei, Lingli; Wang, Fanghui; Han, Kefei; Zhu, Hong

2014-12-01

Over the past decade, Pt based core-shell structured alloys have been studied extensively as oxygen reduction reaction (ORR) catalysts for proton exchange membrane fuel cells (PEMFCs) because of their distinctive electrochemical performance and low Pt loading. In this paper, a facile route based on microwave-assisted polyol method and chemical dealloying process is proposed to synthesize carbon supported core-shell structured nanoparticles (NPs) in gram-level for ORR electrocatalysis in PEMFCs. The obtained samples are characterized by X-ray diffraction (XRD), transmission electron microscopy (TEM), energy dispersive X-ray spectroscopy (EDX), inductively coupled plasma atomic emission spectroscopy (ICP-AES), and X-ray photoelectron spectroscopy (XPS). These physical characterization indicate that the final synthesized NPs are highly dispersed on the carbon support, and in a core-shell structure with CuPt alloy as the core and Pt as the shell. Electrochemical measurements, conducted by cyclic voltammetry (CV) and rotating disk electrode (RDE) tests, show the core-shell structured catalyst exhibit a 3× increase in mass activity and a 2× increase in specific activity over the commercial Pt/C catalyst, respectively. These results demonstrate that this route can be a reliable way to synthesize low-Pt catalyst in large-scale for PEMFCs.

Nuclear fuel elements having a composite cladding

DOEpatents

Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

1983-09-20

An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

POWER GENERATION FROM LIQUID METAL NUCLEAR FUEL

DOEpatents

Dwyer, O.E.

1958-12-23

A nuclear reactor system is described wherein the reactor is the type using a liquid metal fuel, such as a dispersion of fissile material in bismuth. The reactor is designed ln the form of a closed loop having a core sectlon and heat exchanger sections. The liquid fuel is clrculated through the loop undergoing flssion in the core section to produce heat energy and transferrlng this heat energy to secondary fluids in the heat exchanger sections. The fission in the core may be produced by a separate neutron source or by a selfsustained chain reaction of the liquid fuel present in the core section. Additional auxiliary heat exchangers are used in the system to convert water into steam which drives a turbine.

Compressed natural gas fuel use training manual

DOT National Transportation Integrated Search

1992-09-01

The Urban Mass Transportation Administration (UMTA) Alternative Fuel Initiative and the Environmental Protection Agency (EPA) 1991 regulations on transit bus exhaust emissions has resulted in a number of alternative fueled transit bus demonstrations....

RIT Stability through the Transition to Common Core-Aligned MAP® Tests. How Using MAP to Measure Student Learning Growth is Reliable Now and in 2014

ERIC Educational Resources Information Center

Northwest Evaluation Association, 2013

2013-01-01

While many educators expect the Common Core State Standards (CCSS) to be more rigorous than previous state standards, some wonder if the transition to CCSS and to a Common Core aligned MAP test will have an impact on their students' RIT scores or the NWEA norms. MAP assessments use a proprietary scale known as the RIT (Rasch unit) scale to measure…

Optimization of burnable poison design for Pu incineration in fully fertile free PWR core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fridman, E.; Shwageraus, E.; Galperin, A.

2006-07-01

The design challenges of the fertile-free based fuel (FFF) can be addressed by careful and elaborate use of burnable poisons (BP). Practical fully FFF core design for PWR reactor has been reported in the past [1]. However, the burnable poison option used in the design resulted in significant end of cycle reactivity penalty due to incomplete BP depletion. Consequently, excessive Pu loading were required to maintain the target fuel cycle length, which in turn decreased the Pu burning efficiency. A systematic evaluation of commercially available BP materials in all configurations currently used in PWRs is the main objective of thismore » work. The BP materials considered are Boron, Gd, Er, and Hf. The BP geometries were based on Wet Annular Burnable Absorber (WABA), Integral Fuel Burnable Absorber (IFBA), and Homogeneous poison/fuel mixtures. Several most promising combinations of BP designs were selected for the full core 3D simulation. All major core performance parameters for the analyzed cases are very close to those of a standard PWR with conventional UO{sub 2} fuel including possibility of reactivity control, power peaking factors, and cycle length. The MTC of all FFF cores was found at the full power conditions at all times and very close to that of the UO{sub 2} core. The Doppler coefficient of the FFF cores is also negative but somewhat lower in magnitude compared to UO{sub 2} core. The soluble boron worth of the FFF cores was calculated to be lower than that of the UO{sub 2} core by about a factor of two, which still allows the core reactivity control with acceptable soluble boron concentrations. The main conclusion of this work is that judicial application of burnable poisons for fertile free fuel has a potential to produce a core design with performance characteristics close to those of the reference PWR core with conventional UO{sub 2} fuel. (authors)« less

Development of a three-dimensional core dynamics analysis program for commercial boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bessho, Yasunori; Yokomizo, Osamu; Yoshimoto, Yuichiro

1997-03-01

Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates calculational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory size as attained by the latest supercomputer technology. The program consists of six major modules, namely a core neutronics module, a fuel heat conduction/transfer module, a fuel channel thermal-hydraulic module, an upper plenum/separator module, a feedwater/recirculation flow module, and amore » control system module. Its core neutronics module is based on the modified one-group neutron kinetics equation with the prompt jump approximation and with six delayed neutron precursor groups. The module is used to analyze one fuel bundle of the reactor core with one mesh (region). The fuel heat conduction/transfer module solves the one-dimensional heat conduction equation in the radial direction with ten nodes in the fuel pin. The fuel channel thermal-hydraulic module is based on separated three-equation, two-phase flow equations with the drift flux correlation, and it analyzes one fuel bundle of the reactor core with one channel to evaluate flow redistribution between channels precisely. Thermal margin is evaluated by using the GEXL correlation, for example, in the module.« less

Evaluation of HFIR LEU Fuel Using the COMSOL Multiphysics Platform

DOE Office of Scientific and Technical Information (OSTI.GOV)

Primm, Trent; Ruggles, Arthur; Freels, James D

2009-03-01

A finite element computational approach to simulation of the High Flux Isotope Reactor (HFIR) Core Thermal-Fluid behavior is developed. These models were developed to facilitate design of a low enriched core for the HFIR, which will have different axial and radial flux profiles from the current HEU core and thus will require fuel and poison load optimization. This report outlines a stepwise implementation of this modeling approach using the commercial finite element code, COMSOL, with initial assessment of fuel, poison and clad conduction modeling capability, followed by assessment of mating of the fuel conduction models to a one dimensional fluidmore » model typical of legacy simulation techniques for the HFIR core. The model is then extended to fully couple 2-dimensional conduction in the fuel to a 2-dimensional thermo-fluid model of the coolant for a HFIR core cooling sub-channel with additional assessment of simulation outcomes. Finally, 3-dimensional simulations of a fuel plate and cooling channel are presented.« less

A neutral branched platinum-acetylide complex possessing a tetraphenylethylene core: preparation of a luminescent organometallic gelator and its unexpected spectroscopic behaviour during sol-to-gel transition.

PubMed

Ren, Yuan-Yuan; Wu, Nai-Wei; Huang, Junhai; Xu, Zheng; Sun, Dan-Dan; Wang, Cui-Hong; Xu, Lin

2015-10-21

A neutral branched platinum-acetylide complex TPA possessing a tetraphenylethylene core was successfully prepared, which was found to form luminescent organometallic gels in ethyl acetate. Stimulated by temperature or F(-), the reversible gel-sol transition was realized. More interestingly, TPA exhibited an unexpected blue shift of the emission during the sol-to-gel transition.

NEUTRONIC REACTOR OPERATIONAL METHOD AND CORE SYSTEM

DOEpatents

Winters, C.E.; Graham, C.B.; Culver, J.S.; Wilson, R.H.

1960-07-19

Homogeneous neutronic reactor systems are described wherein an aqueous fuel solution is continuously circulated through a spherical core tank. The pumped fuel solution-is injected tangentially into the hollow spherical interior, thereby maintaining vigorous rotation of the solution within the tank in the form of a vortex; gaseous radiolytic decomposition products concentrate within the axial vortex cavity. The evolved gas is continuously discharged through a gas- outlet port registering with an extremity of the vortex cavity. and the solution stream is discharged through an annular liquid outlet port concentrically encircling the gas outlet by virtue of which the vortex and its cavity are maintained precisely axially aligned with the gas outlet. A primary heat exchanger extracts useful heat from the hot effluent fuel solution before its recirculation into the core tank. Hollow cylinders and other alternative core- tank configurations defining geometric volumes of revolution about a principal axis are also covered. AEC's Homogeneous Reactor Experiment No. 1 is a preferred embodiment.

Energy Efficient Engine core design and performance report

NASA Technical Reports Server (NTRS)

Stearns, E. Marshall

1982-01-01

The Energy Efficient Engine (E3) is a NASA program to develop fuel saving technology for future large transport aircraft engines. Testing of the General Electric E3 core showed that the core component performance and core system performance necessary to meet the program goals can be achieved. The E3 core design and test results are described.

Magnetic transition and sound velocities of Fe 3S at high pressure: implications for Earth and planetary cores

NASA Astrophysics Data System (ADS)

Lin, Jung-Fu; Fei, Yingwei; Sturhahn, Wolfgang; Zhao, Jiyong; Mao, Ho-kwang; Hemley, Russell J.

2004-09-01

Magnetic, elastic, thermodynamic, and vibrational properties of the most iron-rich sulfide, Fe3S, known to date have been studied with synchrotron Mössbauer spectroscopy (SMS) and nuclear resonant inelastic X-ray scattering (NRIXS) up to 57 GPa at room temperature. The magnetic hyperfine fields derived from the time spectra of the synchrotron Mössbauer spectroscopy show that the low-pressure magnetic phase displays two magnetic hyperfine field sites and that a magnetic collapse occurs at 21 GPa. The magnetic to non-magnetic transition significantly affects the elastic, thermodynamic, and vibrational properties of Fe3S. The magnetic collapse of Fe3S may also affect the phase relations in the iron-sulfur system, changing the solubility of sulfur in iron under higher pressures. Determination of the physical properties of the non-magnetic Fe3S phase is important for the interpretation of the amount and properties of sulfur present in the planetary cores. Sound velocities of Fe3S obtained from the measured partial phonon density of states (PDOS) for 57Fe incorporated in the alloy show that Fe3S has higher compressional and shear wave velocity than those of hcp-Fe and hcp-Fe0.92Ni0.08 alloy under high pressures, making sulfur a potential light element in the Earth's core based on geophysical arguments. The VP and VS of the non-magnetic Fe3S follow a Birch's law trend whereas the slopes decrease in the magnetic phase, indicating that the decrease of the magnetic moment significantly affects the sound velocities. If the Martian core is in the solid state containing 14.2 wt.% sulfur, it is likely that the non-magnetic Fe3S phase is a dominant component and that our measured sound velocities of Fe3S can be used to construct the corresponding velocity profile of the Martian core. It is also conceivable that Fe3P and Fe3C undergo similar magnetic phase transitions under high pressures.

A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

NASA Technical Reports Server (NTRS)

Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

1996-01-01

Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

A high converter concept for fuel management with blanket fuel assemblies in boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martinez-Frances, N.; Timm, W.; Rossbach, D.

2012-07-01

Studies on the natural Uranium saving and waste reduction potential of a multiple-plant BWR system were performed. The BWR High Converter system should enable a multiple recycling of MOX fuel in current BWR plants by introducing blanket fuel assemblies and burning Uranium and MOX fuel separately. The feasibility of Uranium cores with blankets and full-MOX cores with Plutonium qualities as low as 40% were studied. The power concentration due to blanket insertion is manageable with modern fuel and acceptable values for the thermal limits and reactivity coefficients were obtained. While challenges remain, full-MOX cores also complied with the main designmore » criteria. The combination of Uranium and Plutonium burners in appropriate proportions could enable obtaining as much as 40% more energy out of Uranium ore. Moreover, a proper adjustment of blanket average stay and Plutonium qualities could lead to a system with nearly no Plutonium left for final disposal. The achievement of such goals with current light water technology makes the BWR HC concept an attractive option to improve the fuel cycle until Gen-IV designs are mature. (authors)« less

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1980-04-29

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

DTIC Science & Technology

2013-06-01

Densities ............................................................................................................ 21 2.3 Fuel Mass (Core Total...70 7.1 Geometry, Material Density, and Mass Summary for All Cores...21 Table 3: Fuel Rod Masses for Different Clads

Corrosion resistant PEM fuel cell

DOEpatents

Li, Yang; Meng, Wen-Jin; Swathirajan, Swathy; Harris, Stephen Joel; Doll, Gary Lynn

2001-07-17

The present invention contemplates a PEM fuel cell having electrical contact elements (including bipolar plates/septums) comprising a titanium nitride coated light weight metal (e.g., Al or Ti) core, having a passivating, protective metal layer intermediate the core and the titanium nitride. The protective layer forms a barrier to further oxidation/corrosion when exposed to the fuel cell's operating environment. Stainless steels rich in CR, Ni, and Mo are particularly effective protective interlayers.

Corrosion resistant PEM fuel cell

DOEpatents

Li, Yang; Meng, Wen-Jin; Swathirajan, Swathy; Harris, Stephen Joel; Doll, Gary Lynn

2002-01-01

The present invention contemplates a PEM fuel cell having electrical contact elements (including bipolar plates/septums) comprising a titanium nitride coated light weight metal (e.g., Al or Ti) core, having a passivating, protective metal layer intermediate the core and the titanium nitride. The protective layer forms a barrier to further oxidation/corrosion when exposed to the fuel cell's operating environment. Stainless steels rich in CR, Ni, and Mo are particularly effective protective interlayers.

Corrosion resistant PEM fuel cell

DOEpatents

Li, Yang; Meng, Wen-Jin; Swathirajan, Swathy; Harris, Stephen J.; Doll, Gary L.

1997-01-01

The present invention contemplates a PEM fuel cell having electrical contact elements (including bipolar plates/septums) comprising a titanium nitride coated light weight metal (e.g., Al or Ti) core, having a passivating, protective metal layer intermediate the core and the titanium nitride. The protective layer forms a barrier to further oxidation/corrosion when exposed to the fuel cell's operating environment. Stainless steels rich in CR, Ni, and Mo are particularly effective protective interlayers.

Nuclear reactor fuel element

DOEpatents

Johnson, Carl E.; Crouthamel, Carl E.

1980-01-01

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

Liquefied natural gas fuel use : basic training manual

DOT National Transportation Integrated Search

1994-05-01

The Urban Mass Transportation Administration's Alternative Fuel Initiative and the Environmental Protection Agency's 1991 regulations on transit bus exhaust emissions has resulted in a number of alternative fueled transit bus research and demonstrati...

COVERING A CORE BY EXTRUSION

DOEpatents

Karnie, A.J.

1963-07-16

A method of covering a cylindrical fuel core with a cladding metal ms described. The metal is forced between dies around the core from both ends in two opposing skirts, and as these meet the ends turn outward into an annular recess in the dics. By cutting off the raised portion formed by the recess, oxide impurities are eliminated. (AEC)

Electrocatalysts having gold monolayers on platinum nanoparticle cores, and uses thereof

DOEpatents

Adzic, Radoslav; Zhang, Junliang

2010-04-27

The invention relates to gold-coated particles useful as fuel cell electrocatalysts. The particles are composed of an electrocatalytically active core at least partially encapsulated by an outer shell of gold or gold alloy. The invention more particularly relates to such particles having a noble metal-containing core, and more particularly, a platinum or platinum alloy core. In other embodiments, the invention relates to fuel cells containing these electrocatalysts and methods for generating electrical energy therefrom.

TMI-2 (Three Mile Island Unit 2) core region defueling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rodabaugh, J.M.; Cowser, D.K.

1988-01-01

In July of 1982, a video camera was inserted into the Three Mile Island Unit 2 reactor vessel providing the first visual evidence of core damage. This inspection, and numerous subsequent data acquisition tasks, revealed a central void /approx/1.5 m (5 ft) deep. This void region was surrounded by partial length fuel assemblies and ringed on the periphery by /approx/40 full-length, but partial cross-section, fuel assemblies. All of the original 177 fuel assemblies exhibited signs of damage. The bottom of the void cavity was covered with a bed of granular rubble, fuel assembly upper end fittings, control rod spiders, fuelmore » rod fragments, and fuel pellets. It was obvious that the normal plant refueling system not suitable for removing the damaged core. A new system of defueling tools and equipment was necessary to perform this task. Design of the new system was started immediately, followed by >1 yr of fabrication. Delivery and checkout of the defueling system occurred in mid-1985. Actual defueling was initiated in late 1985 with removal of the debris bed at the bottom of the core void. Obstructions to the debris, such as end fittings and fuel rod fragments ere removed first; then /approx/23,000 kg (50,000lb) of granular debris was quickly loaded into canisters. Core region defueling was completed in late 1987, /approx/2 yr after it was initiated.« less

Corrosion resistant PEM fuel cell

DOEpatents

Li, Y.; Meng, W.J.; Swathirajan, S.; Harris, S.J.; Doll, G.L.

1997-04-29

The present invention contemplates a PEM fuel cell having electrical contact elements (including bipolar plates/septums) comprising a titanium nitride coated light weight metal (e.g., Al or Ti) core, having a passivating, protective metal layer intermediate the core and the titanium nitride. The protective layer forms a barrier to further oxidation/corrosion when exposed to the fuel cell`s operating environment. Stainless steels rich in Cr, Ni, and Mo are particularly effective protective interlayers. 6 figs.

Solar Electricity and Solar Fuels: Status and Perspectives in the Context of the Energy Transition.

PubMed

Armaroli, Nicola; Balzani, Vincenzo

2016-01-04

The energy transition from fossil fuels to renewables is already ongoing, but it will be a long and difficult process because the energy system is a gigantic and complex machine. Key renewable energy production data show the remarkable growth of solar electricity technologies and indicate that crystalline silicon photovoltaics (PV) and wind turbines are the workhorses of the first wave of renewable energy deployment on the TW scale around the globe. The other PV alternatives (e.g., copper/indium/gallium/selenide (CIGS) or CdTe), along with other less mature options, are critically analyzed. As far as fuels are concerned, the situation is significantly more complex because making chemicals with sunshine is far more complicated than generating electric current. The prime solar artificial fuel is molecular hydrogen, which is characterized by an excellent combination of chemical and physical properties. The routes to make it from solar energy (photoelectrochemical cells (PEC), dye-sensitized photoelectrochemical cells (DSPEC), PV electrolyzers) and then synthetic liquid fuels are presented, with discussion on economic aspects. The interconversion between electricity and hydrogen, two energy carriers directly produced by sunlight, will be a key tool to distribute renewable energies with the highest flexibility. The discussion takes into account two concepts that are often overlooked: the energy return on investment (EROI) and the limited availability of natural resources-particularly minerals-which are needed to manufacture energy converters and storage devices on a multi-TW scale. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

FUEL ASSAY REACTOR

DOEpatents

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

Analytical methods in the high conversion reactor core design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zeggel, W.; Oldekop, W.; Axmann, J.K.

High conversion reactor (HCR) design methods have been used at the Technical University of Braunschweig (TUBS) with the technological support of Kraftwerk Union (KWU). The present state and objectives of this cooperation between KWU and TUBS in the field of HCRs have been described using existing design models and current activities aimed at further development and validation of the codes. The hard physical and thermal-hydraulic boundary conditions of pressurized water reactor (PWR) cores with a high degree of fuel utilization result from the tight packing of the HCR fuel rods and the high fissionable plutonium content of the fuel. Inmore » terms of design, the problem will be solved with rod bundles whose fuel rods are adjusted by helical spacers to the proposed small rod pitches. These HCR properties require novel computational models for neutron physics, thermal hydraulics, and fuel rod design. By means of a survey of the codes, the analytical procedure for present-day HCR core design is presented. The design programs are currently under intensive development, as design tools with a solid, scientific foundation and with essential parameters that are widely valid and are required for a promising optimization of the HCR core. Design results and a survey of future HCR development are given. In this connection, the reoptimization of the PWR core in the direction of an HCR is considered a fascinating scientific task, with respect to both economic and safety aspects.« less

Fuel handling system for a nuclear reactor

DOEpatents

Saiveau, James G.; Kann, William J.; Burelbach, James P.

1986-01-01

A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

Fuel handling system for a nuclear reactor

DOEpatents

Saiveau, James G.; Kann, William J.; Burelbach, James P.

1986-12-02

A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

In-situ verification techniques for fast critical assembly cores

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brumbach, S.B.; Amundson, P.I.; Roche, C.T.

1979-01-01

Active and passive autoradiographic techniques were used to obtain piece counts of fuel plates in fast critical assembly drawers and to verify the assembly loading pattern. Active autoradiography using prompt-fission and fission-product radiation was more successful with uranium fuel while passive autoradiography was more successful with plutonium fuel. A source multiplication technique was used to measure changes in reactivity when small quantities (2-2.5 kg) of fissile material were removed from a subcritical reference core of the Zero Power Plutonium Reactor. Efforts to compensate for unsuccessful. Some compensation was achieved by replacing U-238 with polyethylene. The sensitivity for detection of partiallymore » compensated fuel removed from minimum worth regions was approximately 2.5 kg (fissile) for a core containing 2600 kg (fissile). Substitution of polyethylene was detected with a spectral index which was the ratio of the rate of the In-115 (n,..gamma..) reaction to the rate of the In-115 (n,n') reaction. This spectral index was sensitive to the presence of an 0.64-cm-thick, 5.08-cm-high polyethylene column 10-15 cm away from the indium foil. The reactivity worth of Pu-239 was also obtained as a function of location in the reactor core with the use of an inverse kinetics technique. Reactivity worths for Pu-239 varied from a maximum of 58.67 Ih/kg near the core center to a minimum of 14.86 Ih/kg at the core edge.« less

Generalized model for k -core percolation and interdependent networks

NASA Astrophysics Data System (ADS)

Panduranga, Nagendra K.; Gao, Jianxi; Yuan, Xin; Stanley, H. Eugene; Havlin, Shlomo

2017-09-01

Cascading failures in complex systems have been studied extensively using two different models: k -core percolation and interdependent networks. We combine the two models into a general model, solve it analytically, and validate our theoretical results through extensive simulations. We also study the complete phase diagram of the percolation transition as we tune the average local k -core threshold and the coupling between networks. We find that the phase diagram of the combined processes is very rich and includes novel features that do not appear in the models studying each of the processes separately. For example, the phase diagram consists of first- and second-order transition regions separated by two tricritical lines that merge and enclose a two-stage transition region. In the two-stage transition, the size of the giant component undergoes a first-order jump at a certain occupation probability followed by a continuous second-order transition at a lower occupation probability. Furthermore, at certain fixed interdependencies, the percolation transition changes from first-order → second-order → two-stage → first-order as the k -core threshold is increased. The analytic equations describing the phase boundaries of the two-stage transition region are set up, and the critical exponents for each type of transition are derived analytically.

Double-clad nuclear fuel safety rod

DOEpatents

McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

1984-01-01

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Double-clad nuclear-fuel safety rod

DOEpatents

McCarthy, W.H.; Atcheson, D.B.

1981-12-30

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Standardized verification of fuel cycle modeling

DOE PAGES

Feng, B.; Dixon, B.; Sunny, E.; ...

2016-04-05

A nuclear fuel cycle systems modeling and code-to-code comparison effort was coordinated across multiple national laboratories to verify the tools needed to perform fuel cycle analyses of the transition from a once-through nuclear fuel cycle to a sustainable potential future fuel cycle. For this verification study, a simplified example transition scenario was developed to serve as a test case for the four systems codes involved (DYMOND, VISION, ORION, and MARKAL), each used by a different laboratory participant. In addition, all participants produced spreadsheet solutions for the test case to check all the mass flows and reactor/facility profiles on a year-by-yearmore » basis throughout the simulation period. The test case specifications describe a transition from the current US fleet of light water reactors to a future fleet of sodium-cooled fast reactors that continuously recycle transuranic elements as fuel. After several initial coordinated modeling and calculation attempts, it was revealed that most of the differences in code results were not due to different code algorithms or calculation approaches, but due to different interpretations of the input specifications among the analysts. Therefore, the specifications for the test case itself were iteratively updated to remove ambiguity and to help calibrate interpretations. In addition, a few corrections and modifications were made to the codes as well, which led to excellent agreement between all codes and spreadsheets for this test case. Although no fuel cycle transition analysis codes matched the spreadsheet results exactly, all remaining differences in the results were due to fundamental differences in code structure and/or were thoroughly explained. As a result, the specifications and example results are provided so that they can be used to verify additional codes in the future for such fuel cycle transition scenarios.« less

Proliferation resistance of small modular reactors fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Polidoro, F.; Parozzi, F.; Fassnacht, F.

2013-07-01

In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. Inmore » the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.« less

Preparations to ship the TMI-2 damaged reactor core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schmitt, R.C.; Quinn, G.J.

1985-11-01

The March 1979 accident at Three Mile Island Unit 2 (TMI-2) resulted in a severely damaged core. Entries into that core using various tools and inspection devices have shown a significant void, large amounts of rubble, partially intact fuel assemblies, and some resolidified molten materials. The removal and disposition of that core has been of considerable public, regulatory, and governmental interest for some time. In a contractual agreement between General Public Utility Nuclear (GPUN) and the US Department of Energy (DOE), DOE has agreed to accept the TMI-2 core for interim storage at the Idaho National Engineering Laboratory (INEL), conductmore » research on fuel and materials of the core, and eventually dispose of the core either by processing or internment at the national repository. GPUN has removed various samples of material from the core and was scheduled to begin extensive defueling operations in September 1985. EG and G Idaho, Inc. (EG and G), acting on behalf of DOE, is responsible for transporting, receiving, examining, and storing the TMI-2 core. This paper addresses the preparations to ship the core to INEL, which is scheduled to commence in March 1986.« less

Alternative Fuel Vehicle Data Browser

EIA Publications

The annual data for 2015 about fuel use and the number of vehicles in inventory for four types of alternative fuel vehicle (AFV) fleets: federal government, state governments, transit agencies, and fuel providers, is now available. The data is available through an interactive data viewer.

Transition from the adiabatic to the sudden limit in core-level photoemission: A model study of a localized system

NASA Astrophysics Data System (ADS)

Lee, J. D.; Gunnarsson, O.; Hedin, L.

1999-09-01

We consider core electron photoemission in a localized system, where there is a charge transfer excitation. Examples are transition metal and rare earth compounds, chemisorption systems, and high-Tc compounds. The system is modeled by three electron levels, one core level, and two outer levels. In the initital state the core level and one outer level is filled (a spinless two-electron problem). This model system is embedded in a solid state environment, and the implications of our model system results for solid state photoemission are discussed. When the core hole is created, the more localized outer level (d) is pulled below the less localized level (L). The spectrum has a leading peak corresponding to a charge transfer between L and d (``shakedown''), and a satellite corresponding to no charge transfer. The model has a Coulomb interaction between these levels and the continuum states into which the core electron is emitted. The model is simple enough to allow an exact numerical solution, and with a separable potential an analytic solution. Analytic results are also obtained in lowest order perturbation theory, and in the high-energy limit of the semiclassical approximation. We calculate the ratio r(ω) between the weights of the satellite and the main peak as a function of the photon energy ω. The transition from the adiabatic to the sudden limit is found to take place for quite small kinetic energies of the photoelectron. For such small energies, the variation of the dipole matrix elements is substantial and described by the energy scale E~d. Without the coupling to the photoelectron, the corresponding ratio r0(ω) shows a smooth turn-on of the satellite intensity, due to the turn on of the dipole matrix element. The characteristic energy scales are E~d and the satellite excitation energy δE. When the interaction potential with the continuum states is introduced an energy scale E~s=1/(2R~2s) enters, where R~s is a length scale of the interaction (scattering

Understanding Common Core State Standards

ERIC Educational Resources Information Center

Kendall, John S.

2011-01-01

Now that the Common Core standards are coming to just about every school, what every school leader needs is a straightforward explanation that lays out the benefits of the Common Core in plain English, provides a succinct overview, and gets everyone thinking about how to transition to this promising new paradigm. This handy, inexpensive booklet…

Identification of fuel cycle simulator functionalities for analysis of transition to a new fuel cycle

DOE PAGES

Brown, Nicholas R.; Carlsen, Brett W.; Dixon, Brent W.; ...

2016-06-09

Dynamic fuel cycle simulation tools are intended to model holistic transient nuclear fuel cycle scenarios. As with all simulation tools, fuel cycle simulators require verification through unit tests, benchmark cases, and integral tests. Model validation is a vital aspect as well. Although compara-tive studies have been performed, there is no comprehensive unit test and benchmark library for fuel cycle simulator tools. The objective of this paper is to identify the must test functionalities of a fuel cycle simulator tool within the context of specific problems of interest to the Fuel Cycle Options Campaign within the U.S. Department of Energy smore » Office of Nuclear Energy. The approach in this paper identifies the features needed to cover the range of promising fuel cycle options identified in the DOE-NE Fuel Cycle Evaluation and Screening (E&S) and categorizes these features to facilitate prioritization. Features were categorized as essential functions, integrating features, and exemplary capabilities. One objective of this paper is to propose a library of unit tests applicable to each of the essential functions. Another underlying motivation for this paper is to encourage an international dialog on the functionalities and standard test methods for fuel cycle simulator tools.« less

Chapter 3: Simulating fire hazard across landscapes through time: integrating state-and-transition models with the Fuel Characteristic Classification System

Treesearch

Jessica E. Halofsky; Stephanie K. Hart; Miles A. Hemstrom; Joshua S. Halofsky; Morris C. Johnson

2014-01-01

Information on the effects of management activities such as fuel reduction treatments and of processes such as vegetation growth and disturbance on fire hazard can help land managers prioritize treatments across a landscape to best meet management goals. State-and-transition models (STMs) allow landscape-scale simulations that incorporate effects of succession,...

Core cooling under accident conditions at the high-flux beam reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tichler, P.; Cheng, L.; Fauske, H.

The High-Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is cooled and moderated by heavy water and contains {sup 235}U in the form of narrow-channel, parallel-plate-type fuel elements. During normal operation, the flow direction is downward through the core. This flow direction is maintained at a reduced flow rate during routine shutdown and on loss of commercial power by means of redundant pumps and power supplies. However, in certain accident scenarios, e.g. loss-of-coolant accidents (LOCAs), all forced-flow cooling is lost. Although there was experimental evidence during the reactor design period (1958-1963) that the heat removal capacity in the fullymore » developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. Accordingly, a test program was initiated using an electrically heated section to simulate the fuel channel and a cooling loop to simulate the balance of the primary cooling system.« less

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

PubMed

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

PubMed Central

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

Ligand field splittings in core level transitions for transition metal (TM) oxides: Tanabe-Sugano diagrams and (TM) dangling bonds in vacated O-atom defects

NASA Astrophysics Data System (ADS)

Lucovsky, Gerry; Wu, Kun; Pappas, Brian; Whitten, Jerry

2013-04-01

Defect states in the forbidden band-gap below the conduction band edge are active as electron traps in nano-grain high-) transition metal (TM) oxides with thickness >0.3 nm, e.g., ZrO2 and HfO2. These oxides have received considerable attention as gate-dielectrics in complementary metal oxide semiconductor (CMOS) devices, and more recently are emerging as candidates for charge storage and memory devices. To provide a theoretical basis for device functionality, ab-initio many-electron theory is combined with X-ray absorption spectroscopy (XAS) to study O K edge and TM core level transitions. These studies identify ligand field splittings (ΔLF) for defect state features,. When compared with those obtained from O-atom and TM-atom core spectroscopic transitions, this provides direct information about defect state sun-nm bonding arrangements. comparisons are made for (i) elemental TiO2 and Ti2O3 with different formal ionic charges, Ti4+ and Ti3+ and for (ii) Magneli Phase alloys, TinO2n-1, n is an integer 9>=n>3, and (TiO2)x(HfO2)1-x alloys. The alloys display multi-valent behavior from (i) different ionic-charge states, (ii} local bond-strain, and (iii) metallic hopping transport. The intrinsic bonding defects in TM oxides are identified as pairs of singly occupied dangling bonds. For 6-fold coordinated Ti-oxides defect excited states in 2nd derivative O K pre-edge spectra are essentially the same as single Ti-atom d2 transitions in Tanabe-Sugano (T-S) diagrams. O-vacated site defects in 8-fold coordinated ZrO2 and HfO2 are described by d8 T-S diagrams. T-S defect state ordering and splittings are functions of the coordination and symmetry of vacated site bordering TM atoms. ΔLF values from the analysis of T-S diagrams indicate medium range order (MRO) extending to 3rd and 4th nearest-neighbor (NN) TM-atoms. Values are different for 6-fold Ti, and 8-fold ZrO2 and HfO2, and scale inversely with differences in respective formal ionic radii. O-vacated site bonding

Applications of plasma core reactors to terrestrial energy systems

NASA Technical Reports Server (NTRS)

Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

1974-01-01

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

The transition from the core vowels to the following segments in Japanese children who stutter: the second, third and fourth syllables.

PubMed

Matsumoto-Shimamori, Sachiyo; Ito, Tomohiko; Fukuda, Suzy E; Fukuda, Shinji

2011-09-01

Shimamori and Ito (2007 , Syllable weight and phonological encoding in Japanese children who stutter. Japanese Journal of Special Education, 44, 451-462; 2008, Syllable weight and frequency of stuttering: Comparison between children who stutter with and without a family history of stuttering. Japanese Journal of Special Education, 45, 437-445; 2009, Difference in frequency of stuttering between light and heavy syllables in the production of monosyllables: From the viewpoint of phonetic transition. The Japanese Journal of Logopedics and Phoniatrics, 50, 116-122 (in Japanese)) proposed the hypothesis that in Japanese the transition from the core vowels (CVs) to the following segments affected the occurrence of stuttering. However, the transition we investigated was in the first syllables only, and the effect of the transition in second, third and fourth syllables was not addressed. The purpose of this study was to investigate whether the transition from the CVs in the second, third and fourth syllables affected the occurrence of stuttering. The participants were 21 Japanese children. A non-word naming task and a non-word reading task were used. The frequency of stuttering was not significantly different where the number of transitions from the CVs differed on either task. These results suggest that the transition from the CVs in the second, third and fourth syllables does not have a significant effect on the occurrence of stuttering in Japanese.

Seed and blanket fuel arrangement for dual-phase nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Congdon, S.P.; Fawcett, R.M.

1992-09-22

This patent describes a fuel management method for a dual-phase nuclear reactor, it comprises: installing a fuel bundle at a first core location accessed by coolant through a relatively small aperture, each of the bundles having a predetermined group of fuel elements; operating the reactor a first time; shutting down the reactor; reinstalling the fuel bundle at a second core location accessed by coolant through a relatively large aperture; and operating the reactor a second time.

Alternative Fuels Data Center: Alternative Fuel Vehicles Beat the Heat,

Science.gov Websites

standard for biodiesel, American Society for Testing Materials (ASTM) D675 fuel. "E85 is widely maintenance, Roaring Fork Transit Authority, Aspen, Colorado When thinking about buying an alternative fuel Gas Vehicles in Subzero Temps Up in the high terrain of the Colorado mountains, the Roaring Fork

Magnetic moments, E3 transitions and the structure of high-spin core excited states in 211Rn

NASA Astrophysics Data System (ADS)

Poletti, A. R.; Dracoulis, G. D.; Byrne, A. P.; Stuchbery, A. E.; Poletti, S. J.; Gerl, J.; Lewis, P. M.

1985-05-01

The results of g-factor measurements of high-spin states in 211Rn are: Ex = 8856 + Δ' keV (Jπ = 63/2-), g = 0.626(7); 6101 + Δ' KeV (49/2+), 0.766(8); 5347 + Δ' KeV (43/2-), 0.74(2); 3927 + Δ KeV (35/2+), 1.017(12); 1578 + Δ KeV (17/2-), 0.912(9). These results together with measured E3 transition strengths and shell model calculations are used to assign configurations to the core excited states in 211Rn. Mixed configurations are required to explain the g-factors and enhanced E3 strengths simultaneously.

Charge order-superfluidity transition in a two-dimensional system of hard-core bosons and emerging domain structures

NASA Astrophysics Data System (ADS)

Moskvin, A. S.; Panov, Yu. D.; Rybakov, F. N.; Borisov, A. B.

2017-11-01

We have used high-performance parallel computations by NVIDIA graphics cards applying the method of nonlinear conjugate gradients and Monte Carlo method to observe directly the developing ground state configuration of a two-dimensional hard-core boson system with decrease in temperature, and its evolution with deviation from a half-filling. This has allowed us to explore unconventional features of a charge order—superfluidity phase transition, specifically, formation of an irregular domain structure, emergence of a filamentary superfluid structure that condenses within of the charge-ordered phase domain antiphase boundaries, and formation and evolution of various topological structures.

Very high temperature behavior of HTGR core materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soo, P.; Uneberg, G.; Sabatini, R.

1978-01-01

A description is given of experiments to investigate the behavior of HTGR core materials during hypothetical heatup accidents in which the core temperature is assumed to reach values between 2400/sup 0/C and the graphite sublimation range (>3600/sup 0/C). The work includes BISO coated fuel particle failure, simulated fission product migration in core graphite, and graphite sublimation behavior.

Back-Up/ Peak Shaving Fuel Cell System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Staudt, Rhonda L.

2008-05-28

This Final Report covers the work executed by Plug Power from 8/11/03 – 10/31/07 statement of work for Topic 2: advancing the state of the art of fuel cell technology with the development of a new generation of commercially viable, stationary, Back-up/Peak-Shaving fuel cell systems, the GenCore II. The Program cost was $7.2 M with the Department of Energy share being $3.6M and Plug Power’s share being $3.6 M. The Program started in August of 2003 and was scheduled to end in January of 2006. The actual program end date was October of 2007. A no cost extension was grated.more » The Department of Energy barriers addressed as part of this program are: Technical Barriers for Distributed Generation Systems: o Durability o Power Electronics o Start up time Technical Barriers for Fuel Cell Components: o Stack Material and Manufacturing Cost o Durability o Thermal and water management Background The next generation GenCore backup fuel cell system to be designed, developed and tested by Plug Power under the program is the first, mass-manufacturable design implementation of Plug Power’s GenCore architected platform targeted for battery and small generator replacement applications in the telecommunications, broadband and UPS markets. The next generation GenCore will be a standalone, H2 in-DC-out system. In designing the next generation GenCore specifically for the telecommunications market, Plug Power is teaming with BellSouth Telecommunications, Inc., a leading industry end user. The final next generation GenCore system is expected to represent a market-entry, mass-manufacturable and economically viable design. The technology will incorporate: • A cost-reduced, polymer electrolyte membrane (PEM) fuel cell stack tailored to hydrogen fuel use • An advanced electrical energy storage system • A modular, scalable power conditioning system tailored to market requirements • A scaled-down, cost-reduced balance of plant (BOP) • Network Equipment Building Standards

The Rydberg electronic transitions of the hydrogen molecule

DOE Office of Scientific and Technical Information (OSTI.GOV)

Babb, J.F.; Chang, E.S.

1992-01-01

Transition energies and relative line strengths, without Boltzmann weighting, for the electric dipole transitions between Rydberg states n{prime}L{prime} and nL of the hydrogen molecule (one electron in a near-hydrogenic state of high n and L, with n the principal quantum number and L the orbital angular momentum quantum number of the electron) are calculated. Since the H{sup +}{sub 2} core is loosely coupled to the Rydberg electron, numerous lines occur, depending on the vector sum of L and the core rotational angular momentum. For the core vibrational quantum numbers v = 0 to 5 the strongest lines among the P,more » Q, and R branches for the lowest 12 core rotational levels are given for the particular transition arrays 6h-5g, 8i-6h, 7i-6h, 8k-7i, and 9l-8k, for which transitions occur in the wave number range 350 to 1,400 cm {sup {minus}1}.« less

National fuel cell bus program : proterra fuel cell hybrid bus report, Columbia demonstration.

DOT National Transportation Integrated Search

2011-10-01

This report summarizes the experience and early results from a fuel cell bus demonstration funded by the Federal Transit Administration (FTA) under the National Fuel Cell Bus Program. A team led by the Center for Transportation and the Environment an...

America's Next Great Ship: Space Launch System Core Stage Transitioning from Design to Manufacturing

NASA Technical Reports Server (NTRS)

Birkenstock, Benjamin; Kauer, Roy

2014-01-01

The Space Launch System (SLS) Program is essential to achieving the Nation's and NASA's goal of human exploration and scientific investigation of the solar system. As a multi-element program with emphasis on safety, affordability, and sustainability, SLS is becoming America's next great ship of exploration. The SLS Core Stage includes avionics, main propulsion system, pressure vessels, thrust vector control, and structures. Boeing manufactures and assembles the SLS core stage at the Michoud Assembly Facility (MAF) in New Orleans, LA, a historical production center for Saturn V and Space Shuttle programs. As the transition from design to manufacturing progresses, the importance of a well-executed manufacturing, assembly, and operation (MA&O) plan is crucial to meeting performance objectives. Boeing employs classic techniques such as critical path analysis and facility requirements definition as well as innovative approaches such as Constraint Based Scheduling (CBS) and Cirtical Chain Project Management (CCPM) theory to provide a comprehensive suite of project management tools to manage the health of the baseline plan on both a macro (overall project) and micro level (factory areas). These tools coordinate data from multiple business systems and provide a robust network to support Material & Capacity Requirements Planning (MRP/CRP) and priorities. Coupled with these tools and a highly skilled workforce, Boeing is orchestrating the parallel buildup of five major sub assemblies throughout the factory. Boeing and NASA are transforming MAF to host state of the art processes, equipment and tooling, the most prominent of which is the Vertical Assembly Center (VAC), the largest weld tool in the world. In concert, a global supply chain is delivering a range of structural elements and component parts necessary to enable an on-time delivery of the integrated Core Stage. SLS is on plan to launch humanity into the next phase of space exploration.

MOX fuel assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reese, A.P.; Crowther, R.L. Jr.

1992-02-18

This patent describes improvement in a boiling water reactor core having a plurality of vertically upstanding fuel bundles; each fuel bundle containing longitudinally extending sealed rods with fissile material therein; the improvement comprises the fissile material including a mixture of uranium and recovered plutonium in rods of the fuel bundle at locations other than the corners of the fuel bundle; and, neutron absorbing material being located in rods of the fuel bundle at rod locations adjacent the corners of the fuel bundles whereby the neutron absorbing material has decreased shielding from the plutonium and maximum exposure to thermal neutrons formore » shaping the cold reactivity shutdown zone in the fuel bundle.« less

Core thermal response and hydrogen generation of the N Reactor hydrogen mitigation design basis accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, M.D.; Lombardo, N.J.; Heard, F.J.

1988-04-01

Calculations were performed to determine core heatup, core damage, and subsequent hydrogen production of a hypothetical loss-of-cooling accident at the Department of Energy's N Reactor. The thermal transient response of the reactor core was solved using the TRUMP-BD computer program. Estimates of whole-core thermal damage and hydrogen production were made by weighting the results of multiple half-length pressure tube simulations at various power levels. The Baker-Just and Wilson parabolic rate equations for the metal-water chemical reactions modeled the key phenomena of chemical energy and hydrogen evolution. Unlimited steam was assumed available for continuous oxidation of exposed Zircaloy-2 surfaces and formore » uranium metal with fuel cladding beyond the failure temperature (1038 C). Intact fuel geometry was modeled. Maximum fuel temperatures (1181 C) in the cooled central regions of the core were predicted to occur one-half hour into the accident scenario. Maximum fuel temperatures of 1447 C occurred in the core GSCS-regions at the end of the 10-h transient. After 10-h 26% of the fuel inventory was predicted to have failed. Peak hydrogen evolution equaled 42 g/s, while 10-h integrated hydrogen evolution equaled 167 kg. 12 refs., 12 figs., 2 tabs.« less

Promising transit applications of fuel cells and alternative fuels

DOT National Transportation Integrated Search

2002-06-01

For over a decade, the Volpe Center has been providing technical support to the Federal Transit Administration (FTA) Office of Research, Demonstration and Innovation towards the development, deployment, field test and safety evaluation of advanced tr...

Fuel Cell Electric Bus Evaluations | Hydrogen and Fuel Cells | NREL

Science.gov Websites

. Transit Fleets: Current Status 2017, L. Eudy and M. Post (November 2017) Zero Emission Bay Area (ZEBA ) Fuel Cell Bus Demonstration Results: Sixth Report, L. Eudy, M. Post, and M. Jeffers (September 2017 2017) American Fuel Cell Bus Project Evaluation: Third Report, L. Eudy, M. Post, and M. Jeffers (May

Strangeness driven phase transitions in compressed baryonic matter and their relevance for neutron stars and core collapsing supernovae

DOE Office of Scientific and Technical Information (OSTI.GOV)

Raduta, Ad. R.; Gulminelli, F.; Oertel, M.

2015-02-24

We discuss the thermodynamics of compressed baryonic matter with strangeness within non-relativistic mean-field models with effective interactions. The phase diagram of the full baryonic octet under strangeness equilibrium is built and discussed in connection with its relevance for core-collapse supernovae and neutron stars. A simplified framework corresponding to (n, p, Λ)(+e)-mixtures is employed in order to test the sensitivity of the existence of a phase transition on the (poorely constrained) interaction coupling constants and the compatibility between important hyperonic abundances and 2M{sub ⊙} neutron stars.

Performance of U3Si2 Fuel in a Reactivity Insertion Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, Lap Y.; Cuadra, Arantxa; Todosow, Michael

In this study we examined the performance of the U3Si2 fuel cladded with Zircaloy (Zr) in a reactivity insertion accident (RIA) in a PWR core. The power excursion as a result of a $1 reactivity insertion was calculated by a TRACE PWR plant model using point-kinetics, for alternative cores with UO2 and U3Si2 fuel assemblies. The point-kinetics parameters (feedback coefficients, prompt-neutron lifetime and group constants for six delayed-neutron groups) were obtained from beginning-of-cycle equilibrium full core calculations with PARCS. In the PARCS core calculations, the few-group parameters were developed utilizing the TRITON/NEWT tools in the SCALE package. In order tomore » assess the fuel response in finer detail (e.g. the maximum fuel temperature) the power shape and thermal boundary conditions from the TRACE/PARCS calculations were used to drive a BISON model of a fuel pin with U3Si2 and UO2 respectively. For a $1 reactivity transient both TRACE and BISON predicted a higher maximum fuel temperature for the UO2 fuel than the U3Si2 fuel. Furthermore, BISON is noted to calculate a narrower gap and a higher gap heat transfer coefficient than TRACE. This resulted in BISON predicting consistently lower fuel temperatures than TRACE. This study also provides a systematic comparison between TRACE and BISON using consistent transient boundary conditions. The TRACE analysis of the RIA only reflects the core-wide response in power. A refinement to the analysis would be to predict the local peaking in a three-dimensional core as a result of control rod ejection.« less

Mass transit : use of alternative fuels in transit buses

DOT National Transportation Integrated Search

1999-12-14

Improving air quality in urban settings has been a long-standing national objective. Transit buses powered by diesel engines have been identified as contributors to air pollution in these areas. The Transportation Equity Act for the 21st Century (TEA...

Tailored Core Shell Cathode Powders for Solid Oxide Fuel Cells

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swartz, Scott

2015-03-23

In this Phase I SBIR project, a “core-shell” composite cathode approach was evaluated for improving SOFC performance and reducing degradation of lanthanum strontium cobalt ferrite (LSCF) cathode materials, following previous successful demonstrations of infiltration approaches for achieving the same goals. The intent was to establish core-shell cathode powders that enabled high performance to be obtained with “drop-in” process capability for SOFC manufacturing (i.e., rather than adding an infiltration step to the SOFC manufacturing process). Milling, precipitation and hetero-coagulation methods were evaluated for making core-shell composite cathode powders comprised of coarse LSCF “core” particles and nanoscale “shell” particles of lanthanum strontiummore » manganite (LSM) or praseodymium strontium manganite (PSM). Precipitation and hetero-coagulation methods were successful for obtaining the targeted core-shell morphology, although perfect coverage of the LSCF core particles by the LSM and PSM particles was not obtained. Electrochemical characterization of core-shell cathode powders and conventional (baseline) cathode powders was performed via electrochemical impedance spectroscopy (EIS) half-cell measurements and single-cell SOFC testing. Reliable EIS testing methods were established, which enabled comparative area-specific resistance measurements to be obtained. A single-cell SOFC testing approach also was established that enabled cathode resistance to be separated from overall cell resistance, and for cathode degradation to be separated from overall cell degradation. The results of these EIS and SOFC tests conclusively determined that the core-shell cathode powders resulted in significant lowering of performance, compared to the baseline cathodes. Based on the results of this project, it was concluded that the core-shell cathode approach did not warrant further investigation.« less

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

Candidate molten salt investigation for an accelerator driven subcritical core

NASA Astrophysics Data System (ADS)

Sooby, E.; Baty, A.; Beneš, O.; McIntyre, P.; Pogue, N.; Salanne, M.; Sattarov, A.

2013-09-01

We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated. A special thanks is due to Prof. Paul Madden for introducing the ADSMS group to the concept of using the molten salt as the spallation target, rather than a conventional heavy metal spallation target. This feature helps to optimize this core as a Pu/TRU burner.

FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

DOEpatents

Flint, O.

1961-01-10

Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

Injector for liquid fueled rocket engine

NASA Technical Reports Server (NTRS)

Cornelius, Charles S. (Inventor); Myers, W. Neill (Inventor); Shadoan, Michael David (Inventor); Sparks, David L. (Inventor)

2000-01-01

An injector for liquid fueled rocket engines wherein a generally flat core having a frustoconical dome attached to one side of the core to serve as a manifold for a first liquid, with the core having a generally circular configuration having an axis. The other side of the core has a plurality of concentric annular first slots and a plurality of annular concentric second slots alternating with the first slots, the second slots having a greater depth than said first slots. A bore extends through the core for inletting a second liquid into said core, the bore intersecting the second slots to feed the second liquid into the second slots. The core also has a plurality of first passageways leading from the manifold to the first annular slots for feeding the first liquid into said first slots. A faceplate brazed to said other side of the core is provided with apertures extending from the first and second slots through said face plate, these apertures being positioned to direct fuel and liquid oxygen into contact with each other in the combustion chamber. The first liquid may be liquid oxygen and the second liquid may be kerosene or liquid hydrogen.

Application of X-Ray Computer Tomography for Observing the Central Void Formations and the Fuel Pin Deformations of Irradiated FBR Fuel Assemblies

NASA Astrophysics Data System (ADS)

Katsuyama, Kozo; Nagamine, Tsuyoshi; Furuya, Hirotaka

2010-10-01

In order to observe the structural change in the interior of irradiated fuel assemblies, a non-destructive post-irradiation examination (PIE) technique using X-ray computer tomography (X-ray CT) was developed. This X-ray CT technique was applied to observe the central void formations and fuel pin deformations of fuel assemblies which had been irradiated at high linear heat rating. The central void sizes in all fuel pins were measured on five cross sections of the core fuel column as a parameter for evaluating fuel thermal performance. In addition, the fuel pin deformations were analyzed from X-ray CT images obtained along the axial direction of a fuel assembly at the same separation interval. A dependence of void size on the linear heat rating was seen in the fuel assembly irradiated at high linear heat rating. In addition, significant undulations of the fuel pin were observed along the axial direction, coinciding with the wrapping wire pitch in the core fuel column. Application of the developed technique should provide enhanced resolution of measurements and simplify fuel PIEs.

Connecticut Nutmeg Fuel Cell Bus Project : Demonstrating Advanced-Design Hybrid Fuel Cell Buses in Connecticut

DOT National Transportation Integrated Search

2011-07-01

The Federal Transit Administrations (FTA) National Fuel Cell Bus Program (NFCBP) focuses on developing commercially viable fuel cell bus technologies. The Northeast Advanced Vehicle Consortium (NAVC) is one of three non-profit consortia chosen to ...

TREAT Transient Analysis Benchmarking for the HEU Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D. C.; Connaway, H. M.; Wright, A. E.

2014-05-01

This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used tomore » determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term “reported” values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core’s performance.« less

Evaluation of Particle Counter Technology for Detection of Fuel Contamination Detection Utilizing Fuel System Supply Point

DTIC Science & Technology

2014-06-19

product used as a diesel product for ground use (1). Free water contamination (droplets) may appear as fine droplets or slugs of water in the fuel...methods and test procedures for the calibration and use of automatic particle counters. The transition of this technology to the fuel industry is...UNCLASSIFIED 6 UNCLASSIFIED Receipt Vehicle Fuel Tank Fuel Injector Aviation Fuel DEF (AUST) 5695B 18/16/13 Parker 18

Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Douglas W. Akers; Edwin A. Harvego

2012-08-01

This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and datamore » on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.« less

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

DOE PAGES

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; ...

2014-11-04

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

The Future of Low-Carbon Transportation Fuels

NASA Astrophysics Data System (ADS)

Yang, Christopher; Yeh, Sonia

2011-11-01

Petroleum fuel uses make up essentially all of transportation fuel usage today and will continue to dominate transportation fuel usage well into future without any major policy changes. This chapter focuses on low-carbon transportation fuels, specifically, biofuels, electricity and hydrogen, that are emerging options to displace petroleum based fuels. The transition to cleaner, lower carbon fuel sources will need significant technology advancement, and sustained coordination efforts among the vehicle and fuel industry and policymakers/regulators over long period of time in order to overcome market barriers, consumer acceptance, and externalities of imported oil. We discuss the unique infrastructure challenges, and compare resource, technology, economics and transitional issues for each of these fuels. While each fuel type has important technical and implementation challenges to overcome (including vehicle technologies) in order to contribute a large fraction of our total fuel demand, it is important to note that a portfolio approach will give us the best chance of meeting stringent environmental and energy security goals for a sustainable transportation future.

Investigation, quantification, and recommendations : performance of alternatively fueled buses.

DOT National Transportation Integrated Search

2014-08-01

The goal of this project was to continue consistent collection and reporting of data on the performance and costs of alternatively fueled public transit vehicles in the U.S. transit fleet in order to keep the Bus Fuels Fleet Evaluation Tool (BuFFeT; ...

Droplet Core Nuclear Rocket (DCNR)

NASA Technical Reports Server (NTRS)

Anghaie, Samim

1991-01-01

The most basic design feature of the droplet core nuclear reactor is to spray liquid uranium into the core in the form of droplets on the order of five to ten microns in size, to bring the reactor to critical conditions. The liquid uranium fuel ejector is driven by hydrogen, and more hydrogen is injected from the side of the reactor to about one and a half meters from the top. High temperature hydrogen is expanded through a nozzle to produce thrust. The hydrogen pressure in the system can be somewhere between 50 and 500 atmospheres; the higher pressure is more desirable. In the lower core region, hydrogen is tangentially injected to serve two purposes: (1) to provide a swirling flow to protect the wall from impingement of hot uranium droplets: (2) to generate a vortex flow that can be used for fuel separation. The reactor is designed to maximize the energy generation in the upper region of the core. The system can result in and Isp of 2000 per second, and a thrust-to-weight ratio of 1.6 for the shielded reactor. The nuclear engine system can reduce the Mars mission duration to less than 200 days. It can reduce the hydrogen consumption by a factor of 2 to 3, which reduces the hydrogen load by about 130 to 150 metric tons.

Target-fueled nuclear reactor for medical isotope production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coats, Richard L.; Parma, Edward J.

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7more » to 21 days.« less

Thermal-Hydraulic Transient Analysis of a Packed Particle Bed Reactor Fuel Element

DTIC Science & Technology

1990-06-01

long fuel elements, arranged to form a core , were analyzed for an up-power transient from 0 MWt to approximately 18 MWt. The simple model significantly...VARIATIONS IN FUEL ELEMENT GEOMETRY ............. 60 4.4 VARIATIONS IN THE MANNER OF TRANSIENT CONTROL ..... 62 4.5 CORE REPRESENTATION BY MULTIPLE FUEL ...the HTGR , however, the PBR packs small fuel particles between inner and outer retention elements, designated as frits. The PBR is appropriate for a

Induction simulation of gas core nuclear engine

NASA Technical Reports Server (NTRS)

Poole, J. W.; Vogel, C. E.

1973-01-01

The design, construction and operation of an induction heated plasma device known as a combined principles simulator is discussed. This device incorporates the major design features of the gas core nuclear rocket engine such as solid feed, propellant seeding, propellant injection through the walls, and a transpiration cooled, choked flow nozzle. Both argon and nitrogen were used as propellant simulating material, and sodium was used for fuel simulating material. In addition, a number of experiments were conducted utilizing depleted uranium as the fuel. The test program revealed that satisfactory operation of this device can be accomplished over a range of operating conditions and provided additional data to confirm the validity of the gas core concept.

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

NASA Astrophysics Data System (ADS)

Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

2016-05-01

The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.

Power Peaking Effect of OTTO Fuel Scheme Pebble Bed Reactor

NASA Astrophysics Data System (ADS)

Setiadipura, T.; Suwoto; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

Pebble Bed Reactor (PBR) type of Hight Temperature Gas-cooled Reactor (HTGR) is a very interesting nuclear reactor design to fulfill the growing electricity and heat demand with a superior passive safety features. Effort to introduce the PBR design to the market can be strengthen by simplifying its system with the Once-through-then-out (OTTO) cycle PBR in which the pebble fuel only pass the core once. Important challenge in the OTTO fuel scheme is the power peaking effect which limit the maximum nominal power or burnup of the design. Parametric survey is perform in this study to investigate the contribution of different design parameters to power peaking effect of OTTO cycle PBR. PEBBED code is utilized in this study to perform the equilibrium PBR core analysis for different design parameter and fuel scheme. The parameters include its core diameter, height-per-diameter (H/D), power density, and core nominal power. Results of this study show that diameter and H/D effectsare stronger compare to the power density and nominal core power. Results of this study might become an importance guidance for design optimization of OTTO fuel scheme PBR.

Molten core retention assembly

DOEpatents

Lampe, Robert F.

1976-06-22

Molten fuel produced in a core overheating accident is caught by a molten core retention assembly consisting of a horizontal baffle plate having a plurality of openings therein, heat exchange tubes having flow holes near the top thereof mounted in the openings, and a cylindrical, imperforate baffle attached to the plate and surrounding the tubes. The baffle assembly is supported from the core support plate of the reactor by a plurality of hanger rods which are welded to radial beams passing under the baffle plate and intermittently welded thereto. Preferably the upper end of the cylindrical baffle terminates in an outwardly facing lip to which are welded a plurality of bearings having slots therein adapted to accept the hanger rods.

Zero Emission Bay Area (ZEBA) Fuel Cell Bus Demonstration Results. Fourth Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eudy, Leslie; Post, Matthew

This report presents results of a demonstration of fuel cell electric buses (FCEB) operating in Oakland, California. Alameda-Contra Costa Transit District (AC Transit) leads the Zero Emission Bay Area (ZEBA) demonstration, which includes 12 advanced-design fuel cell buses and two hydrogen fueling stations. The FCEBs in service at AC Transit are 40-foot, low-floor buses built by Van Hool with a hybrid electric propulsion system that includes a US Hybrid fuel cell power system and EnerDel lithium-based energy storage system. The buses began revenue service in May 2010.

Preliminary engineering design of sodium-cooled CANDLE core

NASA Astrophysics Data System (ADS)

Takaki, Naoyuki; Namekawa, Azuma; Yoda, Tomoyuki; Mizutani, Akihiko; Sekimoto, Hiroshi

2012-06-01

The CANDLE burning process is characterized by the autonomous shifting of burning region with constant reactivity and constant spacial power distribution. Evaluations of such critical burning process by using widely used neutron diffusion and burning codes under some realistic engineering constraints are valuable to confirm the technical feasibility of the CANDLE concept and to put the idea into concrete core design. In the first part of this paper, it is discussed that whether the sustainable and stable CANDLE burning process can be reproduced even by using conventional core analysis tools such as SLAROM and CITATION-FBR. As a result, it is certainly possible to demonstrate it if the proper core configuration and initial fuel composition required as CANDLE core are applied to the analysis. In the latter part, an example of a concrete image of sodium cooled, metal fuel, 2000MWt rating CANDLE core has been presented by assuming an emerging inevitable technology of recladding. The core satisfies engineering design criteria including cladding temperature, pressure drop, linear heat rate, and cumulative damage fraction (CDF) of cladding, fast neutron fluence and sodium void reactivity which are defined in the Japanese FBR design project. It can be concluded that it is feasible to design CADLE core by using conventional codes while satisfying some realistic engineering design constraints assuming that recladding at certain time interval is technically feasible.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Shackleford, M.H.

1958-12-16

A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

Monitoring arrangement for vented nuclear fuel elements

DOEpatents

Campana, Robert J.

1981-01-01

In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

Radioactivity of spent TRIGA fuel

NASA Astrophysics Data System (ADS)

Usang, M. D.; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

2015-04-01

Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara

2005-02-06

Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an earlymore » prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

DOE PAGES

Carmack, W. Jon; Chichester, Heather M.; Porter, Douglas L.; ...

2016-02-27

The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This then places the peakmore » fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. After comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.« less

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, W. J.; Chichester, H. M.; Porter, D. L.

2016-05-01

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peakmore » fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.« less

Safety and core design of large liquid-metal cooled fast breeder reactors

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

Alternative Fuels Data Center: Active Transit

Science.gov Websites

: Active Transit on AddThis.com... More in this section... Idle Reduction Parts & Equipment Maintenance Reduction Parts & Equipment Maintenance Driving Behavior Fleet Rightsizing System Efficiency Locate

Alternative Fuels in Transportation : Workforce needs and opportunities in support of reducing reliance on petroleum fuels

DOT National Transportation Integrated Search

2016-01-01

An overreliance on foreign oil and the negative impacts of using petroleum fuels on the worlds climate have prompted energy policies that support the diversification of transport fuels and aggressive work to transition to non-petroleum options. Th...

Non-homogeneous hybrid rocket fuel for enhanced regression rates utilizing partial entrainment

NASA Astrophysics Data System (ADS)

Boronowsky, Kenny

A concept was developed and tested to enhance the performance and regression rate of hydroxyl terminated polybutadiene (HTPB), a commonly used hybrid rocket fuel. By adding small nodules of paraffin into the HTPB fuel, a non-homogeneous mixture was created resulting in increased regression rates. The goal was to develop a fuel with a simplified single core geometry and a tailorable regression rate. The new fuel would benefit from the structural stability of HTPB yet not suffer from the large void fraction representative of typical HTPB core geometries. Regression rates were compared between traditional HTPB single core grains, 85% HTPB mixed with 15% (by weight) paraffin cores, 70% HTPB mixed with 30% paraffin cores, and plain paraffin single core grains. Each fuel combination was tested at oxidizer flow rates, ranging from 0.9 - 3.3 g/s of gaseous oxygen, in a small scale hybrid test rocket and average regression rates were measured. While large uncertainties were present in the experimental setup, the overall data showed that the regression rate was enhanced as paraffin concentration increased. While further testing would be required at larger scales of interest, the trends are encouraging. Inclusion of paraffin nodules in the HTPB grain may produce a greater advantage than other more noxious additives in current use. In addition, it may lead to safer rocket motors with higher integrated thrust due to the decreased void fraction.

Emissions of toxic pollutants from compressed natural gas and low sulfur diesel-fueled heavy-duty transit buses tested over multiple driving cycles.

PubMed

Kado, Norman Y; Okamoto, Robert A; Kuzmicky, Paul A; Kobayashi, Reiko; Ayala, Alberto; Gebel, Michael E; Rieger, Paul L; Maddox, Christine; Zafonte, Leo

2005-10-01

The number of heavy-duty vehicles using alternative fuels such as compressed natural gas (CNG) and new low-sulfur diesel fuel formulations and equipped with after-treatment devices are projected to increase. However, few peer-reviewed studies have characterized the emissions of particulate matter (PM) and other toxic compounds from these vehicles. In this study, chemical and biological analyses were used to characterize the identifiable toxic air pollutants emitted from both CNG and low-sulfur-diesel-fueled heavy-duty transit buses tested on a chassis dynamometer over three transient driving cycles and a steady-state cruise condition. The CNG bus had no after-treatment, and the diesel bus was tested first equipped with an oxidation catalyst (OC) and then with a catalyzed diesel particulate filter (DPF). Emissions were analyzed for PM, volatile organic compounds (VOCs; determined on-site), polycyclic aromatic hydrocarbons (PAHs), and mutagenic activity. The 2000 model year CNG-fueled vehicle had the highest emissions of 1,3-butadiene, benzene, and carbonyls (e.g., formaldehyde) of the three vehicle configurations tested in this study. The 1998 model year diesel bus equipped with an OC and fueled with low-sulfur diesel had the highest emission rates of PM and PAHs. The highest specific mutagenic activities (revertants/microg PM, or potency) and the highest mutagen emission rates (revertants/mi) were from the CNG bus in strain TA98 tested over the New York Bus (NYB) driving cycle. The 1998 model year diesel bus with DPF had the lowest VOCs, PAH, and mutagenic activity emission. In general, the NYB driving cycle had the highest emission rates (g/mi), and the Urban Dynamometer Driving Schedule (UDDS) had the lowest emission rates for all toxics tested over the three transient test cycles investigated. Also, transient emissions were, in general, higher than steady-state emissions. The emissions of toxic compounds from an in-use CNG transit bus (without an oxidation

Uranium droplet core nuclear rocket

NASA Technical Reports Server (NTRS)

Anghaie, Samim

1991-01-01

Uranium droplet nuclear rocket is conceptually designed to utilize the broad temperature range ofthe liquid phase of metallic uranium in droplet configuration which maximizes the energy transfer area per unit fuel volume. In a baseline system dissociated hydrogen at 100 bar is heated to 6000 K, providing 2000 second of Isp. Fission fragments and intense radian field enhance the dissociation of molecular hydrogen beyond the equilibrium thermodynamic level. Uranium droplets in the core are confined and separated by an axisymmetric vortex flow generated by high velocity tangential injection of hydrogen in the mid-core regions. Droplet uranium flow to the core is controlled and adjusted by a twin flow nozzle injection system.

CT Scans of Cores Metadata, Barrow, Alaska 2015

DOE Data Explorer

Katie McKnight; Tim Kneafsey; Craig Ulrich

2015-03-11

Individual ice cores were collected from Barrow Environmental Observatory in Barrow, Alaska, throughout 2013 and 2014. Cores were drilled along different transects to sample polygonal features (i.e. the trough, center and rim of high, transitional and low center polygons). Most cores were drilled around 1 meter in depth and a few deep cores were drilled around 3 meters in depth. Three-dimensional images of the frozen cores were constructed using a medical X-ray computed tomography (CT) scanner. TIFF files can be uploaded to ImageJ (an open-source imaging software) to examine soil structure and densities within each core.

Evaluation of nuclear fuel reprocessing strategies. 2. LWR fuel storage, recycle economics and plutonium logistics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prince, B.E.; Hadley, S.W.

1983-10-27

This is the second of a two-part report intended as a critical review of certain issues involved with closing the Light Water Reactor (LWR) fuel cycle and establishing the basis for future transition to commercial breeder applications. The report is divided into four main sections consisting of (1) a review of the status of the LWR spent fuel management and storage problem; (2) an analysis of the economic incentives for instituting reprocessing and recycle in LWRs; (3) an analysis of the time-dependent aspects of plutonium economic value particularly as related to the LWR-breeder transition; and (4) an analysis of themore » time-dependent aspects of plutonium requirements and supply relative to this transition.« less

Experimental investigations of heat transfer and temperature fields in models simulating fuel assemblies used in the core of a nuclear reactor with a liquid heavy-metal coolant

NASA Astrophysics Data System (ADS)

Belyaev, I. A.; Genin, L. G.; Krylov, S. G.; Novikov, A. O.; Razuvanov, N. G.; Sviridov, V. G.

2015-09-01

The aim of this experimental investigation is to obtain information on the temperature fields and heat transfer coefficients during flow of liquid-metal coolant in models simulating an elementary cell in the core of a liquid heavy metal cooled fast-neutron reactor. Two design versions for spacing fuel rods in the reactor core were considered. In the first version, the fuel rods were spaced apart from one another using helical wire wound on the fuel rod external surface, and in the second version spacer grids were used for the same purpose. The experiments were carried out on the mercury loop available at the Moscow Power Engineering Institute National Research University's Chair of Engineering Thermal Physics. Two experimental sections simulating an elementary cell for each of the fuel rod spacing versions were fabricated. The temperature fields were investigated using a dedicated hinged probe that allows temperature to be measured at any point of the studied channel cross section. The heat-transfer coefficients were determined using the wall temperature values obtained at the moment when the probe thermocouple tail end touched the channel wall. Such method of determining the wall temperature makes it possible to alleviate errors that are unavoidable in case of measuring the wall temperature using thermocouples placed in slots milled in the wall. In carrying out the experiments, an automated system of scientific research was applied, which allows a large body of data to be obtained within a short period of time. The experimental investigations in the first test section were carried out at Re = 8700, and in the second one, at five values of Reynolds number. Information about temperature fields was obtained by statistically processing the array of sampled probe thermocouple indications at 300 points in the experimental channel cross section. Reach material has been obtained for verifying the codes used for calculating velocity and temperature fields in channels with

Blue phase liquid crystal phase transition for cyano compound chiral nematic liquid crystal mixtures with three two-ring core structures and chiral dopant concentrations

NASA Astrophysics Data System (ADS)

Shin, Jaesun; Kim, Beomjong; Jung, Wansu; Fahad, Mateen; Park, SangJin; Hong, Sung-Kyu

2017-05-01

Blue phase (BP) temperature range of a chiral nematic liquid crystal (LC) mixture is dependent upon the host nematic LC chemical structure and chiral dopant concentration. In this study, we investigated BP phase transition behaviour and helical twisting power (HTP) using three chiral dopant concentrations of cyano compound chiral nematic LC mixtures incorporating three two-ring core structures in the host nematic LCs. The effect of the host nematic LC core structure, HTP and chiral dopant concentrations were considered on BP temperature ranges, for two types of complete BPI and BPII without isotropic phase (Iso) and two types of coexistence state of BPI+Iso and BPII+Iso.

Combustion characteristics in the transition region of liquid fuel sprays

NASA Technical Reports Server (NTRS)

Cernansky, N. P.; Namer, I.; Tidona, R. J.

1986-01-01

A number of important effects have been observed in the droplet size transition region in spray combustion systems. In this region, where the mechanism of flame propagation is transformed from diffusive to premixed dominated combustion, the following effects have been observed: (1) maxima in burning velocity; (2) extension of flammability limits; (3) minima in ignition energy; and (4) minima in NOx formation. A monodisperse aerosol generator has been used to form and deliver a well controlled liquid fuel spray to the combustion test section where measurements of ignition energy have been made. The ignition studies were performed on monodisperse n-heptane sprays at atmospheric pressure over a range of equivalence ratios and droplet diameters. A capacitive discharge spark ignition system was used as the ignition source, providing independent control of spark energy and duration. Preliminary measurements were made to optimize spark duration and spark gap, optimum conditions being those at which the maximum frequency or probability of ignition was observed. Using the optimum electrode spacing and spark duration, the frequency of ignition was determined as a function of spark energy for three overall equivalence ratios (0.6, 0.8, and 1.0) and for initial droplet diameters of 25, 40, 50, 60, and 70 micro m.

Vibrating fuel grapple

DOEpatents

Chertock, deceased, Alan J.; Fox, Jack N.; Weissinger, Robert B.

1982-01-01

A reactor refueling method utilizing a vibrating fuel grapple for removing spent fuel assemblies from a reactor core which incorporates a pneumatic vibrator in the grapple head, enabling additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

Modelling socio-metabolic transitions: The historical take-off, the acceleration of fossil fuel use, and the 1970s oil price shock - the first trigger of a future decline?

NASA Astrophysics Data System (ADS)

Wiedenhofer, Dominik; Rovenskaya, Elena; Krausmann, Fridolin; Haas, Willi; Fischer-Kowalski, Marina

2013-04-01

By talking about socio-metabolic transitions, we talk about changes in the energy base of socio-economic systems, leading to fundamental changes in social and environmental relations. This refers to the historical shift from a biomass-based (agrarian) economy to a fossil fuel based (industrial) economy just as much as to a future shift from fossil fuels to renewable energy carriers. In our presentation, • We will first show that this pattern of transition can be identified for most high income industrial countries: the later the transition started, the faster it proceeded, and the turning point to stabilization of metabolic rates in all of them happened in the early 1970ies. Due to the inherent non-linearity of this process, two approaches will be aplied to estimate parameters for the starting point, transition speed and saturation level: firstly a combination of an expontential and a generalized logistic function and secondly a Gompertz function. For both an iterative test procedure is applied to find the global minimum of the residual error for the whole function and all its parameters. This theory-based approach allows us to apply a robust methodology across all cases, thereby yielding results which can be generalized. • Next, we will show that this was not just a "historical" socio-ecological transition, however. Currently, a substantial number of countries comprising more than half of the world's population are following a similar transitional pathway at an ever accelerating pace. Based on empirical data on physical resource use and the above sketched methodology, we can show that these so-called emerging economies are currently in the take-off or acceleration phase of the very same transition. • Apart from these "endogenous" processes of socio-metabolic transition, we will investigate the effect of external shocks and their impact on the dynamics of energy and materials use. The first such shock we will explore is the oil crisis of 1972 that possibly

Nuclear design of a vapor core reactor for space nuclear propulsion

NASA Astrophysics Data System (ADS)

Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

1993-01-01

Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

American Fuel Cell Bus Project : First Analysis Report

DOT National Transportation Integrated Search

2013-06-01

This report summarizes the experience and early results from the American Fuel Cell Bus Project, a fuel cell electric bus demonstration funded by the Federal Transit Administration (FTA) under the National Fuel Cell Bus Program. A team led by CALSTAR...

American fuel cell bus project : first analysis report.

DOT National Transportation Integrated Search

2013-06-01

This report summarizes the experience and early results from the American Fuel Cell Bus Project, a fuel cell electric bus demonstration : funded by the Federal Transit Administration (FTA) under the National Fuel Cell Bus Program. A team led by CALST...

Stationary Liquid Fuel Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, Won Sik; Grandy, Andrew; Boroski, Andrew

For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excessmore » reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The

Viscosity of the earth's core.

NASA Technical Reports Server (NTRS)

Gans, R. F.

1972-01-01

Calculation of the viscosity of the core at the boundary of the inner and outer core. It is assumed that this boundary is a melting transition and the viscosity limits of the Andrade (1934,1952) hypothesis (3.7 to 18.5 cp) are adopted. The corresponding kinematic viscosities are such that the precessional system explored by Malkus (1968) would be unstable. Whether it would be sufficiently unstable to overcome a severely subadiabatic temperature gradient cannot be determined.

Fuel shipment experience, fuel movements from the BMI-1 transport cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bauer, Thomas L.; Krause, Michael G

1986-07-01

The University of Texas at Austin received two shipments of irradiated fuel elements from Northrup Aircraft Corporation on April 11 and 16, 1985. A total of 59 elements consisting of standard and instrumented TRIGA fuel were unloaded from the BMI-1 shipping cask. At the time of shipment, the Northrup core burnup was approximately 50 megawatt days with fuel element radiation levels, after a cooling time of three months, of approximately 1.75 rem/hr at 3 feet. In order to facilitate future planning of fuel shipment at the UT facility and other facilities, a summary of the recent transfer process including severalmore » factors which contributed to its success are presented. Numerous color slides were made of the process for future reference by UT and others involved in fuel transfer and handling of the BMI-1 cask.« less

Tracking costs of alternatively fueled buses in Florida : [summary].

DOT National Transportation Integrated Search

2011-01-01

In an effort to address rising fuel costs and environmental concerns, many transit agencies across Florida have introduced alternative fuel technologies to their traditional diesel-powered fleets. Fuel types include biodiesel, compressed natural gas,...

Advanced Nuclear Fuel Cycle Transitions: Optimization, Modeling Choices, and Disruptions

NASA Astrophysics Data System (ADS)

Carlsen, Robert W.

Many nuclear fuel cycle simulators have evolved over time to help understan the nuclear industry/ecosystem at a macroscopic level. Cyclus is one of th first fuel cycle simulators to accommodate larger-scale analysis with it liberal open-source licensing and first-class Linux support. Cyclus also ha features that uniquely enable investigating the effects of modeling choices o fuel cycle simulators and scenarios. This work is divided into thre experiments focusing on optimization, effects of modeling choices, and fue cycle uncertainty. Effective optimization techniques are developed for automatically determinin desirable facility deployment schedules with Cyclus. A novel method fo mapping optimization variables to deployment schedules is developed. Thi allows relationships between reactor types and scenario constraints to b represented implicitly in the variable definitions enabling the usage o optimizers lacking constraint support. It also prevents wasting computationa resources evaluating infeasible deployment schedules. Deployed power capacit over time and deployment of non-reactor facilities are also included a optimization variables There are many fuel cycle simulators built with different combinations o modeling choices. Comparing results between them is often difficult. Cyclus flexibility allows comparing effects of many such modeling choices. Reacto refueling cycle synchronization and inter-facility competition among othe effects are compared in four cases each using combinations of fleet of individually modeled reactors with 1-month or 3-month time steps. There are noticeable differences in results for the different cases. The larges differences occur during periods of constrained reactor fuel availability This and similar work can help improve the quality of fuel cycle analysi generally There is significant uncertainty associated deploying new nuclear technologie such as time-frames for technology availability and the cost of buildin advanced reactors

Information Is a Common Core Dish Best Served First

ERIC Educational Resources Information Center

Walkowiak, Temple A.

2015-01-01

There has been a spike in negative comments about mathematics and standards since the implementation of the Common Core State Standards. Regardless of whether the comments are fueled by social media or traditional media, educators need to be armed with strategies for helping parents understand, navigate, and embrace the Common Core's mathematics…

SunLine Transit Agency Hydrogen-Powered Transit Buses : Evaluation Results Update

DOT National Transportation Integrated Search

2007-10-01

In early 2007, the National Renewable Energy Laboratory (NREL) published a preliminary evaluation results report (January 2006 through November 2006) on hydrogen- and CNG-fueled buses operating at SunLine Transit Agency (SunLine) in Thousand Palms, C...

Phase transition of a cobalt-free perovskite as a high-performance cathode for intermediate-temperature solid oxide fuel cells.

PubMed

Jiang, Shanshan; Zhou, Wei; Niu, Yingjie; Zhu, Zhonghua; Shao, Zongping

2012-10-01

It is generally recognized that the phase transition of a perovskite may be detrimental to the connection between cathode and electrolyte. Moreover, certain phase transitions may induce the formation of poor electronic and ionic conducting phase(s), thereby lowering the electrochemical performance of the cathode. Here, we present a study on the phase transition of a cobalt-free perovskite (SrNb(0.1)Fe(0.9)O(3-δ), SNF) and evaluate its effect on the electrochemical performance of the fuel cell. SNF exists as a primitive perovskite structure with space group P4mm (99) at room temperature. As evidenced by in situ high-temperature X-ray diffraction measurements over the temperature range of 600 to 1000 °C, SNF undergoes a transformation to a tetragonal structure with a space group I4/m (87). This phase transition is accompanied by a moderate change in the volume, allowing a good cathode/electrolyte interface on thermal cycling. According to the electrochemical impedance spectroscopy evaluation, the I4/m phase exhibits positive effects on the cathode's performance, showing the highest oxygen reduction reaction activity of cobalt-free cathodes reported so far. This activity improvement is attributed to enhanced oxygen surface processes. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Morphology of single inhalable particle inside public transit biodiesel fueled bus.

PubMed

Shandilya, Kaushik K; Kumar, Ashok

2010-01-01

In an urban-transit bus, fueled by biodiesel in Toledo, Ohio, single inhalable particle samples in October 2008 were collected and detected by scanning electron microscopy and energy dispersive X-ray spectrometry (SEM/EDS). Particle size analysis found bimodal distribution at 0.2 and 0.5 microm. The particle morphology was characterized by 14 different shape clusters: square, pentagon, hexagon, heptagon, octagon, nonagon, decagon, agglomerate, sphere, triangle, oblong, strip, line or stick, and unknown, by quantitative order. The square particles were common in the samples. Round and triangle particles are more, and pentagon, hexagon, heptagon, octagon, nonagon, decagon, strip, line or sticks are less. Agglomerate particles were found in abundance. The surface of most particles was coarse with a fractal edge that can provide a suitable chemical reaction bed in the polluted atmospheric environment. The three sorts of surface patterns of squares were smooth, semi-smooth, and coarse. The three sorts of square surface patterns represented the morphological characteristics of single inhalable particles in the air inside the bus in Toledo. The size and shape distribution results were compared to those obtained for a bus using ultra low sulfur diesel.

Pu-Zr alloy for high-temperature foil-type fuel

DOEpatents

McCuaig, Franklin D.

1977-01-01

A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron reflux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

Dissolution flowsheet for high flux isotope reactor fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Foster, T.

2016-09-27

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3O 8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2. The HFIR fuel cores will be dissolved and the recovered U will be down-blendedmore » into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H 2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H 2 and other permanent gases in the dissolution offgas allowing the development of H 2 generation rate profiles. The H 2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower flammability limit (LFL) for H 2 at a given Hg

Numerical modeling of laboratory-scale surface-to-crown fire transition

NASA Astrophysics Data System (ADS)

Castle, Drew Clayton

Understanding the conditions leading to the transition of fire spread from a surface fuel to an elevated (crown) fuel is critical to effective fire risk assessment and management. Surface fires that successfully transition to crown fires can be very difficult to suppress, potentially leading to damages in the natural and built environments. This is relevant to chaparral shrub lands which are common throughout parts of the Southwest U.S. and represent a significant part of the wildland urban interface. The ability of the Wildland-Urban Interface Fire Dynamic Simulator (WFDS) to model surface-to-crown fire transition was evaluated through comparison to laboratory experiments. The WFDS model is being developed by the U.S. Forest Service (USFS) and the National Institute of Standards and Technology. The experiments were conducted at the USFS Forest Fire Laboratory in Riverside, California. The experiments measured the ignition of chamise (Adenostoma fasciculatum) crown fuel held above a surface fire spreading through excelsior fuel. Cases with different crown fuel bulk densities, crown fuel base heights, and imposed wind speeds were considered. Cold-flow simulations yielded wind speed profiles that closely matched the experimental measurements. Next, fire simulations with only the surface fuel were conducted to verify the rate of spread while factors such as substrate properties were varied. Finally, simulations with both a surface fuel and a crown fuel were completed. Examination of specific surface fire characteristics (rate of spread, flame angle, etc.) and the corresponding experimental surface fire behavior provided a basis for comparison of the factors most responsible for transition from a surface fire to the raised fuel ignition. The rate of spread was determined by tracking the flame in the Smokeview animations using a tool developed for tracking an actual flame in a video. WFDS simulations produced results in both surface fire spread and raised fuel bed

Moving, Moving, Moving- A Giant Rocket Fuel Tank

NASA Image and Video Library

2016-10-07

Technicians moved a giant fuel tank from the Vertical Assembly Center where the tank recently completed friction stir welding to an adjacent work area at NASA's Michoud Assembly Facility in New Orleans. More than 1.7 miles of welds have been completed for core stage hardware at Michoud. This liquid hydrogen fuel tank is the largest piece of the core stage that will provide the fuel for the first flight of NASA's new rocket, the Space Launch System, with the Orion spacecraft in 2018. The tank is more than 130 feet long, and together with the liquid oxygen tank holds 733,000 gallons of propellant to feed the vehicle's four RS-25 engines to produce a total of 2 million pounds of thrust. SLS will have the power and capacity to carry humans to Mars. For more information on the core stage: http://www.nasa.gov/exploration/syste... Video Credit: NASA/MAF/Eric Bordelon

PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

NASA Astrophysics Data System (ADS)

Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

2014-06-01

The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

Soldatov, Alexey I.

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

Fuel cell development for transportation: Catalyst development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doddapaneni, N.

1996-04-01

Fuel cells are being considered as alternate power sources for transportation and stationary applications. With proton exchange membrane (PEM) fuel cells the fuel crossover to cathodes causes severe thermal management and cell voltage drop due to oxidation of fuel at the platinized cathodes. The main goal of this project was to design, synthesize, and evaluate stable and inexpensive transition metal macrocyclic catalysts for the reduction of oxygen and be electrochemically inert towards anode fuels such as hydrogen and methanol.

Molten salt considerations for accelerator-driven subcritical fission to close the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Sooby, Elizabeth; Adams, Marvin; Baty, Austin; Gerity, James; McIntyre, Peter; Melconian, Karie; Phongikaroon, Supathorn; Pogue, Nathaniel; Sattarov, Akhdiyor; Simpson, Michael; Tripathy, Prabhat; Tsevkov, Pavel

2013-04-01

The host salt selection, molecular modeling, physical chemistry, and processing chemistry are presented here for an accelerator-driven subcritical fission in a molten salt core (ADSMS). The core is fueled solely with the transuranics (TRU) and long-lived fission products (LFP) from used nuclear fuel. The neutronics and salt composition are optimized to destroy the transuranics by fission and the long-lived fission products by transmutation. The cores are driven by proton beams from a strong-focusing cyclotron stack. One such ADSMS system can destroy the transuranics in the used nuclear fuel produced by a 1GWe conventional reactor. It uniquely provides a method to close the nuclear fuel cycle for green nuclear energy.

Alternative Fuels Data Center: Worcester Regional Transit Authority Drives

Science.gov Websites

Clean Cities. Download QuickTime Video QuickTime (.mov) Download Windows Media Video Windows Media (.wmv ) Video Download Help Text version See more videos provided by Clean Cities TV and FuelEconomy.gov Fuel Cell Vehicles in California Nov. 18, 2017 Photo of a car Smart Car Shopping Nov. 4, 2017 Image of

Transit Bus Fuel Economy and Performance Simulation

DOT National Transportation Integrated Search

1984-01-01

This report presents the results of bus simulation studies to determine the effects of various design and operating parameters on bus fuel economy and performance. The bus components are first described in terms of how they are modeled. Then a variat...

Composite fuel cell membranes

DOEpatents

Plowman, K.R.; Rehg, T.J.; Davis, L.W.; Carl, W.P.; Cisar, A.J.; Eastland, C.S.

1997-08-05

A bilayer or trilayer composite ion exchange membrane is described suitable for use in a fuel cell. The composite membrane has a high equivalent weight thick layer in order to provide sufficient strength and low equivalent weight surface layers for improved electrical performance in a fuel cell. In use, the composite membrane is provided with electrode surface layers. The composite membrane can be composed of a sulfonic fluoropolymer in both core and surface layers.

Composite fuel cell membranes

DOEpatents

Plowman, Keith R.; Rehg, Timothy J.; Davis, Larry W.; Carl, William P.; Cisar, Alan J.; Eastland, Charles S.

1997-01-01

A bilayer or trilayer composite ion exchange membrane suitable for use in a fuel cell. The composite membrane has a high equivalent weight thick layer in order to provide sufficient strength and low equivalent weight surface layers for improved electrical performance in a fuel cell. In use, the composite membrane is provided with electrode surface layers. The composite membrane can be composed of a sulfonic fluoropolymer in both core and surface layers.

77 FR 72746 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-06

...EPA published a direct final rule on October 9, 2012 to amend the definition of heating oil in 40 CFR 80.1401 in the Renewable Fuel Standard (``RFS'') program under section 211(o) of the Clean Air Act. The direct final rule also amended requirements under EPA's diesel sulfur program related to the sulfur content of locomotive and marine diesel fuel produced by transmix processors, and the fuel marker requirements for 500 ppm sulfur locomotive and marine (LM) diesel fuel to allow for solvent yellow 124 marker to transition out of the distribution system. Because EPA received adverse comments on the heating oil definition and transmix amendments, we are withdrawing those portions of the direct final rule. Because EPA did not receive adverse comments with respect to the yellow marker amendments, those amendments will become effective as indicated in the direct final rule.

Gaseous fuel reactors for power systems

NASA Technical Reports Server (NTRS)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

40 CFR 86.113-15 - Fuel specifications.

Code of Federal Regulations, 2014 CFR

2014-07-01

... Heavy-Duty Vehicles; Test Procedures § 86.113-15 Fuel specifications. Section 86.113-15 includes text... transition to an ethanol-blend test fuel for vehicles certified under subpart S of this part. You may use the test fuels specified in § 86.113-04(a) for vehicles that are not yet subject to testing with the new...

A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2013-01-01

This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

Atomic Migration Induced Crystal Structure Transformation and Core-Centered Phase Transition in Single Crystal Ge2Sb2Te5 Nanowires.

PubMed

Lee, Jun-Young; Kim, Jeong-Hyeon; Jeon, Deok-Jin; Han, Jaehyun; Yeo, Jong-Souk

2016-10-12

A phase change nanowire holds a promise for nonvolatile memory applications, but its transition mechanism has remained unclear due to the analytical difficulties at atomic resolution. Here we obtain a deeper understanding on the phase transition of a single crystalline Ge 2 Sb 2 Te 5 nanowire (GST NW) using atomic scale imaging, diffraction, and chemical analysis. Our cross-sectional analysis has shown that the as-grown hexagonal close-packed structure of the single crystal GST NW transforms to a metastable face-centered cubic structure due to the atomic migration to the pre-existing vacancy layers in the hcp structure going through iterative electrical switching. We call this crystal structure transformation "metastabilization", which is also confirmed by the increase of set-resistance during the switching operation. For the set to reset transition between crystalline and amorphous phases, high-resolution imaging indicates that the longitudinal center of the nanowire mainly undergoes phase transition. According to the atomic scale analysis of the GST NW after repeated electrical switching, partial crystallites are distributed around the core-centered amorphous region of the nanowire where atomic migration is mainly induced, thus potentially leading to low power electrical switching. These results provide a novel understanding of phase change nanowires, and can be applied to enhance the design of nanowire phase change memory devices for improved electrical performance.

Heat exchanger for fuel cell power plant reformer

DOEpatents

Misage, Robert; Scheffler, Glenn W.; Setzer, Herbert J.; Margiott, Paul R.; Parenti, Jr., Edmund K.

1988-01-01

A heat exchanger uses the heat from processed fuel gas from a reformer for a fuel cell to superheat steam, to preheat raw fuel prior to entering the reformer and to heat a water-steam coolant mixture from the fuel cells. The processed fuel gas temperature is thus lowered to a level useful in the fuel cell reaction. The four temperature adjustments are accomplished in a single heat exchanger with only three heat transfer cores. The heat exchanger is preheated by circulating coolant and purge steam from the power section during startup of the latter.

An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

NASA Technical Reports Server (NTRS)

Clark, John S.; Walton, James T.; Mcguire, Melissa L.

1992-01-01

Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines.

Evidence for core 2 to core 1 O-glycan remodeling during the recycling of MUC1

PubMed Central

Razawi, Hanieh; Kinlough, Carol L; Staubach, Simon; Poland, Paul A; Rbaibi, Youssef; Weisz, Ora A; Hughey, Rebecca P; Hanisch, Franz-Georg

2013-01-01

The apical transmembrane glycoprotein MUC1 is endocytosed to recycle through the trans-Golgi network (TGN) or Golgi complex to the plasma membrane. We followed the hypothesis that not only the known follow-up sialylation of MUC1 in the TGN is associated with this process, but also a remodeling of O-glycan core structures, which would explain the previously described differential core 2- vs core 1-based O-glycosylation of secreted, single Golgi passage and recycling membrane MUC1 isoforms (Engelmann K, Kinlough CL, Müller S, Razawi H, Baldus SE, Hughey RP, Hanisch F-G. 2005. Glycobiology. 15:1111–1124). Transmembrane and secreted MUC1 probes show trafficking-dependent changes in O-glycan core profiles. To address this novel observation, we used recombinant epitope-tagged MUC1 (MUC1-M) and mutant forms with abrogated clathrin-mediated endocytosis (MUC1-M-Y20,60N) or blocked recycling (palmitoylation-defective MUC1-M-CQC/AQA). We show that the CQC/AQA mutant transits the TGN at significantly lower levels, concomitant with a strongly reduced shedding from the plasma membrane and its accumulation in endosomal compartments. Intriguingly, the O-glycosylation of the shed MUC1 ectodomain subunit changes from preponderant sialylated core 1 (MUC1-M) to core 2 glycans on the non-recycling CQC/AQA mutant. The O-glycoprofile of the non-recycling CQC/AQA mutant resembles the core 2 glycoprofile on a secretory MUC1 probe that transits the Golgi complex only once. In contrast, the MUC1-M-Y20,60N mutant recycles via flotillin-dependent pathways and shows the wild-type phenotype with dominant core 1 expression. Differential radiolabeling of protein with [35S]Met/Cys or glycans with [3H]GlcNH2 in pulse-chase experiments of surface biotinylated MUC1 revealed a significantly shorter half-life of [3H]MUC1 when compared with [35S]MUC1, whereas the same ratio for the CQC/AQA mutant was close to one. This finding further supports the novel possibility of a recycling-associated O

Elastic properties and phase transitions of Fe7C3 and new constraints on the light element budget of the Earth's inner core

NASA Astrophysics Data System (ADS)

Prescher, C.; Bykova, E.; Kupenko, I.; Glazyrin, K.; Kantor, A.; McCammon, C. A.; Mookherjee, M.; Miyajima, N.; Cerantola, V.; Nakajima, Y.; Prakapenka, V.; Rüffer, R.; Chumakov, A.; Dubrovinsky, L. S.

2013-12-01

The Earth's inner core consists mainly of iron (or iron-nickel alloy) with some amount of light element(s) whereby their nature remains controversial. Seismological data suggest that the material forming Earth's inner core (pressures over 330 GPa and temperatures above 5000 K) has an enigmatically high Poisson's ratio ~0.44, while iron or it alloys with Si, S, O, or H expected to have at appropriate thermodynamic conditions Poisson's ratio well below 0.39. We will present an experimental study on a new high pressure variant in the iron carbide system. We have synthesized and solved structure of high-pressure orthorhombic phase of o-Fe7C3, and investigated its stability and behavior at pressures over 180 GPa and temperatures above 3500 K by means of different methods including single crystal X-ray diffraction, Mössbauer spectroscopy, and nuclear resonance scattering. O-Fe7C3 is structurally stable to at least outer core conditions and demonstrates magnetic or electronic transitions at ~18 GPa and ~70 GPa. The high pressure phase of o-Fe7C3 above 70 GPa exhibits anomalous elastic properties. When extrapolated to the conditions of the Earth's inner core it shows shear wave velocities and Poisson's ratios close to the values inferred by seismological models. Our results not only support earlier works suggesting that carbon may be an important component of Earth's core, but shows that it may drastically change iron's elastic properties, thus explaining anomalous Earth's inner core elastic properties.

76 FR 1213 - Core Principles and Other Requirements for Swap Execution Facilities

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-07

... Part II Commodity Futures Trading Commission 17 CFR Part 37 Core Principles and Other Requirements... RIN Number 3038-AD18 Core Principles and Other Requirements for Swap Execution Facilities AGENCY... Compliance With the Core Principles III. Effective Date and Transition Period IV. Related Matters A...

Pulsed Magnetic Welding for Advanced Core and Cladding Steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cao, Guoping; Yang, Yong

2013-12-19

To investigate a solid-state joining method, pulsed magnetic welding (PMW), for welding the advanced core and cladding steels to be used in Generation IV systems, with a specific application for fuel pin end-plug welding. As another alternative solid state welding technique, pulsed magnetic welding (PMW) has not been extensively explored on the advanced steels. The resultant weld can be free from microstructure defects (pores, non-metallic inclusions, segregation of alloying elements). More specifically, the following objectives are to be achieved: 1. To design a suitable welding apparatus fixture, and optimize welding parameters for repeatable and acceptable joining of the fuel pinmore » end-plug. The welding will be evaluated using tensile tests for lap joint weldments and helium leak tests for the fuel pin end-plug; 2 Investigate the microstructural and mechanical properties changes in PMW weldments of proposed advanced core and cladding alloys; 3. Simulate the irradiation effects on the PWM weldments using ion irradiation.« less

Overview of Fuel Rod Simulator Usage at ORNL

NASA Astrophysics Data System (ADS)

Ott, Larry J.; McCulloch, Reg

2004-02-01

During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

Transitional Gas Jet Diffusion Flames in Microgravity

NASA Technical Reports Server (NTRS)

Agrawal, Ajay K.; Alammar, Khalid; Gollahalli, S. R.; Griffin, DeVon (Technical Monitor)

2000-01-01

Drop tower experiments were performed to identify buoyancy effects in transitional hydrogen gas jet diffusion flames. Quantitative rainbow schlieren deflectometry was utilized to optically visualize the flame and to measure oxygen concentration in the laminar portion of the flame. Test conditions consisted of atmospheric pressure flames burning in quiescent air. Fuel from a 0.3mm inside diameter tube injector was issued at jet exit Reynolds numbers (Re) of 1300 to 1700. Helium mole percentage in the fuel was varied from 0 to 40%. Significant effects of buoyancy were observed in near field of the flame even-though the fuel jets were momentum-dominated. Results show an increase of breakpoint length in microgravity. Data suggest that transitional flames in earth-gravity at Re<1300 might become laminar in microgravity.

American Fuel Cell Bus Project Evaluation. Second Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eudy, Leslie; Post, Matthew

2015-09-01

This report presents results of the American Fuel Cell Bus (AFCB) Project, a demonstration of fuel cell electric buses operating in the Coachella Valley area of California. The prototype AFCB was developed as part of the Federal Transit Administration's (FTA's) National Fuel Cell Bus Program. Through the non-profit consortia CALSTART, a team led by SunLine Transit Agency and BAE Systems developed a new fuel cell electric bus for demonstration. SunLine added two more AFCBs to its fleet in 2014 and another in 2015. FTA and the AFCB project team are collaborating with the U.S. Department of Energy (DOE) and DOE'smore » National Renewable Energy Laboratory to evaluate the buses in revenue service. This report summarizes the performance results for the buses through June 2015.« less

Pm-1 Reactor Core Final Design Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bagley, R. O.; Cox, F. H.; Carnasale, A.

1962-01-01

The PM-1 water cooled and moderated core contains 741 highly enriched stainless steel cermet tubular fuel elements and 90 lumped B stainless steel burnable poison elements, and it is controlled by 6 Y-shaped europium titanate movable control rods. The core has a lifetime of 1.95 years when operated at its design power level of 9.37 mw of thermal energy. The control of the core is designed so that there is a positive shutdown margin at all times with either one rod stuck completely out or the core or with two rods stuck in the operating condition. The core power ismore » removed by 2125 gpm of pressurized water at an average temperature of 463 deg F and pressure of 1300 psia. In reactors of this type, the core is stable with a negative temperature coefficient of approximately 2.5 x 10/sup -4/ DELTA K/K/ deg F.« less

IMPROVEMENTS RELATING TO NUCLEAR REACTOR CORE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bell, F.R.

1963-03-01

A nuclear reactor core composed of a number of stacked horizontal layers is described. Each layer is made up of elements of moderator material of equal height and of generally hexagonal cross-section. Each element has holes containing nuclear fuel and separate ones for coolant. (C.E.S.)

Core power and decay time limits for a disabled LOFT ECCS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Atkinson, S.A.

1978-01-09

An analysis was done to determine at what LOFT total core power (nuclear plus decay power) the ECCS could be inoperable. The criteria used for the analysis was that the maximum fuel clad temperature should not exceed 1650/sup 0/F given a loss of coolant. Calculations for natural convection cooling of the fuel by air with an inlet temperature of 580/sup 0/F determined that the limiting core power is 25 kW (discounted by 15 percent to 20 percent for potential uncertainties). Shutdown times are listed for when the LOFT ECCS can be safely bypassed or disabled.

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

NASA Astrophysics Data System (ADS)

Alekseev, P. N.; Bobrov, E. A.; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A.

2015-12-01

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U-Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium-plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: 235U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or 233U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

Core cooling under accident conditions at the high flux beam reactor (HFBR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tichler, P.; Cheng, L.; Fauske, H.

In certain accident scenarios, e.g. loss of coolant accidents (LOCA) all forced flow cooling is lost. Decay heating causes a temperature increase in the core coolant and the resulting thermal buoyancy causes a reversal of the flow direction to a natural circulation mode. Although there was experimental evidence during the reactor design period (1958--1963) that the heat removal capacity in the fully developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. In a LOCA scenario where even limited fuelmore » damage occurs and natural circulation is established, fission product gases could be carried from the damaged fuel by steam into areas where operator access is required to maintain the core in a coolable configuration. This would force evacuation of the building and lead to extensive core damage. As a result the HFBR was shut down by the Department of Energy (DOE) and an extensive review of the HFBR was initiated. In an effort to address this issue BNL developed a model designed to predict the heat removal limit during flow reversal that was found to be in good agreement with the test results. Currently a thermal-hydraulic test program is being developed to provide a more realistic and defensible estimate of the flow reversal heat removal limit so that the reactor power level can be increased.« less

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabricationmore » must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.« less

A sharp and flat section of the core-mantle boundary

USGS Publications Warehouse

Vidale, J.E.; Benz, H.M.

1992-01-01

THE transition zone between the Earth's core and mantle plays an important role as a boundary layer for mantle and core convection1. This zone conducts a large amount of heat from the core to the mantle, and contains at least one thermal boundary layer2,3; the proximity of reactive silicates and molten iron leads to the possibility of zones of intermediate composition4. Here we investigate one region of the core-mantle boundary using seismic waves that are converted from shear to compressional waves by reflection at the boundary. The use of this phase (known as ScP), the large number of receiving stations, and the large aperture of our array all provide higher resolution than has previously been possible5-7. For the 350-km-long section of the core-mantle boundary under the northeast Pacific sampled by the reflections, the local boundary topography has an amplitude of less than 500 m, no sharp radial gradients exist in the 400 km above the boundary, and the mantle-lo-core transition occurs over less than 1 km. The simplicity of the structure near and above the core-mantle boundary argues against chemical heterogeneity at the base of the mantle in this location.

Connecting Core Percolation and Controllability of Complex Networks

PubMed Central

Jia, Tao; Pósfai, Márton

2014-01-01

Core percolation is a fundamental structural transition in complex networks related to a wide range of important problems. Recent advances have provided us an analytical framework of core percolation in uncorrelated random networks with arbitrary degree distributions. Here we apply the tools in analysis of network controllability. We confirm analytically that the emergence of the bifurcation in control coincides with the formation of the core and the structure of the core determines the control mode of the network. We also derive the analytical expression related to the controllability robustness by extending the deduction in core percolation. These findings help us better understand the interesting interplay between the structural and dynamical properties of complex networks. PMID:24946797

Vibrating fuel grapple. [LMFBR

DOEpatents

Chertock, A.J.; Fox, J.N.; Weissinger, R.B.

A reactor refueling method is described which utilizes a vibrating fuel grapple for removing spent fuel assemblies from a reactor core. It incorporates a pneumatic vibrator in the grapple head which allows additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

Gaseous fuel nuclear reactor research

NASA Technical Reports Server (NTRS)

Schwenk, F. C.; Thom, K.

1975-01-01

Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

Apparatus and method for controlling the secondary injection of fuel

DOEpatents

Martin, Scott M.; Cai, Weidong; Harris, Jr., Arthur J.

2013-03-05

A combustor (28) for a gas turbine engine is provided comprising a primary combustion chamber (30) for combusting a first fuel to form a combustion flow stream (50) and a transition piece (32) located downstream from the primary combustion chamber (30). The transition piece (32) comprises a plurality of injectors (66) located around a circumference of the transition piece (32) for injecting a second fuel into the combustion flow stream (50). The injectors (66) are effective to create a radial temperature profile (74) at an exit (58) of the transition piece (32) having a reduced coefficient of variation relative to a radial temperature profile (64) at an inlet (54) of the transition piece (32). Methods for controlling the temperature profile of a secondary injection are also provided.

Investigation, quantification, and recommendations : performance of alternatively fueled buses : [summary].

DOT National Transportation Integrated Search

2014-08-01

Volatile fuel prices and the need to reduce carbon emissions have caused many U.S. transit agencies to invest in vehicles that use alternative fuels. These vehicles have different patterns of costs than traditional, diesel-fueled vehicles, and the Fl...

Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N.

2012-07-01

The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existingmore » facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)« less

Nuclear reactor fuel element with vanadium getter on cladding

DOEpatents

Johnson, Carl E.; Carroll, Kenneth G.

1977-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

Determining Reactor Fuel Type from Continuous Antineutrino Monitoring

NASA Astrophysics Data System (ADS)

Jaffke, Patrick; Huber, Patrick

2017-09-01

We investigate the ability of an antineutrino detector to determine the fuel type of a reactor. A hypothetical 5-ton antineutrino detector is placed 25 m from the core and measures the spectral shape and rate of antineutrinos emitted by fission fragments in the core for a number of 90-d periods. Our results indicate that four major fuel types can be differentiated from the variation of fission fractions over the irradiation time with a true positive probability of detection at approximately 95%. In addition, we demonstrate that antineutrinos can identify the burnup at which weapons-grade mixed-oxide (MOX) fuel would be reduced to reactor-grade MOX, on average, providing assurance that plutonium-disposition goals are met. We also investigate removal scenarios where plutonium is purposefully diverted from a mixture of MOX and low-enriched uranium fuel. Finally, we discuss how our analysis is impacted by a spectral distortion around 6 MeV observed in the antineutrino spectrum measured from commercial power reactors.

Analysis of maximum allowable fragment heights during dissolution of high flux isotope reactor fuel in an h-canyon dissolver

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel, G.; Rudisill, T.

2017-07-17

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3O 8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2. The HFIR fuel cores will be dissolved using a flowsheet developed by the Savannahmore » River National Laboratory (SRNL) in either the 6.4D or 6.1D dissolver using a unique insert. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The recovered U will be down-blended into low-enriched U for subsequent use as commercial reactor fuel. During the development of the HFIR fuel dissolution flowsheet, the cycle time for the initial core was estimated at 28 to 40 h. Once the cycle is complete, H-Canyon personnel will open the dissolver and probe the HFIR insert wells to determine the height of any fuel fragments which did not dissolve. Before the next core can be charged to the dissolver, an analysis of the potential for H 2 gas generation must show that the combined surface area of the fuel fragments and the subsequent core will not generate H 2 concentrations in the dissolver offgas which exceeds 60% of the lower flammability limit (LFL) of H 2 at 200 °C. The objective of this study is to identify the maximum fuel fragment height as a function of the Al concentration in the dissolving solution which will provide criteria for charging successive HFIR cores to an H-Canyon dissolver.« less

Transitional justice as social control: political transitions, human rights norms and the reclassification of the past.

PubMed

Dudai, Ron

2017-09-12

This article offers an interpretation of transitional justice policies - the efforts of post-conflict and post-dictatorship societies to address the legacy of past abuses - as a form of social control. While transitional justice is commonly conceptualized as responding to a core problem of impunity, this article argues that such formulation is too narrow and leads to lack of coherence in the analysis of the diverse array of transitional mechanisms, which include among others trials, truth commissions, reparations for victims and apologies. Building on the work of Stanley Cohen, the article contends that the core transitional problem is the denial of human rights violations, and consequently that the common purpose of all transitional justice mechanisms is to reclassify the past: redefining as deviant some acts and individuals which prior to the transition were considered 'normal'. The article identifies and analyses three themes in the application of a social control framework to transitional justice: (1) truth, memory and retroactive social control, pertains to the way truth-seeking transitional justice mechanisms reclassify past events by engaging in social control of and through memory; (2) censure, celebration and transitional social control refers to the reclassification of categories of individuals through expressions of both social disapproval and praise; and (3) civil society and social control from below concerns the role of social movements, organizations and groups as informal agents of social control during transitions. The concluding section recaps and briefly explores the concept of 'good moral panic' in the context of political transitions. While the concept of social control tends to have negative connotations for critical sociologists, this work suggests that efforts to categorize, punish and disapprove certain behaviours as deviant may not only be viewed as supporting a conservative status-quo, but also as promoting fledging human rights norms. �

Simulating the minimum core for hydrophobic collapse in globular proteins.

PubMed Central

Tsai, J.; Gerstein, M.; Levitt, M.

1997-01-01

To investigate the nature of hydrophobic collapse considered to be the driving force in protein folding, we have simulated aqueous solutions of two model hydrophobic solutes, methane and isobutylene. Using a novel methodology for determining contacts, we can precisely follow hydrophobic aggregation as it proceeds through three stages: dispersed, transition, and collapsed. Theoretical modeling of the cluster formation observed by simulation indicates that this aggregation is cooperative and that the simulations favor the formation of a single cluster midway through the transition stage. This defines a minimum solute hydrophobic core volume. We compare this with protein hydrophobic core volumes determined from solved crystal structures. Our analysis shows that the solute core volume roughly estimates the minimum core size required for independent hydrophobic stabilization of a protein and defines a limiting concentration of nonpolar residues that can cause hydrophobic collapse. These results suggest that the physical forces driving aggregation of hydrophobic molecules in water is indeed responsible for protein folding. PMID:9416609

Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

DOE PAGES

Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

2016-11-18

Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies ismore » a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium

Design of a fuel element for a lead-cooled fast reactor

NASA Astrophysics Data System (ADS)

Sobolev, V.; Malambu, E.; Abderrahim, H. Aït

2009-03-01

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg-1 of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg-1 of HM.

Benchmark Evaluation of Fuel Effect and Material Worth Measurements for a Beryllium-Reflected Space Reactor Mockup

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, Margaret A.; Bess, John D.

2015-02-01

The critical configuration of the small, compact critical assembly (SCCA) experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) in 1962-1965 have been evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The initial intent of these experiments was to support the design of the Medium Power Reactor Experiment (MPRE) program, whose purpose was to study “power plants for the production of electrical power in space vehicles.” The third configuration in this series of experiments was a beryllium-reflected assembly of stainless-steel-clad, highly enriched uranium (HEU)-O 2 fuel mockup of a potassium-cooledmore » space power reactor. Reactivity measurements cadmium ratio spectral measurements and fission rate measurements were measured through the core and top reflector. Fuel effect worth measurements and neutron moderating and absorbing material worths were also measured in the assembly fuel region. The cadmium ratios, fission rate, and worth measurements were evaluated for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The fuel tube effect and neutron moderating and absorbing material worth measurements are the focus of this paper. Additionally, a measurement of the worth of potassium filling the core region was performed but has not yet been evaluated Pellets of 93.15 wt.% enriched uranium dioxide (UO 2) were stacked in 30.48 cm tall stainless steel fuel tubes (0.3 cm tall end caps). Each fuel tube had 26 pellets with a total mass of 295.8 g UO 2 per tube. 253 tubes were arranged in 1.506-cm triangular lattice. An additional 7-tube cluster critical configuration was also measured but not used for any physics measurements. The core was surrounded on all side by a beryllium reflector. The fuel effect worths were measured by removing fuel tubes at various radius. An accident scenario was also simulated by

Transition-Metal Nitride Core@Noble-Metal Shell Nanoparticles as Highly CO Tolerant Catalysts

DOE PAGES

Garg, Aaron; Milina, Maria; Ball, Madelyn; ...

2017-05-25

Core–shell architectures offer an effective way to tune and enhance the properties of noble-metal catalysts. Herein, we demonstrate the synthesis of Pt shell on titanium tungsten nitride core nanoparticles (Pt/TiWN) by high temperature ammonia nitridation of a parent core–shell carbide material (Pt/TiWC). X-ray photoelectron spectroscopy revealed significant core-level shifts for Pt shells supported on TiWN cores, corresponding to increased stabilization of the Pt valence d-states. The modulation of the electronic structure of the Pt shell by the nitride core translated into enhanced CO tolerance during hydrogen electrooxidation in the presence of CO. In conclusion, the ability to control shell coveragemore » and vary the heterometallic composition of the shell and nitride core opens up attractive opportunities to synthesize a broad range of new materials with tunable catalytic properties.« less

Transition-Metal Nitride Core@Noble-Metal Shell Nanoparticles as Highly CO Tolerant Catalysts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garg, Aaron; Milina, Maria; Ball, Madelyn

Core–shell architectures offer an effective way to tune and enhance the properties of noble-metal catalysts. Herein, we demonstrate the synthesis of Pt shell on titanium tungsten nitride core nanoparticles (Pt/TiWN) by high temperature ammonia nitridation of a parent core–shell carbide material (Pt/TiWC). X-ray photoelectron spectroscopy revealed significant core-level shifts for Pt shells supported on TiWN cores, corresponding to increased stabilization of the Pt valence d-states. The modulation of the electronic structure of the Pt shell by the nitride core translated into enhanced CO tolerance during hydrogen electrooxidation in the presence of CO. In conclusion, the ability to control shell coveragemore » and vary the heterometallic composition of the shell and nitride core opens up attractive opportunities to synthesize a broad range of new materials with tunable catalytic properties.« less

American Fuel Cell Bus Project Evaluation : Second Report

DOT National Transportation Integrated Search

2015-09-01

This report presents results of the American Fuel Cell Bus (AFCB) Project, a demonstration of fuel cell electric buses (FCEB) operating in the Coachella Valley area of California. The prototype AFCB was developed as part of the Federal Transit Admini...

Evolution of oceanic core complex domes and corrugations

NASA Astrophysics Data System (ADS)

Cann, J.; Escartin, J.; Smith, D.; Schouten, H.

2007-12-01

In regions of the oceans where detachment faulting is developed widely, individual core complex domes (elevated massifs capped by corrugated detachment surfaces) show a consistent morphology. At their outward sides, most core complex domes are attached to a planar slope, interpreted (Smith et al., 2006) as an originally steep inward-facing normal fault that has been rotated to shallower angles. We suggest that the break in slope where the originally steep normal fault meets the domal corrugated surface marks the trace of the brittle-ductile transition at the base of the original normal fault. The steep faults originate within a short distance of the spreading axis. This means that the arcuate shape of the intersection of the steep fault with the dome must indicate the shape of the brittle-ductile transition very close to the spreading axis. The transition must be very shallow close to the summit of the dome and deeper on each flank. Evidence from drilling of some core complexes (McCaig et al, 2007) shows that while the domal detachment faults are active they may channel hydrothermal flow at black smoker temperatures and may be simultaneously injected by magma from below. This indicates a close link between igneous activity, hydrothermal flow and deformation while a core complex is forming. Once the shape of the core complex dome is established, it persists as the ductile footwall mantle rising from below is shaped by the overlying brittle hanging wall that has been cooled by the hydrothermal circulation. The corrugations in the footwall must be moulded into it by irregularities in the brittle hanging wall, as suggested by Spencer (1999). The along-axis arched shape of the hanging wall helps to stabilise the domal shape of the footwall as it rises and cools.

Regular transition zone biopsy during active surveillance for prostate cancer may improve detection of pathological progression.

PubMed

Wong, Lih-Ming; Toi, Ants; Van der Kwast, Theodorus; Trottier, Greg; Alibhai, Shabbir M H; Timilshina, Narhari; Evans, Andrew; Zlotta, Alexandre; Fleshner, Neil; Finelli, Antonio

2014-10-01

We investigated the frequency of cancer and pathological progression in transition zone biopsies in men undergoing multiple rebiopsies while on active surveillance. Eligibility criteria of the active surveillance prostate cancer database (1997 to 2012) at our tertiary center includes prostate specific antigen 10 ng/ml or less, cT2 or less, no Gleason grade 4 or 5, 3 or fewer positive cores, no core with greater than 50% involvement, patient age 75 years or less and 1 or more biopsies after initial diagnostic biopsy. We excluded from analysis men with fewer than 10 cores at diagnostic biopsy and/or confirmatory biopsy greater than 24 months after diagnostic biopsy. Multiparametric magnetic resonance imaging was performed selectively to investigate incongruity between prostate specific antigen and biopsy findings. Pathological progression was defined by grade and/or volume (greater than 50% of core involved). Transition zone progression was subdivided into exclusively transition zone and combined transition zone (transition and peripheral zones). A multivariate Cox proportional hazards model was used to determine predictors of transition zone progression. A total of 392 men were considered in analysis. Median followup was 45.5 months. At each biopsy during active surveillance (confirmatory biopsy to biopsy 5+) there were transition zone positive cores in 18.6% to 26.7% of cases, all transition zone progression in 5.9% to 11.1% and exclusively transition zone progression in 2.7% to 6.7%. Volume related progression was noted more frequently than grade related progression (24 vs 9 cases). Predictors of only transition zone progression were the maximum percent in a single core (HR 1.99, 95% CI 1.30-3.04, p = 0.002) and cancer on magnetic resonance imaging (HR 3.19, 95% CI 1.23-8.27, p = 0.02). Across multiple active surveillance biopsies 2.7% to 6.7% of men had only transition zone progression. We recommend that transition zone biopsy be considered in all men at

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Alternative Fuel Transit Bus Evaluation Program Results

DOT National Transportation Integrated Search

1996-05-06

The objective of this program, which is supported by the U.S. Department of : Energy (DOE) through the National Renewable Energy Laboratory (NREL), is to : provide an unbiased and comprehensive comparison of transit buses operating on : alternative f...

Common Core Literacy Lesson Plans: Ready-to-Use Resources, 9-12

ERIC Educational Resources Information Center

Davis, Lauren

2013-01-01

Schools nationwide are transitioning to the Common Core--our advice to you: Be prepared, but don't go it alone! Our new book, "Common Core Literacy Lesson Plans: Ready-to-Use Resources, 9-12," shows you that teaching the Common Core State Standards in high school doesn't have to be intimidating! This easy-to-use guide meets the particular needs of…

Common Core Literacy Lesson Plans: Ready-to-Use Resources, 6-8

ERIC Educational Resources Information Center

Davis, Lauren

2013-01-01

Schools nationwide are transitioning to the Common Core--our advice to you: Don't go it alone! Our new book, "Common Core Literacy Lesson Plans: Ready-to-Use Resources, 6-8," shows you that teaching the Common Core State Standards in the middle grades doesn't have to be intimidating! This easy-to-use guide meets the particular needs of middle…

Common Core Literacy Lesson Plans: Ready-to-Use Resources, K-5

ERIC Educational Resources Information Center

Davis, Lauren

2013-01-01

Schools nationwide are transitioning to the Common Core--our advice to you: Don't go it alone! Our new book, "Common Core Literacy Lesson Plans: Ready-to-Use Resources, K-5," shows you that teaching the Common Core State Standards in the elementary grades doesn't have to be intimidating! This easy-to-use guide provides model lesson plans for…

Optimisation of composite metallic fuel for minor actinide transmutation in an accelerator-driven system

NASA Astrophysics Data System (ADS)

Uyttenhove, W.; Sobolev, V.; Maschek, W.

2011-09-01

A potential option for neutralization of minor actinides (MA) accumulated in spent nuclear fuel of light water reactors (LWRs) is their transmutation in dedicated accelerator-driven systems (ADS). A promising fuel candidate dedicated to MA transmutation is a CERMET composite with Mo metal matrix and (Pu, Np, Am, Cm)O 2-x fuel particles. Results of optimisation studies of the CERMET fuel targeting to increasing the MA transmutation efficiency of the EFIT (European Facility for Industrial Transmutation) core are presented. In the adopted strategy of MA burning the plutonium (Pu) balance of the core is minimized, allowing a reduction in the reactivity swing and the peak power form-factor deviation and an extension of the cycle duration. The MA/Pu ratio is used as a variable for the fuel optimisation studies. The efficiency of MA transmutation is close to the foreseen theoretical value of 42 kg TW -1 h -1 when level of Pu in the actinide mixture is about 40 wt.%. The obtained results are compared with the reference case of the EFIT core loaded with the composite CERCER fuel, where fuel particles are incorporated in a ceramic magnesia matrix. The results of this study offer additional information for the EFIT fuel selection.

Neutronics calculation of RTP core

NASA Astrophysics Data System (ADS)

Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

2017-01-01

Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

Biodiesel fuel management best practices for transit

DOT National Transportation Integrated Search

2007-11-27

Public transportation systems play a key role throughout the country not only in providing vital services to citizens but also in the environmental quality of our communities. Transit systems nationwide are seeking out new technologies in order to in...

Performance of miniaturized direct methanol fuel cell (DMFC) devices using micropump for fuel delivery

NASA Astrophysics Data System (ADS)

Zhang, Tao; Wang, Qing-Ming

A fuel cell is a device that can convert chemical energy into electricity directly. Among various types of fuel cells, both polymer electrolyte membrane fuel cells (PEMFCs) and direct methanol fuel cells (DMFCs) can work at low temperature (<80 °C). Therefore, they can be used to supply power for commercial portable electronics such as laptop computers, digital cameras, PDAs and cell phones. The focus of this paper is to investigate the performance of a miniaturized DMFC device using a micropump to deliver fuel. The core of this micropump is a piezoelectric ring-type bending actuator and the associated nozzle/diffuser for directing fuel flow. Based on the experimental measurements, it is found that the performance of the fuel cell can be significantly improved if enough fuel flow is induced by the micropump at anode. Three factors may contribute to the performance enhancement including replenishment of methanol, decrease of diffusion resistance and removal of carbon dioxide. In comparison with conventional mini pumps, the size of the piezoelectric micropump is much smaller and the energy consumption is much lower. Thus, it is very viable and effective to use a piezoelectric valveless micropump for fuel delivery in miniaturized DMFC power systems.

Detection of Ketenimine (CH2CNH) in SGRB2(N) Hot Cores

NASA Technical Reports Server (NTRS)

Lovas, F. J.; Hollis, J. M.; Remijan, Anthony J.; Jewell, P. R.

2006-01-01

Ketenimine (CH2CNH) has been detected in absorption toward the starforming region Sagittarius B2(N) with the 100-m Green Bank Telescope (GBT) by means of three rotational transitions: 7(sub 16)-8(sub 08) at 41.5 GHz, 8(sub 19)-9(sub 09), at 23.2 GHz, and 9(sub 18)-10(sub 0,10) at 4.9 GHz. Ketenimine has a sparse rotational spectrum below 50 GHz. From transition line strength arguments, the spectral lines found are the ones most likely to be detected and occur in spectral regions that have little possibility of confusion with other molecular species. Partially resolved hyperfine structure is apparent in the 4.9 GHz transition which has energy levels approximately 50 K above ground state level; the absorption seen in this transition appears to be emanating from gas in close proximity to the LMH hot core that has a systemic LSR velocity of +64 kilometers per second. By comparison, the 41.5 GHz and 23.2 GHz transitions have lower energy levels of approximately 33 K and approximately 41 K, respectively; and show absorption against the two star-forming SgrB2(N) hot cores with systematic LSR velocities of +64 (the LMH) and +82 kilometers per second. These ketenimine data show that the hot core at +82 kilometers per second is cooler than the hot core at +64 kilometers per second. Ketenimine is likely formed directly from its isomer methyl cyanide (CH3CN) by tautomerization driven by shocks that pervade the star-forming region.

Cool Core Disruption in Abell 1763

NASA Astrophysics Data System (ADS)

Douglass, Edmund; Blanton, Elizabeth L.; Clarke, Tracy E.; Randall, Scott W.; Edwards, Louise O. V.; Sabry, Ziad

2017-01-01

We present the analysis of a 20 ksec Chandra archival observation of the massive galaxy cluster Abell 1763. A model-subtracted image highlighting excess cluster emission reveals a large spiral structure winding outward from the core to a radius of ~950 kpc. We measure the gas of the inner spiral to have significantly lower entropy than non-spiral regions at the same radius. This is consistent with the structure resulting from merger-induced motion of the cluster’s cool core, a phenomenon seen in many systems. Atypical of spiral-hosting clusters, an intact cool core is not detected. Its absence suggests the system has experienced significant disruption since the initial dynamical encounter that set the sloshing core in motion. Along the major axis of the elongated ICM distribution we detect thermal features consistent with the merger event most likely responsible for cool core disruption. The merger-induced transition towards non-cool core status will be discussed. The interaction between the powerful (P1.4 ~ 1026 W Hz-1) cluster-center WAT radio source and its ICM environment will also be discussed.

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, David

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allowmore » for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.« less

Advanced Fuels for LWRs: Fully-Ceramic Microencapsulated and Related Concepts FY 2012 Interim Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

R. Sonat Sen; Brian Boer; John D. Bess

2012-03-01

This report summarizes the progress in the Deep Burn project at Idaho National Laboratory during the first half of fiscal year 2012 (FY2012). The current focus of this work is on Fully-Ceramic Microencapsulated (FCM) fuel containing low-enriched uranium (LEU) uranium nitride (UN) fuel kernels. UO2 fuel kernels have not been ruled out, and will be examined as later work in FY2012. Reactor physics calculations confirmed that the FCM fuel containing 500 mm diameter kernels of UN fuel has positive MTC with a conventional fuel pellet radius of 4.1 mm. The methodology was put into place and validated against MCNP tomore » perform whole-core calculations using DONJON, which can interpolate cross sections from a library generated using DRAGON. Comparisons to MCNP were performed on the whole core to confirm the accuracy of the DRAGON/DONJON schemes. A thermal fluid coupling scheme was also developed and implemented with DONJON. This is currently able to iterate between diffusion calculations and thermal fluid calculations in order to update fuel temperatures and cross sections in whole-core calculations. Now that the DRAGON/DONJON calculation capability is in place and has been validated against MCNP results, and a thermal-hydraulic capability has been implemented in the DONJON methodology, the work will proceed to more realistic reactor calculations. MTC calculations at the lattice level without the correct burnable poison are inadequate to guarantee zero or negative values in a realistic mode of operation. Using the DONJON calculation methodology described in this report, a startup core with enrichment zoning and burnable poisons will be designed. Larger fuel pins will be evaluated for their ability to (1) alleviate the problem of positive MTC and (2) increase reactivity-limited burnup. Once the critical boron concentration of the startup core is determined, MTC will be calculated to verify a non-positive value. If the value is positive, the design will be changed to

OTTER: An Optimized Transit Tool And Easy Reference

DTIC Science & Technology

2016-03-01

as stated in the CNO’s Position Report: 2014. While a number of fuel -saving measures have been implemented in recent years, the effects of...saving measures have been implemented in recent years, the effects of operational transit speed on fuel consumption have not been adequately...their estimated savings As clearly seen, operating configuration has the most effect by far on fuel savings. Source: Fonte S (2009). In 2015

FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

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

1963-05-01

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

FTA fuel cell bus program : research accomplishments through 2011.