National Library of Energy BETA

Sample records for graphite research reactor

  1. Brookhaven Graphite Research Reactor Workshop

    Broader source: Energy.gov [DOE]

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

  2. Graphite Reactor | ornl.gov

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

    Graphite Reactor 'In the early, desperate days of World War II, the United States launched the top-secret, top-priority Manhattan Project...' In the early, desperate days of U.S....

  3. Deployment of Smart 3D Subsurface Contaminant Characterization at the Brookhaven Graphite Research Reactor

    SciTech Connect (OSTI)

    Sullivan, T.; Heiser, J.; Kalb, P.; Milian, L.; Newson, C.; Lilimpakas, M.; Daniels, T.

    2002-02-26

    The Brookhaven Graphite Research Reactor (BGRR) Historical Site Assessment (BNL 1999) identified contamination inside the Below Grade Ducts (BGD) resulting from the deposition of fission and activation products from the pile on the inner carbon steel liner during reactor operations. Due to partial flooding of the BGD since shutdown, some of this contamination may have leaked out of the ducts into the surrounding soils. The baseline remediation plan for cleanup of contaminated soils beneath the BGD involves complete removal of the ducts, followed by surveying the underlying and surrounding soils, then removing soil that has been contaminated above cleanup goals. Alternatively, if soil contamination around and beneath the BGD is either non-existent/minimal (below cleanup goals) or is very localized and can be ''surgically removed'' at a reasonable cost, the BGD can be decontaminated and left in place. The focus of this Department of Energy Accelerated Site Technology Deployment (DOE ASTD) project was to determine the extent (location, type, and level) of soil contamination surrounding the BGD and to present this data to the stakeholders as part of the Engineering Evaluation/Cost Analysis (EE/CA) process. A suite of innovative characterization tools was used to complete the characterization of the soil surrounding the BGD in a cost-effective and timely fashion and in a manner acceptable to the stakeholders. The tools consisted of a tracer gas leak detection system that was used to define the gaseous leak paths out of the BGD and guide soil characterization studies, a small-footprint Geoprobe to reach areas surrounding the BGD that were difficult to access, two novel, field-deployed, radiological analysis systems (ISOCS and BetaScint) and a three-dimensional (3D) visualization system to facilitate data analysis/interpretation. All of the technologies performed as well or better than expected and the characterization could not have been completed in the same time or at the same cost without implementing this approach.

  4. Nuclear Graphite -Fission Reactor Brief Outline of Experience and

    E-Print Network [OSTI]

    McDonald, Kirk

    Nuclear Graphite - Fission Reactor Brief Outline of Experience and Understanding Professor Barry J Marsden and Dr. Graham N Hall Nuclear Graphite Research Group The University of Manchester 20 March 201313 9PL Tel: +44 (0) 161 275 4399, barry.marsden@manchester.ac.uk #12;Overview · Nuclear Graphite

  5. US graphite reactor D&D experience

    SciTech Connect (OSTI)

    Garrett, S.M.K.; Williams, N.C.

    1997-02-01

    This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE).

  6. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE BROOKHAVEN GRAPHITE RESEARCH REACTOR ENGINEERED CAP, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK DCN 5098-SR-07-0

    SciTech Connect (OSTI)

    Evan Harpenau

    2011-07-15

    The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the Brookhaven Graphite Research Reactor (BGRR) Engineered Cap at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Science Associates (BSA) have completed removal of affected soils and performed as-left surveys by BSA associated with the BGRR Engineered Cap. Sample results have been submitted, as required, to demonstrate that remediation efforts comply with the cleanup goal of {approx}15 mrem/yr above background to a resident in 50 years (BNL 2011a).

  7. Graphite Technology Development Plan

    SciTech Connect (OSTI)

    W. Windes; T. Burchell; M.Carroll

    2010-10-01

    The Next Generation Nuclear Plant (NGNP) will be a helium-cooled High Temperature Gas Reactor (HTGR) with a large graphite core. Graphite physically contains the fuel and comprises the majority of the core volume. Graphite has been used effectively as a structural and moderator material in both research and commercial high-temperature gas-cooled reactors. This development has resulted in graphite being established as a viable structural material for HTGRs. While the general characteristics necessary for producing nuclear grade graphite are understood, historical “nuclear” grades no longer exist. New grades must be fabricated, characterized, and irradiated to demonstrate that current grades of graphite exhibit acceptable non-irradiated and irradiated properties upon which the thermomechanical design of the structural graphite in NGNP is based. This Technology Development Plan outlines the research and development (R&D) activities and associated rationale necessary to qualify nuclear grade graphite for use within the NGNP reactor.

  8. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    SciTech Connect (OSTI)

    Mcwilliams, A. J.

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  9. Treatment of Irradiated Graphite from French Bugey Reactor - 13424

    SciTech Connect (OSTI)

    Brown, Thomas [Studsvik, Inc., 5605 Glenridge Drive NE, Suite 705, Atlanta, GA (United States)] [Studsvik, Inc., 5605 Glenridge Drive NE, Suite 705, Atlanta, GA (United States); Poncet, Bernard [electricite de France, 154 Avenue Thiers, CS 60018, 69458 Lyon Cedex 06 (France)] [electricite de France, 154 Avenue Thiers, CS 60018, 69458 Lyon Cedex 06 (France)

    2013-07-01

    Beginning in 2009, in order to determine an alternative to direct disposal for decommissioned irradiated graphite from EDF's Bugey NPP, Studsvik and EDF began a test program to determine if graphite decontamination and destruction were practicable using Studsvik's thermal organic reduction (THOR) technology. The testing program focused primarily on the release of C-14, H-3, and Cl-36 and also monitored graphite mass loss. For said testing, a bench-scale steam reformer (BSSR) was constructed with the capability of flowing various compositions of gases at temperatures up to 1300 deg. C over uniformly sized particles of graphite for fixed amounts of time. The BSSR was followed by a condenser, thermal oxidizer, and NaOH bubbler system designed to capture H-3 and C-14. Also, in a separate series of testing, high concentration acid and peroxide solutions were used to soak the graphite and leach out and measure Cl-36. A series of gasification tests were performed to scope gas compositions and temperatures for graphite gasification using steam and oxygen. Results suggested higher temperature steam (1100 deg. C vs. 900 deg. C) yielded a practicable gasification rate but that lower temperature (900 deg. C) gasification was also a practicable treatment alternative if oxygen is fed into the process. A series of decontamination tests were performed to determine the release behavior of and extent to which C-14 and H-3 were released from graphite in a high temperature (900-1300 deg. C), low flow roasting gas environment. In general, testing determined that higher temperatures and longer roasting times were efficacious for releasing H-3 completely and the majority (80%) of C-14. Manipulating oxidizing and reducing gas environments was also found to limit graphite mass loss. A series of soaking tests was performed to measure the amount of Cl-36 in the samples of graphite before and after roasting in the BSSR. Similar to C-14 release, these soaking tests revealed that 70-80% Cl-36 is released during roasting tests. (authors)

  10. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The first experiment was inserted in the ATR in August 2009 and started its irradiation in September 2009. It is anticipated to complete its irradiation in early calendar 2011. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and the irradiation experience to date.

  11. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    SciTech Connect (OSTI)

    Rosenthal, Murray Wilford

    2009-08-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  12. Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics

    E-Print Network [OSTI]

    S. R. Mirfayzi

    2013-01-08

    Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermal neutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermal neutron diffusion length in graphite is evaluated at $50.85 \\pm 0.3cm$ using COMSOL Multiphysics and $50.95 \\pm 0.5cm$ using MCNP. Finally the total neutron cross-section is derived using FEA in an inverse iteration form.

  13. Advanced Reactor Research and Development Funding Opportunity...

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

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

  14. Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

    SciTech Connect (OSTI)

    Romano, T.

    1997-09-29

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  15. Status of the NGNP graphite creep experiments AGC-1 and AGC-2 irradiated in the advanced test reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2014-05-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the next generation nuclear plant (NGNP) very high temperature gas reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have three different compressive loads applied to the top half of three diametrically opposite pairs of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment.

  16. Oxidation Resistant Graphite Studies

    SciTech Connect (OSTI)

    W. Windes; R. Smith

    2014-07-01

    The Very High Temperature Reactor (VHTR) Graphite Research and Development Program is investigating doped nuclear graphite grades exhibiting oxidation resistance. During a oxygen ingress accident the oxidation rates of the high temperature graphite core region would be extremely high resulting in significant structural damage to the core. Reducing the oxidation rate of the graphite core material would reduce the structural effects and keep the core integrity intact during any air-ingress accident. Oxidation testing of graphite doped with oxidation resistant material is being conducted to determine the extent of oxidation rate reduction. Nuclear grade graphite doped with varying levels of Boron-Carbide (B4C) was oxidized in air at nominal 740°C at 10/90% (air/He) and 100% air. The oxidation rates of the boronated and unboronated graphite grade were compared. With increasing boron-carbide content (up to 6 vol%) the oxidation rate was observed to have a 20 fold reduction from unboronated graphite. Visual inspection and uniformity of oxidation across the surface of the specimens were conducted. Future work to determine the remaining mechanical strength as well as graphite grades with SiC doped material are discussed.

  17. 2012 Annual Report Research Reactor Infrastructure Program

    SciTech Connect (OSTI)

    Douglas Morrell

    2012-11-01

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

  18. International Research Reactor Decommissioning Project

    SciTech Connect (OSTI)

    Leopando, Leonardo; Warnecke, Ernst

    2008-01-15

    Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement to the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.

  19. Status of the NGNP Graphite Creep Experiments AGC-1 and AGC-2 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.

  20. Baseline Graphite Characterization: First Billet

    SciTech Connect (OSTI)

    Mark C. Carroll; Joe Lords; David Rohrbaugh

    2010-09-01

    The Next Generation Nuclear Plant Project Graphite Research and Development program is currently establishing the safe operating envelope of graphite core components for a very high temperature reactor design. To meet this goal, the program is generating the extensive amount of quantitative data necessary for predicting the behavior and operating performance of the available nuclear graphite grades. In order determine the in-service behavior of the graphite for the latest proposed designs, two main programs are underway. The first, the Advanced Graphite Creep (AGC) program, is a set of experiments that are designed to evaluate the irradiated properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences, and compressive loads. Despite the aggressive experimental matrix that comprises the set of AGC test runs, a limited amount of data can be generated based upon the availability of space within the Advanced Test Reactor and the geometric constraints placed on the AGC specimens that will be inserted. In order to supplement the AGC data set, the Baseline Graphite Characterization program will endeavor to provide supplemental data that will characterize the inherent property variability in nuclear-grade graphite without the testing constraints of the AGC program. This variability in properties is a natural artifact of graphite due to the geologic raw materials that are utilized in its production. This variability will be quantified not only within a single billet of as-produced graphite, but also from billets within a single lot, billets from different lots of the same grade, and across different billets of the numerous grades of nuclear graphite that are presently available. The thorough understanding of this variability will provide added detail to the irradiated property data, and provide a more thorough understanding of the behavior of graphite that will be used in reactor design and licensing. This report covers the development of the Baseline Graphite Characterization program from a testing and data collection standpoint through the completion of characterization on the first billet of nuclear-grade graphite. This data set is the starting point for all future evaluations and comparisons of material properties.

  1. Irradiation Creep in Graphite

    SciTech Connect (OSTI)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  2. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    SciTech Connect (OSTI)

    Douglas Morrell

    2011-03-01

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

  3. Brookhaven Lab Completes Decommissioning of Graphite Research Reactor:

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirley Ann JacksonDepartment|Marketing, LLC |EnergycurrentlyJune 2012 This booklet

  4. Removal of 14C from Irradiated Graphite for Graphite Recycle and Waste Volume Reduction

    SciTech Connect (OSTI)

    Dunzik-Gougar, Mary Lou; Windes, Will; Marsden, Barry

    2014-06-10

    The aim of the research presented here was to identify the chemical form of 14C in irradiated graphite. A greater understanding of the chemical form of this longest-lived isotope in irradiated graphite will inform not only management of legacy waste, but also development of next generation gas-cooled reactors. Approximately 250,000 metric tons of irradiated graphite waste exists worldwide, with the largest single quantity originating in the Magnox and AGR reactors of UK. The waste quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation I gas-cooled, graphite moderated reactors. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 14C, with a half-life of 5730 years.

  5. Role of Nuclear Grade Graphite in Oxidation in Modular HTGRs

    SciTech Connect (OSTI)

    Willaim Windes; G. Strydom; J. Kane; R. Smith

    2014-11-01

    The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of core environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.

  6. Evaluation of graphite/steam interactions for ITER (International Thermonuclear Experimental Reactor)

    SciTech Connect (OSTI)

    Smolik, G.R.; Merrill, B.J.; Piet, S.J.; Holland, D.F.

    1990-09-01

    In this report we present the results of an experimental/analytical study designed to determine the quantity of hydrogen generated during a coolant inleakage accident in ITER. This hydrogen could represent a potential explosive hazard, provided the proper conditions exist, causing machine damage and release of radioactive material. We have measured graphite/steam reaction rates for several graphites and carbon-based composites at temperatures between 1000 C and 1700 C. The effects of steam flow rate, and partial pressure were also examined. The measured reaction rates correlated well with two Arrhenius type relationships. We have used the relationships for GraphNOL N3M in a thermal model to determine that for ITER the quantity of hydrogen produced would range between 5 and 35 kg, depending upon how the graphite tiles are attached to the first wall. While 5 kg is not a significant concern, 35 kg presents an explosive hazard. 20 refs., 14 figs., 1 tab.

  7. Supply of enriched uranium for research reactors

    SciTech Connect (OSTI)

    Mueller, H. [NUKEM GmbH, Alzenau (Germany)

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  8. Fuel elements of research reactor CM

    SciTech Connect (OSTI)

    Kozlov, A.V.; Morozov, A.V.; Vatulin, A.V.; Ershov, S.A.

    2013-07-01

    In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.

  9. Experimental Study of the Thermal-Hydraulic Phenomena in the Reactor Cavity Cooling System and Analysis of the Effects of Graphite Dispersion 

    E-Print Network [OSTI]

    Vaghetto, Rodolfo

    2012-07-16

    An experimental activity was performed to observe and study the effects of graphite dispersion and deposition on thermal hydraulic phenomena in a Reactor Cavity Cooling System (RCCS). The small scale RCCS experimental facility (16.5cm x 16.5cm x 30...

  10. Probabilistic Safety Assessment of Tehran Research Reactor

    SciTech Connect (OSTI)

    Hosseini, Seyed Mohammad Hadi; Nematollahi, Mohammad Reza; Sepanloo, Kamran

    2004-07-01

    Probabilistic Safety Assessment (PSA) application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this paper the application of the Probabilistic Safety Assessment to the Tehran Research Reactor (TRR) is presented. The level 1 PSA application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using SAPHIRE software. This Study shows that the obtained core damage frequency for Tehran Research Reactor (8.368 E-6 per year) well meets the IAEA criterion for existing nuclear power plants (1E-4). But safety improvement suggestions are offered to decrease the most probable accidents. (authors)

  11. Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities 

    E-Print Network [OSTI]

    Rauch, Eric B.

    2010-07-14

    this type of reactor desirable also make it suspicious to the international community as a possible means to shorten that state?s nuclear latency. If a safeguards approach could be developed for a fuel cycle featuring one of these reactors, it would ease...

  12. Light Water Reactor Fuel Cladding Research and Testing | ornl...

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

    Light Water Reactor Fuel Cladding Research June 01, 2013 Severe Accident Test Station ORNL is the focus point for Light Water Reactor (LWR) fuel cladding research and testing. The...

  13. Research and Medical Isotope Reactor Supply | Y-12 National Security...

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

    Y-12 tops the short list of the world's most secure, reliable uranium feedstock suppliers for dozens of research and test reactors on six continents. These reactors can be...

  14. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    SciTech Connect (OSTI)

    Not Available

    1987-05-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions.

  15. NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944

    E-Print Network [OSTI]

    #12;#12;11 #12;2 NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944 Nuclear fission discovered 430 nuclear power reactors are operating in the world, and 103 nuclear power plants produce 20, naval reactors, and nuclear power plants. Oak Ridge experiments byArt Snell in 1944 showed that 10 tons

  16. Research Reactor Conversion | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Reactor Conversion | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the...

  17. GRAFEC: A New Spanish Program to Investigate Waste Management Options for Radioactive Graphite - 12399

    SciTech Connect (OSTI)

    Marquez, Eva; Pina, Gabriel; Rodriguez, Marina; Fachinger, Johannes; Grosse, Karl-Heinz; Leganes Nieto, Jose Luis; Quiros Gracian, Maria

    2012-07-01

    Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m{sup 3} each and placed in interim storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like {sup 14}C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of {sup 14}C by thermal methods. This method is based on results obtained at the Research Centre Juelich (FZJ) in the Frame of the EC programs 'Raphael' and 'Carbowaste'. The process developed at FZJ is based on a preferential oxidation of {sup 14}C in comparison to the bulk {sup 12}C. Explanations for this effect are the inhomogeneous distribution and a weaker bounding of {sup 14}C which is not incorporated in the graphite lattice. However these investigations have only been performed with graphite from the high temperature reactor Arbeitsgemeinschaft Versuchsreaktor Juelich AVR which has been operated in a non-oxidising condition or research reactor graphite operated at room temperature. The reactor Vandellos I has been operated with CO{sub 2} as coolant and significant amounts of graphite have been already oxidised. The aim of the project is to validate whether a {sup 14}C decontamination can also been achieved with graphite from Vandellos I. A second possibility under investigation is the encapsulation of the graphite in a long term stable glass matrix. The principal applicability has been already proved by FNAG. Crushed graphite mixed with a suitable glass powder has been pressed at elevated temperature under vacuum. The vacuum is required to avoid gas enclosures in the obtained product. The obtained products, named IGM for 'Impermeable Graphite Matrix', have densities above 99% of theoretical density. The amount of glass has been chosen with respect to the pore volume of the former graphite parts. The method allows the production of encapsulated graphite without increasing the disposal volume. This paper will give a short overview of characterisation results of different irradiated graphite materials obtained at CIEMAT and in the Carbowaste project as well as the proposed methods and the actual status of the program including first results about leaching of non-radioactive IGM samples and hopefully first tendencies concerning the C-14 separation from graphite of Vandellos I by thermal treatment. Both processes, the thermal treatment as well as the IGM, have the potential to solve problems related to the management of irradiated graphite in Spain. However the methods have only been tested with different types of i-graphite and virgin graphite, respectively. Only investigations with real i-graphite from Spain will reveal whether the described methods are applicable to graphite from Vandellos I. However all partners are convinced that one of these new methods or a combination of them will lead to a feasible option to manage i-graphite in Spain on an industrial scale. (authors)

  18. Uranium Oxide Aerosol Transport in Porous Graphite

    SciTech Connect (OSTI)

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

    2012-01-23

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

  19. Sodium fast reactor safety and licensing research plan. Volume II.

    SciTech Connect (OSTI)

    Ludewig, H.; Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.; Lambert, J.; Hayes, S.; Sackett, J.; Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  20. On the RA research reactor fuel management problems

    SciTech Connect (OSTI)

    Matausek, M.V.; Marinkovic, N.

    1997-12-01

    After 25 yr of operation, the Soviet-origin 6.5-MW heavy water RA research reactor was shut down in 1984. Basic facts about RA reactor operation, aging, reconstruction, and spent-fuel disposal have been presented and discussed in earlier papers. The following paragraphs present recent activities and results related to important fuel management problems.

  1. Spent graphite fuel element processing

    SciTech Connect (OSTI)

    Holder, N.D.; Olsen, C.W.

    1981-07-01

    The Department of Energy currently sponsors two programs to demonstrate the processing of spent graphite fuel elements. General Atomic in San Diego operates a cold pilot plant to demonstrate the processing of both US and German high-temperature reactor fuel. Exxon Nuclear Idaho Company is demonstrating the processing of spent graphite fuel elements from Rover reactors operated for the Nuclear Rocket Propulsion Program. This work is done at Idaho National Engineering Laboratory, where a hot facility is being constructed to complete processing of the Rover fuel. This paper focuses on the graphite combustion process common to both programs.

  2. Annular Core Research Reactor - Critical to Science-Based Weapons...

    National Nuclear Security Administration (NNSA)

    to science-based weapons design and certification. The ACRR is a pool-type research reactor (Hazard Category 2 Nuclear Facility) that has been in operation since the 1970s...

  3. Development of Technical Nuclear Forensics for Spent Research Reactor Fuel 

    E-Print Network [OSTI]

    Sternat, Matthew Ryan 1982-

    2012-11-20

    Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis method was employed to reconstruct ...

  4. The fight to save the university research reactors

    SciTech Connect (OSTI)

    Bobeck, L.M.; Perez, P.B.

    1993-10-01

    This article looks at impacts of Nuclear Regulatory Commission actions on nonprofit educational reactors. In mid-July the NRC issued a ruling on fee policy, which eliminated the historical fee exemeption for nonprofit research reactors. The ensuing regulatory fees placed an economic burden on these facilities which was likely to close many of them. On September 13, the NRC agreed to reconsider this rule. In part this reflects that this rule had an impact on a larger user base than just research reactors. The article summarizes this problem, and tries to put it in perspective for the reader.

  5. Graphite Oxidation Simulation in HTR Accident Conditions

    SciTech Connect (OSTI)

    El-Genk, Mohamed

    2012-10-19

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air ?helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  6. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    SciTech Connect (OSTI)

    Diamond, D. J.; Baek, J. S.; Hanson, A. L.; Cheng, L-Y; Brown, N.; Cuadra, A.

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  7. Background Radiation Measurements at High Power Research Reactors

    E-Print Network [OSTI]

    Ashenfelter, J; Baldenegro, C X; Band, H R; Barclay, G; Bass, C D; Berish, D; Bowden, N S; Bryan, C D; Cherwinka, J J; Chu, R; Classen, T; Davee, D; Dean, D; Deichert, G; Dolinski, M J; Dolph, J; Dwyer, D A; Fan, S; Gaison, J K; Galindo-Uribarri, A; Gilje, K; Glenn, A; Green, M; Han, K; Hans, S; Heeger, K M; Heffron, B; Jaffe, D E; Kettell, S; Langford, T J; Littlejohn, B R; Martinez, D; McKeown, R D; Morrell, S; Mueller, P E; Mumm, H P; Napolitano, J; Norcini, D; Pushin, D; Romero, E; Rosero, R; Saldana, L; Seilhan, B S; Sharma, R; Stemen, N T; Surukuchi, P T; Thompson, S J; Varner, R L; Wang, W; Watson, S M; White, B; White, C; Wilhelmi, J; Williams, C; Wise, T; Yao, H; Yeh, M; Yen, Y -R; Zhang, C; Zhang, X

    2015-01-01

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including $\\gamma$-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  8. Background Radiation Measurements at High Power Research Reactors

    E-Print Network [OSTI]

    J. Ashenfelter; B. Balantekin; C. X. Baldenegro; H. R. Band; G. Barclay; C. D. Bass; D. Berish; N. S. Bowden; C. D. Bryan; J. J. Cherwinka; R. Chu; T. Classen; D. Davee; D. Dean; G. Deichert; M. J. Dolinski; J. Dolph; D. A. Dwyer; S. Fan; J. K. Gaison; A. Galindo-Uribarri; K. Gilje; A. Glenn; M. Green; K. Han; S. Hans; K. M. Heeger; B. Heffron; D. E. Jaffe; S. Kettell; T. J. Langford; B. R. Littlejohn; D. Martinez; R. D. McKeown; S. Morrell; P. E. Mueller; H. P. Mumm; J. Napolitano; D. Norcini; D. Pushin; E. Romero; R. Rosero; L. Saldana; B. S. Seilhan; R. Sharma; N. T. Stemen; P. T. Surukuchi; S. J. Thompson; R. L. Varner; W. Wang; S. M. Watson; B. White; C. White; J. Wilhelmi; C. Williams; T. Wise; H. Yao; M. Yeh; Y. -R. Yen; C. Zhang; X. Zhang

    2015-06-11

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including $\\gamma$-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  9. Background Radiation Measurements at High Power Research Reactors

    E-Print Network [OSTI]

    J. Ashenfelter; B. Balantekin; C. X. Baldenegro; H. R. Band; G. Barclay; C. D. Bass; D. Berish; N. S. Bowden; C. D. Bryan; J. J. Cherwinka; R. Chu; T. Classen; D. Davee; D. Dean; G. Deichert; M. J. Dolinski; J. Dolph; D. A. Dwyer; S. Fan; J. K. Gaison; A. Galindo-Uribarri; K. Gilje; A. Glenn; M. Green; K. Han; S. Hans; K. M. Heeger; B. Heffron; D. E. Jaffe; S. Kettell; T. J. Langford; B. R. Littlejohn; D. Martinez; R. D. McKeown; S. Morrell; P. E. Mueller; H. P. Mumm; J. Napolitano; D. Norcini; D. Pushin; E. Romero; R. Rosero; L. Saldana; B. S. Seilhan; R. Sharma; N. T. Stemen; P. T. Surukuchi; S. J. Thompson; R. L. Varner; W. Wang; S. M. Watson; B. White; C. White; J. Wilhelmi; C. Williams; T. Wise; H. Yao; M. Yeh; Y. -R. Yen; C. Zhang; X. Zhang

    2015-11-11

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including $\\gamma$-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  10. Research into Oil-based Colloidal-Graphite Lubricants for Forging of Al-based Alloys

    SciTech Connect (OSTI)

    Petrov, A.; Petrov, P.; Petrov, M.

    2011-05-04

    The presented paper describes the topical problem in metal forging production. It deals with the choice of an optimal lubricant for forging of Al-based alloys. Within the scope of the paper, the properties of several oil-based colloidal-graphite lubricants were investigated. The physicochemical and technological properties of these lubricants are presented. It was found that physicochemical properties of lubricant compositions have an influence on friction coefficient value and quality of forgings.The ring compression method was used to estimate the friction coefficient value. Hydraulic press was used for the test. The comparative analysis of the investigated lubricants was carried out. The forging quality was estimated on the basis of production test. The practical recommendations were given to choose an optimal oil-based colloidal-graphite lubricant for isothermal forging of Al-based alloy.

  11. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  12. Monash researchers led by Dr. Dan Li have developed a novel method for converting natural graphite into highly porous

    E-Print Network [OSTI]

    Albrecht, David

    natural graphite into highly porous graphene film for advanced applications. Figure 1: illustrates the conversion of inexpensive & abundant graphite into highly porous, mechanically robust conductive films (eg capacitors, batteries and fuel cells) n LCD displays and photovoltaic devices n Composites

  13. Research and educational activities at the MIT Research Reactor : Fiscal year 1968

    E-Print Network [OSTI]

    Massachusetts Institute of Technology. Department of Nuclear Engineering; 7102 Massachusetts Institute of Technology. Research Reactor. Staff; U.S. Atomic Energy Commission

    1968-01-01

    A report of research and educational activities which utilized the Massachusetts Institute of Technology, five-megawatt, heavy water, research reactor during fiscal year 1968 has been prepared for administrative use at MIT ...

  14. Inhibition of Oxidation in Nuclear Graphite

    SciTech Connect (OSTI)

    Phil Winston; James W. Sterbentz; William E. Windes

    2013-10-01

    Graphite is a fundamental material of high temperature gas cooled nuclear reactors, providing both structure and neutron moderation. Its high thermal conductivity, chemical inertness, thermal heat capacity, and high thermal structural stability under normal and off normal conditions contribute to the inherent safety of these reactor designs. One of the primary safety issues for a high temperature graphite reactor core is the possibility of rapid oxidation of the carbon structure during an off normal design basis event where an oxidizing atmosphere (air ingress) can be introduced to the hot core. Although the current Generation IV high temperature reactor designs attempt to mitigate any damage caused by a postualed air ingress event, the use of graphite components that inhibit oxidation is a logical step to increase the safety of these reactors. Recent experimental studies of graphite containing between 5.5 and 7 wt% boron carbide (B4C) indicate that oxidation is dramatically reduced even at prolonged exposures at temperatures up to 900°C. The proposed addition of B4C to graphite components in the nuclear core would necessarily be enriched in B-11 isotope in order to minimize B-10 neutron absorption and graphite swelling. The enriched boron can be added to the graphite during billet fabrication. Experimental oxidation rate results and potential applications for borated graphite in nuclear reactor components will be discussed.

  15. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.

  16. Graphite Gamma Scan Results

    SciTech Connect (OSTI)

    Mark W. Drigert

    2014-04-01

    This report documents the measurement and data analysis of the radio isotopic content for a series of graphite specimens irradiated in the first Advanced Graphite Creep (AGC) experiment, AGC-1. This is the first of a series of six capsules planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphites. The AGC-1 capsule was irradiated in the Advanced Test Reactor (ATR) at INL at approximately 700 degrees C and to a peak dose of 7 dpa (displacements per atom). Details of the irradiation conditions and other characterization measurements performed on specimens in the AGC-1 capsule can be found in “AGC-1 Specimen Post Irradiation Data Report” ORNL/TM 2013/242. Two specimens from six different graphite types are analyzed here. Each specimen is 12.7 mm in diameter by 25.4 mm long. The isotope with the highest activity was 60Co. Graphite type NBG-18 had the highest content of 60Co with an activity of 142.89 µCi at a measurement distance of 47 cm.

  17. Reactor Safety Research: Semiannual report, July-December 1986

    SciTech Connect (OSTI)

    Not Available

    1987-11-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions.

  18. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect (OSTI)

    Not Available

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  19. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    SciTech Connect (OSTI)

    Heeger, Karsten M.

    2014-09-13

    This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta_{13}$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

  20. Graphite in Science and Nuclear Technique

    E-Print Network [OSTI]

    E. I. Zhmurikov; I. A. Bubnenkov; V. V. Dremov; S. I. Samarin; A. S. Pokrovsky; D. V. Harkov

    2015-08-14

    The monograph is devoted to the application of graphite and graphite composites in science and technology. The structure and electrical properties, the technological aspects of production of high-strength synthetic graphites, the dynamics of the graphite destruction, traditionally used in the nuclear industry are discussed. It is focuses on the characteristics of graphitization and properties of graphite composites based on carbon isotope 13C. The book is based, generally, on the original results, and concentrated on the actual problems of application and testing of graphite materials in modern nuclear physics, in scientific and technical applications. For scientists and engineers specializing in nuclear physics and engineering, physics of nuclear reactors, condensed matter, for undergraduate, graduate and post-graduate students of universities physical specialties.

  1. Photo of the Week: The Brookhaven Graphite Research Reactor | Department of

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative Fuelsof EnergyAprilEnergyPartnership for EnergyPetroleumZeroEnergy The

  2. Friction pressure drop measurements and flow distribution analysis for LEU conversion study of MIT Research Reactor

    E-Print Network [OSTI]

    Wong, Susanna Yuen-Ting

    2008-01-01

    The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched ...

  3. Design and optimization of a high thermal flux research reactor via Kriging-based algorithm

    E-Print Network [OSTI]

    Kempf, Stephanie Anne

    2011-01-01

    In response to increasing demands for the services of research reactors, a 5 MW LEU-fueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic ...

  4. Very High Temperature Reactor (VHTR) Survey of Materials Research and Development Needs to Support Early Deployment

    SciTech Connect (OSTI)

    Eric Shaber; G. Baccaglini; S. Ball; T. Burchell; B. Corwin; T. Fewell; M. Labar; P. MacDonald; P. Rittenhouse; Russ Vollam; F. Southworth

    2003-01-01

    The VHTR reference concept is a helium-cooled, graphite moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 C or higher. It is expected that the VHTR will be purchased in the future as either an electricity producing plant with a direct cycle gas turbine or a hydrogen producing (or other process heat application) plant. The process heat version of the VHTR will require that an intermediate heat exchanger (IHX) and primary gas circulator be located in an adjoining power conversion vessel. A third VHTR mission - actinide burning - can be accomplished with either the hydrogen-production or gas turbine designs. The first ''demonstration'' VHTR will produce both electricity and hydrogen using the IHX to transfer the heat to either a hydrogen production plant or the gas turbine. The plant size, reactor thermal power, and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. The purpose of this report is to identify the materials research and development needs for the VHTR. To do this, we focused on the plant design described in Section 2, which is similar to the GT-MHR plant design (850 C core outlet temperature). For system or component designs that present significant material challenges (or far greater expense) there may be some viable design alternatives or options that can reduce development needs or allow use of available (cheaper) materials. Nevertheless, we were not able to assess those alternatives in the time allotted for this report and, to move forward with this material research and development assessment, the authors of this report felt that it was necessary to use a GT-MHR type design as the baseline design.

  5. Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

    SciTech Connect (OSTI)

    Smith, Anthony A. [Research Sites Restoration Ltd, Winfrith, Dorset (United Kingdom)] [Research Sites Restoration Ltd, Winfrith, Dorset (United Kingdom)

    2013-07-01

    The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] it is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)

  6. Advanced sodium fast reactor accident source terms : research needs.

    SciTech Connect (OSTI)

    Powers, Dana Auburn; Clement, Bernard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  7. Role of research reactors in training of NPP personnel with special focus on training reactor VR-1

    SciTech Connect (OSTI)

    Sklenka, L.; Rataj, J.; Frybort, J.; Huml, O. [Dept. of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical Univ. in Prague, V Holesovickach 2, Prague 8, 180 00 (Czech Republic)

    2012-07-01

    Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training program are demonstrated. (authors)

  8. Eastern Europe Research Reactor Initiative nuclear education and training courses - Current activities and future challenges

    SciTech Connect (OSTI)

    Snoj, L.; Sklenka, L.; Rataj, J.; Boeck, H.

    2012-07-01

    The Eastern Europe Research Reactor Initiative was established in January 2008 to enhance cooperation between the Research Reactors in Eastern Europe. It covers three areas of research reactor utilisation: irradiation of materials and fuel, radioisotope production, neutron beam experiments, education and training. In the field of education and training an EERRI training course was developed. The training programme has been elaborated with the purpose to assist IAEA Member States, which consider building a research reactor (RR) as a first step to develop nuclear competence and infrastructure in the Country. The major strength of the reactor is utilisation of three different research reactors and a lot of practical exercises. Due to high level of adaptability, the course can be tailored to specific needs of institutions with limited or no access to research reactors. (authors)

  9. University Research Reactor Task Force to the Nuclear Energy Research

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIX E LIST OF APPLICABLEStatutoryin theNuclear Energy Research andofRod BeeverDataAdvisory

  10. AGC-3 Graphite Preirradiation Data Analysis Report

    SciTech Connect (OSTI)

    William Windes; David Swank; David Rohrbaugh; Joseph Lord

    2013-09-01

    This report describes the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the third Advanced Graphite Capsule (AGC-3) irradiation capsule. The AGC-3 capsule is third in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. The general design of AGC-3 test capsule is similar to the AGC-2 test capsule, material property tests were conducted on graphite specimens prior to loading into the AGC-3 irradiation assembly. However the 6 major nuclear graphite grades in AGC-2 were modified; two previous graphite grades (IG-430 and H-451) were eliminated and one was added (Mersen’s 2114 was added). Specimen testing from three graphite grades (PCEA, 2114, and NBG-17) was conducted at Idaho National Laboratory (INL) and specimen testing for two grades (IG-110 and NBG-18) were conducted at Oak Ridge National Laboratory (ORNL) from May 2011 to July 2013. This report also details the specimen loading methodology for the graphite specimens inside the AGC-3 irradiation capsule. The AGC-3 capsule design requires "matched pair" creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-3 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce "matched pairs" of graphite samples above and below the AGC-3 capsule elevation mid-point to provide specimens with similar neutron dose levels.

  11. Opportunities for Materials Science and Biological Research at the OPAL Research Reactor

    SciTech Connect (OSTI)

    Kennedy, S. J.

    2008-03-17

    Neutron scattering techniques have evolved over more than 1/2 century into a powerful set of tools for determination of atomic and molecular structures. Modern facilities offer the possibility to determine complex structures over length scales from {approx}0.1 nm to {approx}500 nm. They can also provide information on atomic and molecular dynamics, on magnetic interactions and on the location and behaviour of hydrogen in a variety of materials. The OPAL Research Reactor is a 20 megawatt pool type reactor using low enriched uranium fuel, and cooled by water. OPAL is a multipurpose neutron factory with modern facilities for neutron beam research, radioisotope production and irradiation services. The neutron beam facility has been designed to compete with the best beam facilities in the world. After six years in construction, the reactor and neutron beam facilities are now being commissioned, and we will commence scientific experiments later this year. The presentation will include an outline of the strengths of neutron scattering and a description of the OPAL research reactor, with particular emphasis on it's scientific infrastructure. It will also provide an overview of the opportunities for research in materials science and biology that will be possible at OPAL, and mechanisms for accessing the facilities. The discussion will emphasize how researchers from around the world can utilize these exciting new facilities.

  12. IGORR-IV -- Proceedings of the fourth meeting of the International Group on Research Reactors

    SciTech Connect (OSTI)

    Rosenbalm, K.F. [comp.] [comp.

    1995-12-31

    The International Group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Twenty-nine papers were presented in five sessions and written versions of the papers or hard copies of the vugraphs used are published in these proceedings. The five sessions were: (1) Operating Research Reactors and Facility Upgrades; (2) Research Reactors in Design and Construction; (3) ANS Closeout Activities; (4) and (5) Research, Development, and Analysis Results.

  13. Irradiation Induced Dimensional Changes in Bulk Graphite; The theory

    E-Print Network [OSTI]

    S. V. Panyukov; A. V. Subbotin; M. V. Arjakov

    2012-10-14

    Basing on experimental data on irradiation-induced deformation of graphite we introduced a concept of diffuse domain structure developed in reactor graphite produced by extrusion. Such domains are considered as random continuous deviations of local graphite texture from the global one. We elucidate the origin of domain structure and estimate the size and the degree of orientational ordering of its domains. Using this concept we explain the well known radiation-induced size effect observed in reactor graphite. We also propose a method for converting the experimental data on shape-change of finite-size samples to bulk graphite. This method gives a more accurate evaluation of corresponding data used in estimations of reactor graphite components lifetime under irradiation.

  14. Reprocessing of research reactor fuel the Dounreay option

    SciTech Connect (OSTI)

    Cartwright, P.

    1997-08-01

    Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.

  15. Sodium fast reactor fuels and materials : research needs.

    SciTech Connect (OSTI)

    Denman, Matthew R.; Porter, Douglas; Wright, Art; Lambert, John; Hayes, Steven; Natesan, Ken; Ott, Larry J.; Garner, Frank; Walters, Leon; Yacout, Abdellatif

    2011-09-01

    An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

  16. Nuclear plant-aging research on reactor protection systems

    SciTech Connect (OSTI)

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed.

  17. Sodium fast reactor safety and licensing research plan. Volume I.

    SciTech Connect (OSTI)

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  18. Instrumentation for Neutron Scattering at the Missouri University Research Reactor Paul F. Miceli

    E-Print Network [OSTI]

    Montfrooij, Wouter

    Instrumentation for Neutron Scattering at the Missouri University Research Reactor Paul F. Miceli Research Reactor (MURR) provides significant thermal neutron flux, which enables neutron scattering]. There are presently 5 instruments located on the beam port floor that are dedicated to neutron scattering: (1) TRIAX

  19. Preparation of graphitic articles

    DOE Patents [OSTI]

    Phillips, Jonathan; Nemer, Martin; Weigle, John C.

    2010-05-11

    Graphitic structures have been prepared by exposing templates (metal, metal-coated ceramic, graphite, for example) to a gaseous mixture that includes hydrocarbons and oxygen. When the template is metal, subsequent acid treatment removes the metal to yield monoliths, hollow graphitic structures, and other products. The shapes of the coated and hollow graphitic structures mimic the shapes of the templates.

  20. Second generation Research Reactor Fuel Container (RRFC-II).

    SciTech Connect (OSTI)

    Abhold, M. E.; Baker, M. C.; Bourret, S. C.; Harker, W. C.; Pelowitz, D. G.; Polk, P. J.

    2001-01-01

    The second generation Research Reactor Fuel Counter (RRFC-II) has been developed to measure the remaining {sup 235}U content in foreign spent Material Test Reactor (MTR)-type fuel being returned to the Westinghouse Savannah River Site (WSRS) for interim storage and subsequent disposal. The fuel to be measured started as fresh fuel nominally with 93% enriched Uraniuin alloyed with A1 clad in Al. The fuel was irradiated to levels of up to 65% burnup. The RRFC-II, which will be located in the L-Basin spent fuel pool, is intended to assay the {sup 235}U content using a combination of passive neutron coincidence counting, active neutron coincidence counting, and active-multiplicity analysis. Measurements will be done underwater, eliminating the need for costly and hazardous handling operations of spent fuel out of water. The underwater portion of the RRFC-II consists of a watertight stainless steel housing containing neutron and gamma detectors and a scanning active neutron source. The portion of the system that resides above water consists of data-processing electronics; electromechanical drive electronics; a computer to control the operation of the counter, to collect, and to analyze data; and a touch screen interface located at the equipment rack. The RRFC-II is an improved version of the Los Alamos-designed RRFC already installed in the SRS Receipts Basin for Offsite Fuel. The RRFC-II has been fabricated and is scheduled for installation in late FY 2001 pending acceptance testing by Savannah River Site personnel.

  1. The Optimized Integration of the Decontamination Plan and the Radwaste Management Plan into Decommissioning Plan to the VVR-S Research Reactor from Romania

    SciTech Connect (OSTI)

    Barariu, G. [National Authority for Nuclear Activity-Subsidiary of Technology and Engineering for Nuclear Projects (Romania)

    2008-07-01

    The paper presents the progress of the Decontamination Plan and Radioactive Waste Management Plan which accompanies the Decommissioning Plan for research reactor VVR-S located in Magurele, Ilfov, near Bucharest, Romania. The new variant of the Decommissioning Plan was elaborated taking into account the IAEA recommendation concerning radioactive waste management. A new feasibility study for VVR-S decommissioning was also elaborated. The preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as rehabilitation of the existing Radioactive Waste Treatment Plant and the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and reusing of cleaned reactor building is envisaged. An inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the IAEA assistance. Environmental concerns are a part of the radioactive waste management strategy. In conclusion: The current version 8 of the Draft Decommissioning Plan which include the Integrated concept of Decontamination and Decommissioning and Radwaste Management, reflects the substantial work that has been incorporated by IFIN-HH in collaboration with SITON, which has resulted in substantial improvement in document The decommissioning strategy must take into account costs for VVR-S Reactor decommissioning, as well as costs for much needed refurbishments to the radioactive waste treatment plant and the Baita-Bihor waste disposal repository. Several improvements to the Baita-Bihor repository and IFIN-HH waste treatment facility were proposed. The quantities and composition of the radioactive waste generated by VVR-S Reactor dismantling were again estimated by streams and the best demonstrated practicable processing solution was proposed. The estimated quantities of materials to be managed in the near future raise some issues that need to be solved swiftly, such as treatment of aluminum and lead and graphite management. It is envisaged that these materials to be treated to Subsidiary for Nuclear Research (SCN) Pitesti. (authors)

  2. Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    National Nuclear Security Administration (NNSA)

    * Complete reactor control rod system. * Note: Does not include the steam turbine generator portion of the power plant. - Sensitive nuclear technology: Any information...

  3. Annular Core Research Reactor at Sandia National Laboratories...

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

    at Sandia National Laboratories achieves 10,000th reactor pulse operation | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the...

  4. U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization

    SciTech Connect (OSTI)

    Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

    2004-10-03

    The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies.

  5. DOE-NE Light Water Reactor Sustainability Program and EPRI Long Term Operations Program – Joint Research and Development

    Broader source: Energy.gov [DOE]

    Description of Joint DOE and EPRI research and development programs related to reactor sustainability INL/EXT-12-24562

  6. Modeling Fission Product Sorption in Graphite Structures

    SciTech Connect (OSTI)

    Szlufarska, Izabela; Morgan, Dane; Allen, Todd

    2013-04-08

    The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high- temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products on each type of graphite site. The model will include multiple simultaneous adsorbing species, which will allow for competitive adsorption effects between different fission product species and O and OH (for modeling accident conditions).

  7. AGC-2 Graphite Preirradiation Data Analysis Report

    SciTech Connect (OSTI)

    William Windes; W. David Swank; David Rohrbaugh; Joseph Lord

    2013-08-01

    This report described the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the second Advanced Graphite Capsule (AGC-2) irradiation capsule. The AGC-2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. Similar to the AGC-1 specimen pre-irradiation examination report, material property tests were conducted on specimens from 18 nuclear graphite types but on an increased number of specimens (512) prior to loading into the AGC-2 irradiation assembly. All AGC-2 specimen testing was conducted at Idaho National Laboratory (INL) from October 2009 to August 2010. This report also details the specimen loading methodology for the graphite specimens inside the AGC-2 irradiation capsule. The AGC-2 capsule design requires “matched pair” creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-2 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce “matched pairs” of graphite samples above and below the AGC-2 capsule elevation mid-point to provide specimens with similar neutron dose levels.

  8. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    SciTech Connect (OSTI)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  9. Chapter 20: Graphite

    SciTech Connect (OSTI)

    Burchell, Timothy D

    2012-01-01

    Graphite is truly a unique material. Its structure, from the nano- to the millimeter scale give it remarkable properties that lead to numerous and diverse applications. Graphite bond anisotropy, with strong in-plane covalent bonds and weak van der Waals type bonding between the planes, gives graphite its unique combination of properties. Easy shear of the crystal, facilitated by weak interplaner bonds allows graphite to be used as a dry lubricant, and is responsible for the substances name! The word graphite is derived from the Greek to write because of graphites ability to mark writing surfaces. Moreover, synthetic graphite contains within its structure, porosity spanning many orders of magnitude in size. The thermal closure of these pores profoundly affects the properties for example, graphite strength increases with temperature to temperatures in excess of 2200 C. Consequently, graphite is utilized in many high temperature applications. The basic physical properties of graphite are reviewed here. Graphite applications include metallurgical; (aluminum and steel production), single crystal silicon production, and metal casting; electrical (motor brushes and commutators); mechanical (seals, bearings and bushings); and nuclear applications, (see Chapter 91, Nuclear Graphite). Here we discuss the structure, manufacture, properties, and applications of Graphite.

  10. Fission Product Sorptivity in Graphite

    SciTech Connect (OSTI)

    Tompson, Jr., Robert V.; Loyalka, Sudarshan; Ghosh, Tushar; Viswanath, Dabir; Walton, Kyle; Haffner, Robert

    2015-04-01

    Both adsorption and absorption (sorption) of fission product (FP) gases on/into graphite are issues of interest in very high temperature reactors (VHTRs). In the original proposal, we proposed to use packed beds of graphite particles to measure sorption at a variety of temperatures and to use an electrodynamic balance (EDB) to measure sorption onto single graphite particles (a few ?m in diameter) at room temperature. The use of packed beds at elevated temperature is not an issue. However, the TPOC requested revision of this initial proposal to included single particle measurements at elevated temperatures up to 1100 °C. To accommodate the desire of NEUP to extend the single particle EDB measurements to elevated temperatures it was necessary to significantly revise the plan and the budget. These revisions were approved. In the EDB method, we levitate a single graphite particle (the size, surface characteristics, morphology, purity, and composition of the particle can be varied) or agglomerate in the balance and measure the sorption of species by observing the changes in mass. This process involves the use of an electron stepping technique to measure the total charge on a particle which, in conjunction with the measured suspension voltages for the particle, allows for determinations of mass and, hence, of mass changes which then correspond to measurements of sorption. Accommodating elevated temperatures with this type of system required a significant system redesign and required additional time that ultimately was not available. These constraints also meant that the grant had to focus on fewer species as a result. Overall, the extension of the original proposed single particle work to elevated temperatures added greatly to the complexity of the proposed project and added greatly to the time that would eventually be required as well. This means that the bulk of the experimental progress was made using the packed bed sorption systems. Only being able to recruit one graduate student meant that data acquisition with the packed bed systems ended up competing for the graduate student’s available time with the electrodynamic balance redesign and assembly portions of the project. This competition for available time was eventually mitigated to some extent by the later recruitment of an undergraduate student to help with data collection using the packed bed system. It was only the recruitment of the second student that allowed the single particle balance design and construction efforts to proceed as far as they did during the project period. It should be added that some significant time was also spent by the graduate student cataloging previous work involving graphite. This eventually resulted in a review paper being submitted and accepted (“Adsorption of Iodine on Graphite in High Temperature Gas-Cooled Reactor Systems: A Review,” Kyle L. Walton, Tushar K. Ghosh, Dabir S. Viswanath, Sudarshan K. Loyalka, Robert V. Tompson). Our specific revised objectives in this project were as follows: Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using an EDB and a temperature controlled EDB; Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using a packed column bed apparatus; Explore the effect that charge has on the adsorption isotherms of iodine by varying the charges on and the voltages used to suspend the microscopic particles in the EDB; and To interpret these results in terms of the existing models (Langmuir, BET, Freundlich, and others) which we will modify as necessary to include charge related effects.

  11. Nondestructive evaluation of nuclear-grade graphite

    SciTech Connect (OSTI)

    Kunerth, D. C.; McJunkin, T. R. [Idaho National Laboratory, P.O. Box 1625, MS 2209, Idaho Falls, ID 83415-2209 (United States)

    2012-05-17

    The material of choice for the core of the high-temperature gas-cooled reactors being developed by the U.S. Department of Energy's Next Generation Nuclear Plant Program is graphite. Graphite is a composite material whose properties are highly dependent on the base material and manufacturing methods. In addition to the material variations intrinsic to the manufacturing process, graphite will also undergo changes in material properties resulting from radiation damage and possible oxidation within the reactor. Idaho National Laboratory is presently evaluating the viability of conventional nondestructive evaluation techniques to characterize the material variations inherent to manufacturing and in-service degradation. Approaches of interest include x-ray radiography, eddy currents, and ultrasonics.

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    This EA evaluates the environmental impacts of a proposal to accept 409 spent fuel elements from eight foreign research reactors in seven European countries.  The spent fuel would be shipped across...

  13. Two U.S. University Research Reactors to be Converted From Highly...

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

    of Energy (DOE) has begun to convert research reactors from using highly-enriched uranium (HEU) to low-enriched uranium fuel (LEU) at the University of Florida and Texas A&M...

  14. Sensitivity and uncertainty analyses for thermo-hydraulic calculation of research reactor

    SciTech Connect (OSTI)

    Hartini, Entin; Andiwijayakusuma, Dinan [Center for Development of Nuclear Informatics - National Nuclear Energy Agency PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)] [Center for Development of Nuclear Informatics - National Nuclear Energy Agency PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia); Isnaeni, Muh Darwis [Center for Reactor Technology and Nuclear Safety- National Nuclear Energy Agency PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)] [Center for Reactor Technology and Nuclear Safety- National Nuclear Energy Agency PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)

    2013-09-09

    The sensitivity and uncertainty analysis of input parameters on thermohydraulic calculations for a research reactor has successfully done in this research. The uncertainty analysis was carried out on input parameters for thermohydraulic calculation of sub-channel analysis using Code COOLOD-N. The input parameters include radial peaking factor, the increase bulk coolant temperature, heat flux factor and the increase temperature cladding and fuel meat at research reactor utilizing plate fuel element. The input uncertainty of 1% - 4% were used in nominal power calculation. The bubble detachment parameters were computed for S ratio (the safety margin against the onset of flow instability ratio) which were used to determine safety level in line with the design of 'Reactor Serba Guna-G. A. Siwabessy' (RSG-GA Siwabessy). It was concluded from the calculation results that using the uncertainty input more than 3% was beyond the safety margin of reactor operation.

  15. Pyrolytic graphite production : automation of material placement

    E-Print Network [OSTI]

    Olle, Chase R

    2014-01-01

    This research examines the process and challenges associated with the addition of an autonomous transfer robot to a manufacturing line for AvCarb Material Solutions for use in production of pyrolytic graphite. Development ...

  16. Thermal hydraulic limits analysis for the MIT Research Reactor low enrichment uranium core conversion using statistical propagation of parametric uncertainties

    E-Print Network [OSTI]

    Chiang, Keng-Yen

    2012-01-01

    The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design from 15 to 18 plates per element, a reactor ...

  17. MODULAR PEBBLE BED REACTOR PROJECT UNIVERSITY RESEARCH CONSORTIUM

    E-Print Network [OSTI]

    includes the development of a fission gas release model, particle temperature distributions, internal conceptual design for a gas-cooled, modular, pebble bed reactor. Key technology areas associated particle pressure, migration of fission products, and chemical attack of fuel particle layers. · A balance

  18. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    SciTech Connect (OSTI)

    Not Available

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  19. Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania

    SciTech Connect (OSTI)

    K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin; C. Paunoiu; M. Ciocanescu

    2010-03-01

    In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.

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

    SciTech Connect (OSTI)

    Vinson, D.

    2010-07-11

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

  1. Golden Mountains The Altai Republic in the Russian Federation Kyoto University Research Reactor Institute Kobayashi Tooru

    E-Print Network [OSTI]

    Taskaev, Sergey Yur'evich

    1 Golden Mountains ­ The Altai Republic in the Russian Federation Kyoto University Research Reactor Institute Kobayashi Tooru It is said that Altai also means golden mountains. The development to these mountains easier especially during the warm summer season. Being geographically located at the central

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

    SciTech Connect (OSTI)

    Not Available

    1994-04-01

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

  3. Research and Medical Isotope Reactor Supply | Y-12 National Security

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration wouldMassR&D100 WinnersAffiliatesMadden-JulianOut withResearch Sheds LightComplex

  4. Understanding Creep Mechanisms in Graphite with Experiments, Multiscale Simulations, and Modeling

    SciTech Connect (OSTI)

    Eapen, Jacob; Murty, Korukonda; Burchell, Timothy

    2014-06-02

    Disordering mechanisms in graphite have a long history with conflicting viewpoints. Using Raman and x-ray photon spectroscopy, electron microscopy, x-ray diffraction experiments and atomistic modeling and simulations, the current project has developed a fundamental understanding of early-to-late state radiation damage mechanisms in nuclear reactor grade graphite (NBG-18 and PCEA). We show that the topological defects in graphite play an important role under neutron and ion irradiation.

  5. Development of Regulatory Technical Requirements for the Advanced Integral Type Research Reactor

    SciTech Connect (OSTI)

    Jo, Jong Chull; Yune, Young Gill; Kim, Woong Sik; Kim, Hho Jung

    2004-07-01

    This paper presents the current status of the study on the development of regulatory technical requirements for the licensing review of an advanced integral type research reactor of which the license application is expected in a few years. According to the Atomic Energy Act of Korea, both research and education reactors are subject to the technical requirements for power reactors in the licensing review. But, some of the requirements may not be applicable or insufficient for the licensing reviews of reactors with unique design features. Thus it is necessary to identify which review topics or areas can not be addressed by the existing requirements and to develop the required ones newly or supplement appropriately. Through the study performed so far, it has been identified that the following requirements need to be developed newly for the licensing review of SMART-P: the use of proven technology, the interfacial facility, the non-safety systems, and the metallic fuels. The approach and basis for the development of each of the requirements are discussed. (authors)

  6. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    SciTech Connect (OSTI)

    Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  7. U.S. Department of Energy Instrumentation and Controls Technology Research for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Wood, Richard Thomas

    2012-01-01

    Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD&D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, key DOE programs have substantial ICHMI RD&D elements to their respective research portfolio. This article describes current ICHMI research to support the development of advanced small modular reactors.

  8. Carbon Characterization Laboratory Readiness to Receive Irradiated Graphite Samples

    SciTech Connect (OSTI)

    Karen A. Moore

    2011-05-01

    The Carbon Characterization Laboratory (CCL) is located in Labs C19 and C20 of the Idaho National Laboratory Research Center. The CCL was established under the Next Generation Nuclear Plant Project to support graphite and ceramic composite research and development activities. The research conducted in this laboratory will support the Advanced Graphite Creep experiments—a major series of material irradiation experiments within the Next Generation Nuclear Plant Graphite program. The CCL is designed to characterize and test low activated irradiated materials such as high purity graphite, carbon-carbon composites, silicon-carbide composite, and ceramic materials. The laboratory is fully capable of characterizing material properties for both irradiated and nonirradiated materials. Major infrastructural modifications were undertaken to support this new radiological facility at Idaho National Laboratory. Facility modifications are complete, equipment has been installed, radiological controls and operating procedures have been established and work management documents have been created to place the CCL in readiness to receive irradiated graphite samples.

  9. Graphite Sublimation Tests for the Muon Collider/Neutrino Factory

    E-Print Network [OSTI]

    McDonald, Kirk

    cooled graphite target was proposed for a 1.5 MW neutrino production research facility because of its simplicity and favorable performance as a target material for neutrino production (Ref. 1). The conceptual handling performance of radiatively cooled graphite targets, a helium cover gas at nominally one atmosphere

  10. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    SciTech Connect (OSTI)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  11. A simple setup for neutron tomography at the Portuguese Nuclear Research Reactor

    E-Print Network [OSTI]

    M. A. Stanojev Pereira; J. G. Marques; R. Pugliesi

    2012-05-15

    A simple setup for neutron radiography and tomography was recently installed at the Portuguese Research Reactor. The objective of this work was to determine the operational characteristics of the installed setup, namely the irradiation time to obtain the best dynamic range for individual images and the spatial resolution. The performance of the equipment was demonstrated by imaging a fragment of a 17th century decorative tile.

  12. Graphite and Beryllium Reflector Critical Assemblies of UO2 (Benchmark Experiments 2 and 3)

    SciTech Connect (OSTI)

    Margaret A. Marshall; John D. Bess

    2012-11-01

    INTRODUCTION A series of experiments was carried out in 1962-65 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for use in space reactor research programs. A core containing 93.2 wt% enriched UO2 fuel rods was used in these experiments. The first part of the experimental series consisted of 252 tightly-packed fuel rods (1.27-cm triangular pitch) with graphite reflectors [1], the second part used 252 graphite-reflected fuel rods organized in a 1.506-cm triangular-pitch array [2], and the final part of the experimental series consisted of 253 beryllium-reflected fuel rods in a 1.506-cm-triangular-pitch configuration and in a 7-tube-cluster configuration [3]. Fission rate distribution and cadmium ratio measurements were taken for all three parts of the experimental series. Reactivity coefficient measurements were taken for various materials placed in the beryllium reflected core. All three experiments in the series have been evaluated for inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP) [4] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbooks, [5]. The evaluation of the first experiment in the series was discussed at the 2011 ANS Winter meeting [6]. The evaluations of the second and third experiments are discussed below. These experiments are of interest as benchmarks because they support the validation of compact reactor designs with similar characteristics to the design parameters for a space nuclear fission surface power systems [7].

  13. Decommissioning of German Research Reactors Under the Governance of the Federal Ministry of Education and Research - 12154

    SciTech Connect (OSTI)

    Weigl, M. [Karlsruhe Institute of Technology, Projekttraeger Karlsruhe (PTKA-WTE), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2012-07-01

    Since 1956, nuclear research and development (R and D) in Germany has been supported by the Federal Government. The goal was to help German industry to become competitive in all fields of nuclear technology. National research centers were established and demonstration plants were built. In the meantime, all these facilities were shut down and are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactor with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. Another big project was finished in 2008. The Forschungs-Reaktor Juelich 1 (FRJ1), a research reactor with a thermal power of 10 MW was completely dismantled and in September 2008 an oak tree was planted on a green field at the site, where the FRJ1 was standing before. This is another example for German success in the field of D and D. Within these projects a lot of new solutions and innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). Some examples are underwater-cutting technologies like plasma arc cutting and contact arc metal cutting. This clearly shows that research on the field of D and D is important for the future. Moreover, these research activities are important to save the know-how in nuclear engineering in Germany and will enable enterprises to compete on the increasing market of D and D services. The author assumes that an efficient decommissioning of nuclear installations will help stabilize the credibility of nuclear energy. Some critics of nuclear energy are insisting that a return to 'green field sites' is not possible. The successful completion of two big D and D projects (HDR and KKN), which reached green field conditions, are showing quite the contrary. Moreover, research on D and D technologies offers the possibility to educate students on a field of nuclear technology, which will be very important in the future. In these days D and D companies are seeking for a lot of young engineers and this will not change in the coming years. (authors)

  14. Recompressed exfoliated graphite articles

    DOE Patents [OSTI]

    Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

    2013-08-06

    This invention provides an electrically conductive, less anisotropic, recompressed exfoliated graphite article comprising a mixture of (a) expanded or exfoliated graphite flakes; and (b) particles of non-expandable graphite or carbon, wherein the non-expandable graphite or carbon particles are in the amount of between about 3% and about 70% by weight based on the total weight of the particles and the expanded graphite flakes combined; wherein the mixture is compressed to form the article having an apparent bulk density of from about 0.1 g/cm.sup.3 to about 2.0 g/cm.sup.3. The article exhibits a thickness-direction conductivity typically greater than 50 S/cm, more typically greater than 100 S/cm, and most typically greater than 200 S/cm. The article, when used in a thin foil or sheet form, can be a useful component in a sheet molding compound plate used as a fuel cell separator or flow field plate. The article may also be used as a current collector for a battery, supercapacitor, or any other electrochemical cell.

  15. X-10 Graphite Reactor | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal Gas &SCE-Sessions |discussed how savingDepartment ofFinalCapitan AsWes Kelley

  16. X-10 Graphite Reactor | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative FuelsofProgram: Report1538-1950DepartmentWaveWind(NREL)Zappos.comDepartmentAirportX-10

  17. Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion

    E-Print Network [OSTI]

    Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

    2008-01-01

    The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the ...

  18. NUMERICAL SIMULATION FOR MECHANICAL BEHAVIOR OF U10MO MONOLITHIC MINIPLATES FOR RESEARCH AND TEST REACTORS

    SciTech Connect (OSTI)

    Hakan Ozaltun & Herman Shen

    2011-11-01

    This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

  19. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    SciTech Connect (OSTI)

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  20. Improved graphite furnace atomizer

    DOE Patents [OSTI]

    Siemer, D.D.

    1983-05-18

    A graphite furnace atomizer for use in graphite furnace atomic absorption spectroscopy is described wherein the heating elements are affixed near the optical path and away from the point of sample deposition, so that when the sample is volatilized the spectroscopic temperature at the optical path is at least that of the volatilization temperature, whereby analyteconcomitant complex formation is advantageously reduced. The atomizer may be elongated along its axis to increase the distance between the optical path and the sample deposition point. Also, the atomizer may be elongated along the axis of the optical path, whereby its analytical sensitivity is greatly increased.

  1. Sandia National Laboratories: Research: Research Foundations...

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

    of gamma-ray radiation. The Annular Core Research Reactor (ACRR) is used for reactor-driven laser experiments, space reactor fuels development, pulse reactor kinetics, reactor...

  2. Diamond-graphite field emitters

    DOE Patents [OSTI]

    Valone, Steven M. (Santa Fe, NM)

    1997-01-01

    A field emission electron emitter comprising an electrode of diamond and a conductive carbon, e.g., graphite, is provided.

  3. RELAP5 Application to Accident Analysis of the NIST Research Reactor

    SciTech Connect (OSTI)

    Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

    2012-03-18

    Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

  4. Graphite Fiber Brush Anodes for Increased Power Production in

    E-Print Network [OSTI]

    Graphite Fiber Brush Anodes for Increased Power Production in Air-Cathode Microbial Fuel Cells B R a conductive, but noncorrosive metal core, were examined for power production in cube (C-MFC) and bottle (B-MFC) air-cathode MFCs. Power production in C-MFCs containing brush electrodes at 9600 m2/m3 reactor volume

  5. An American company thinks it can have a commercial reactor ready and working within a decade ONE of the clichs of nuclear-power research is that a commercial fusion reactor is only 30 years away, and always1

    E-Print Network [OSTI]

    South Bohemia, University of

    of the clichés of nuclear-power research is that a commercial fusion reactor is only 30 years away, and always1An American company thinks it can have a commercial reactor ready and working within a decade ONE that Lockheed Martin,2 a big American engineering and defence company, has a new design for a fusion reactor

  6. Graphite-based photovoltaic cells

    DOE Patents [OSTI]

    Lagally, Max (Madison, WI); Liu, Feng (Salt Lake City, UT)

    2010-12-28

    The present invention uses lithographically patterned graphite stacks as the basic building elements of an efficient and economical photovoltaic cell. The basic design of the graphite-based photovoltaic cells includes a plurality of spatially separated graphite stacks, each comprising a plurality of vertically stacked, semiconducting graphene sheets (carbon nanoribbons) bridging electrically conductive contacts.

  7. Carbide Coatings for Nickel Alloys, Graphite and Carbon/Carbon Composites to be used in Fluoride Salt Valves

    SciTech Connect (OSTI)

    Nagle, Denis; Zhang, Dajie

    2015-10-22

    The focus of this research was concerned with developing materials technology that supports the evolution of Generation IV Advanced High Temperature Reactor (AHTR) concepts. Specifically, we investigate refractory carbide coatings for 1) nickel alloys, and 2) commercial carbon-carbon composites (CCCs). Numerous compelling reasons have driven us to focus on carbon and carbide materials. First, unlike metals, the strength and modulus of CCCs increase with rising temperature. Secondly, graphite and carbon composites have been proven effective for resisting highly corrosive fluoride melts such as molten cryolite [Na3AlF6] at ~1000oC in aluminum reduction cells. Thirdly, graphite and carbide materials exhibit extraordinary radiation damage tolerance and stability up to 2000°C. Finally, carbides are thermodynamically more stable in liquid fluoride salt than the corresponding metals (i.e. Cr and Zr) found in nickel based alloys.

  8. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    SciTech Connect (OSTI)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  9. System Upgrades at the Advanced Test Reactor Help Ensure that Nuclear Energy Research Continues at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Craig Wise

    2011-12-01

    Fully operational in 1967, the Advanced Test Reactor (ATR) is a first-of-its-kind materials test reactor. Located on the Idaho National Laboratory’s desert site, this reactor remains at the forefront of nuclear science, producing extremely high neutron irradiation in a relatively short time span. The Advanced Test Reactor is also the only U.S. reactor that can replicate multiple reactor environments concurrently. The Idaho National Laboratory and the Department of Energy recently invested over 13 million dollars to replace three of ATR’s instrumentation and control systems. The new systems offer the latest software and technology advancements, ensuring the availability of the reactor for future energy research. Engineers and project managers successfully completed the four year project in March while the ATR was in a scheduled maintenance outage. “These new systems represent state-of-the-art monitoring and annunciation capabilities,” said Don Feldman, ATR Station Manager. “They are comparable to systems currently used for advanced reactor designs planned for construction in the U.S. and in operation in some foreign countries.”

  10. Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability

    SciTech Connect (OSTI)

    Philip E. MacDonald

    2003-09-01

    Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment. • Reactor pressure vessel • Pumps and piping

  11. Tritium Related Material Research -Irradiation Effect on Isotropic...

    Office of Environmental Management (EM)

    Related Material Research -Irradiation Effect on Isotropic Graphite Utilizing Heavy Ion-Irradiation- Tritium Related Material Research -Irradiation Effect on Isotropic Graphite...

  12. NOVEL CRYOGENIC ENGINEERING SOLUTIONS FOR THE NEW AUSTRALIAN RESEARCH REACTOR OPAL

    SciTech Connect (OSTI)

    Olsen, S. R.; Kennedy, S. J.; Kim, S.; Schulz, J. C.; Thiering, R.; Gilbert, E. P.; Lu, W.; James, M.; Robinson, R. A.

    2008-03-16

    In August 2006 the new 20MW low enriched uranium research reactor OPAL went critical. The reactor has 3 main functions, radio pharmaceutical production, silicon irradiation and as a neutron source. Commissioning on 7 neutron scattering instruments began in December 2006. Three of these instruments (Small Angle Neutron Scattering, Reflectometer and Time-of-flight Spectrometer) utilize cold neutrons.The OPAL Cold Neutron Source, located inside the reactor, is a 20L liquid deuterium moderated source operating at 20K, 330kPa with a nominal refrigeration capacity of 5 kW and a peak flux at 4.2meV (equivalent to a wavelength of 0.4nm). The Thermosiphon and Moderator Chamber are cooled by helium gas delivered at 19.8K using the Brayton cycle. The helium is compressed by two 250kW compressors (one with a variable frequency drive to lower power consumption).A 5 Tesla BSCCO (2223) horizontal field HTS magnet will be delivered in the 2{sup nd} half of 2007 for use on all the cold neutron instruments. The magnet is cooled by a pulse tube cryocooler operating at 20K. The magnet design allows for the neutron beam to pass both axially and transverse to the field. Samples will be mounted in a 4K to 800K Gifford-McMahon (GM) cryofurnace, with the ability to apply a variable electric field in-situ. The magnet is mounted onto a tilt stage. The sample can thus be studied under a wide variety of conditions.A cryogen free 7.4 Tesla Nb-Ti vertical field LTS magnet, commissioned in 2005 will be used on neutron diffraction experiments. It is cooled by a standard GM cryocooler operating at 4.2K. The sample is mounted in a 2{sup nd} GM cryocooler (4K-300K) and a variable electric field can be applied.

  13. Heat exchanger using graphite foam

    DOE Patents [OSTI]

    Campagna, Michael Joseph; Callas, James John

    2012-09-25

    A heat exchanger is disclosed. The heat exchanger may have an inlet configured to receive a first fluid and an outlet configured to discharge the first fluid. The heat exchanger may further have at least one passageway configured to conduct the first fluid from the inlet to the outlet. The at least one passageway may be composed of a graphite foam and a layer of graphite material on the exterior of the graphite foam. The layer of graphite material may form at least a partial barrier between the first fluid and a second fluid external to the at least one passageway.

  14. Development of a Scanning Microscale Fast Neutron Irradiation Platform for Examining the Correlation Between Local Neutron Damage and Graphite Microstructure

    SciTech Connect (OSTI)

    Pinhero, Patrick; Windes, William

    2015-03-10

    The fast particle radiation damage effect of graphite, a main material in current and future nuclear reactors, has significant influence on the utilization of this material in fission and fusion plants. Atoms on graphite crystals can be easily replaced or dislocated by fast protons and result in interstitials and vacancies. The currently accepted model indicates that after most of the interstitials recombine with vacancies, surviving interstitials form clusters and furthermore gather to create loops with each other between layers. Meanwhile, surviving vacancies and interstitials form dislocation loops on the layers. The growth of these inserted layers cause the dimensional increase, i.e. swelling, of graphite. Interstitial and vacancy dislocation loops have been reported and they can easily been observed by electron microscope. However, observation of the intermediate atom clusters becomes is paramount in helping prove this model. We utilize fast protons generated from the University of Missouri Research Reactor (MURR) cyclotron to irradiate highly- oriented pyrolytic graphite (HOPG) as target for this research. Post-irradiation examination (PIE) of dosed targets with high-resolution transmission electron microscopy (HRTEM) has permit observation and analysis of clusters and dislocation loops to support the proposed theory. Another part of the research is to validate M.I. Heggie’s Ruck and Tuck model, which introduced graphite layers may fold under fast particle irradiation. Again, we employed microscopy to image irradiated specimens to determine how the extent of Ruck and Tuck by calculating the number of folds as a function of dose. Our most significant accomplishment is the invention of a novel class of high-intensity pure beta-emitters for long-term lightweight batteries. We have filed four invention disclosure records based on the research conducted in this project. These batteries are lightweight because they consist of carbon and tritium and can be fabricated to conform to many geometric shapes. In addition, we have published eight peer-reviewed American Nuclear Society (ANS) transactions, and presented our findings at ANS National Meetings, and several universities.

  15. Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

    SciTech Connect (OSTI)

    Chang Ho Oh; Eung Soo Kim; Hee Cheon No; Nam Zin Cho

    2008-12-01

    The US Department of Energy is performing research and development (R&D) that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP) Program / GEN-IV Very High Temperature Reactor (VHTR). Phenomena identification and ranking studies (PIRT) to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Schultz et al., 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) are very high priority for the NGNP program. Following a loss of coolant and system depressurization, air will enter the core through the break. Air ingress leads to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heat-up of the bottom reflector and the reactor core and will cause the release of fission products eventually. The potential collapse of the bottom reflector because of burn-off and the release of CO lead to serious safety problems. For estimation of the proper safety margin we need experimental data and tools, including accurate multi-dimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model. We also need to develop effective strategies to mitigate the effects of oxidation. The results from this research will provide crucial inputs to the INL NGNP/VHTR Methods R&D project. This project is focused on (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the bottom reflector, (d) structural tests of the burnt-off bottom reflector, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i) verification and validation of the coupled models.

  16. Sept.. 1060 HYDROGENS O R P T I O N O X GRAPHITEAT E L E V A T E D TEMPERATURES 1093 HYDROGEN SORPTIOK ON GRAPHITE AT ELEVATED

    E-Print Network [OSTI]

    Sept.. 1060 HYDROGENS O R P T I O N O X GRAPHITEAT E L E V A T E D TEMPERATURES 1093 HYDROGEN SORPTIOK ON GRAPHITE AT ELEVATED BYJ. P. REDMONDAND 1'. L. W..ZLHER,JR. Lkpaitmcnt of Fuel Technology does n>toccur. Introduction drtificial graphite of use in graphite-moderated nuclear reactors

  17. Neutronic safety and transient analyses for potential LEU conversion of the IR-8 research reactor.

    SciTech Connect (OSTI)

    Deen, J. R.; Hanan, N. A.; Smith, R. S.; Matos, J. E.; Egorenkov, P. M.; Nasonov, V. A.

    1999-09-27

    Kinetic parameters, isothermal reactivity feedback coefficients and three transients for the IR-8 research reactor cores loaded with either HEU(90%), HEU(36%), or LEU (19.75%) fuel assemblies (FA) were calculated using three dimensional diffusion theory flux solutions, RELAP5/MOD3.2 and PARET. The prompt neutron generation time and effective delayed neutron fractions were calculated for fresh and beginning-of-equilibrium-cycle cores. Isothermal reactivity feedback coefficients were calculated for changes in coolant density, coolant temperature and fuel temperature in fresh and equilibrium cores. These kinetic parameters and reactivity coefficients were used in transient analysis models to predict power histories, and peak fuel, clad and coolant temperatures. The transients modeled were a rapid and slow loss-of-flow, a slow reactivity insertion, and a fast reactivity insertion.

  18. The progress of neutron texture diffractometer at China Advanced Research Reactor

    E-Print Network [OSTI]

    Li, MeiJuan; Liu, YunTao; Tian, GengFang; Gao, JianBo; Yu, ZhouXiang; Li, YuQing; Wu, LiQi; Yang, LinFeng; Sun, Kai; Wang, HongLi; Chen, DongFeng

    2015-01-01

    The first neutron texture diffractometer in China has been built at China Advanced Research Reactor due to the strong demands of texture measurement with neutrons from domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study the texture in the commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of Zr alloy plate are presented. The comparison of texture measurement among different neutron texture diffractometer of HIPPO at LANSCE, Kowari at ANSTO and neutron texture diffractometer at CARR illustrates the reliable performance of this texture diffractometer.

  19. Graphitic packing removal tool

    DOE Patents [OSTI]

    Meyers, Kurt Edward (Avella, PA); Kolsun, George J. (Pittsburgh, PA)

    1997-01-01

    Graphitic packing removal tools for removal of the seal rings in one piece. he packing removal tool has a cylindrical base ring the same size as the packing ring with a surface finish, perforations, knurling or threads for adhesion to the seal ring. Elongated leg shanks are mounted axially along the circumferential center. A slit or slits permit insertion around shafts. A removal tool follower stabilizes the upper portion of the legs to allow a spanner wrench to be used for insertion and removal.

  20. Graphitic packing removal tool

    DOE Patents [OSTI]

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

    1997-11-11

    Graphitic packing removal tools for removal of the seal rings in one piece are disclosed. The packing removal tool has a cylindrical base ring the same size as the packing ring with a surface finish, perforations, knurling or threads for adhesion to the seal ring. Elongated leg shanks are mounted axially along the circumferential center. A slit or slits permit insertion around shafts. A removal tool follower stabilizes the upper portion of the legs to allow a spanner wrench to be used for insertion and removal. 5 figs.

  1. Understanding Interfaces in Metal-Graphitic Hybrid Nanostructures”

    SciTech Connect (OSTI)

    Ding, Mengning; Tang, Yifan; Star, Alexander

    2013-01-03

    Metal–graphitic interfaces formed between metal nanoparticles (MNPs) and carbon nanotubes (CNTs) or graphene play an important role in the properties of such hybrid nanostructures. This Perspective summarizes different types of interfaces that exist within the metal–carbon nanoassemblies and discusses current efforts on understanding and modeling the interfacial conditions and interactions. Characterization of the metal–graphitic interfaces is described here, including microscopy, spectroscopy, electrochemical techniques, and electrical measurements. Recent studies on these nanohybrids have shown that the metal–graphitic interfaces play critical roles in both controlled assembly of nanoparticles and practical applications of nanohybrids in chemical sensors and fuel cells. Better understanding, design, and manipulation of metal–graphitic interfaces could therefore become the new frontier in the research of MNP/CNT or MNP/graphene hybrid systems.

  2. Dismantling Structures and Equipment of the MR Reactor and its Loop Facilities at the National Research Center 'Kurchatov Institute' - 12051

    SciTech Connect (OSTI)

    Volkov, V.G.; Danilovich, A.S.; Zverkov, Yu. A.; Ivanov, O.P.; Kolyadin, V.I.; Lemus, A.V.; Muzrukova, V.D.; Pavlenko, V.I.; Semenov, S.G.; Fadin, S.Yu.; Shisha, A.D.; Chesnokov, A.V.

    2012-07-01

    In 2008 a design of decommissioning of research reactors MR and RFT has been developed in the National research Center 'Kurchatov institute'. The design has been approved by Russian State Authority in July 2009 year and has received the positive conclusion of ecological expertise. In 2009-2010 a preparation for decommissioning of reactors MR and RFT was spent. Within the frames of a preparation a characterization, sorting and removal of radioactive objects, including the irradiated fuel, from reactor storage facilities and pool have been executed. During carrying out of a preparation on removal of radioactive objects from reactor sluice pool water treating has been spent. For these purposes modular installation for clearing and processing of a liquid radioactive waste 'Aqua - Express' was used. As a result of works it was possible to lower volume activity of water on three orders in magnitude that has allowed improving essentially of radiating conditions in a reactor hall. Auxiliary systems of ventilation, energy and heat supplies, monitoring systems of radiating conditions of premises of the reactor and its loop-back installations are reconstructed. In 2011 the license for a decommissioning of the specified reactors has been received and there are begun dismantling works. Within the frames of works under the design the armature and pipelines are dismantled in a under floor space of a reactor hall where a moving and taking away pipelines of loop facilities and the first contour of the MR reactor were replaced. A dismantle of the main equipment of loop facility with the gas coolant has been spent. Technologies which were used on dismantle of the radioactive contaminated equipment are presented, the basic works on reconstruction of systems of maintenance of on the decommissioning works are described, the sequence of works on the decommissioning of reactors MR and RFT is shown. Dismantling works were carried out with application of means of a dust suppression that, in aggregate with standard means at such works of individual protection of the personnel and devices of radiating control, has allowed to lower risk of action of radiation on the personnel, the population and environment at the expense of reduction of volume activity of radioactive aerosols in air. (authors)

  3. Testing of Small Graphite Samples for Nuclear Qualification

    SciTech Connect (OSTI)

    Julie Chapman

    2010-11-01

    Accurately determining the mechanical properties of small irradiated samples is crucial to predicting the behavior of the overal irradiated graphite components within a Very High Temperature Reactor. The sample size allowed in a material test reactor, however, is limited, and this poses some difficulties with respect to mechanical testing. In the case of graphite with a larger grain size, a small sample may exhibit characteristics not representative of the bulk material, leading to inaccuracies in the data. A study to determine a potential size effect on the tensile strength was pursued under the Next Generation Nuclear Plant program. It focuses first on optimizing the tensile testing procedure identified in the American Society for Testing and Materials (ASTM) Standard C 781-08. Once the testing procedure was verified, a size effect was assessed by gradually reducing the diameter of the specimens. By monitoring the material response, a size effect was successfully identified.

  4. Oxidation of PCEA nuclear graphite by low water concentrations in helium

    SciTech Connect (OSTI)

    Contescu, Cristian I; Mee, Robert; Wang, Peng; Romanova, Anna V; Burchell, Timothy D

    2014-10-01

    Accelerated oxidation tests were performed to determine kinetic parameters of the chronic oxidation reaction of PCEA graphite in contact with helium coolant containing low moisture concentrations in high temperature gas-cooled reactors. To the authors best knowledge such a study has not been done since the detailed analysis of reaction of H-451 graphite with steam [Velasquez, Hightower, Burnette, 1978]. Since that H-451 graphite is now unavailable, it is urgently needed to characterize chronic oxidation behavior of new graphite grades under qualification for gas-cooled reactors. The Langmuir-Hinshelwood mechanism of carbon oxidation by water results in a non-linear reaction rate expression, with at least six different parameters. They were determined in accelerated oxidation experiments that covered a large range of temperatures (800 to 1100 oC), and partial pressures of water (15 to 850 Pa) and hydrogen (30 to 150 Pa) and used graphite specimens thin enough (4 mm) in order to avoid diffusion effects. Data analysis employed a statistical method based on multiple likelihood estimation of parameters and simultaneous fitting of non-linear equations. The results show significant material-specific differences between graphite grades PCEA and H-451 which were attributed to microstructural dissimilarity of the two materials. It is concluded that kinetic data cannot be transferred from one graphite grade to another.

  5. Reactor safety research programs. Quarterly report, July 1-September 30, 1979

    SciTech Connect (OSTI)

    Hooper, J.L.

    1980-03-01

    The programs include: ultimate heat sink performance measurement; experimental verification of steady state codes: Task A - irradiation results and Task C - code development; graphite nondestructive testing; acoustic emission-flaw relationship for in-service monitoring of nuclear pressure vessels; fuel subassembly procurement and irradiation test program; report of resident engineer at Cadarache, France; core thermal model development; integration of nondestructive examination reliability and fracture mechanics; and steam generator tube integrity.

  6. INITIAL COMPARISON OF BASELINE PHYSICAL AND MECHANICAL PROPERTIES FOR THE VHTR CANDIDATE GRAPHITE GRADES

    SciTech Connect (OSTI)

    Carroll, Mark C

    2014-09-01

    High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration that is capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is endeavoring to minimize the conservative estimates of as-manufactured mechanical and physical properties in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered “candidate” grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding process. An analysis of the comparison between each of these grades will include not only the differences in fundamental and statistically-significant individual strength levels, but also the differences in variability in properties within each of the grades that will ultimately provide the basis for the prediction of in-service performance. The comparative performance of the different types of nuclear-grade graphites will continue to evolve as thousands more specimens are fully characterized from the numerous grades of graphite being evaluated.

  7. Status of Initial Assessment of Physical and Mechanical Properties of Graphite Grades for NGNP Appkications

    SciTech Connect (OSTI)

    Strizak, Joe P [ORNL; Burchell, Timothy D [ORNL; Windes, Will [Idaho National Laboratory (INL)

    2011-12-01

    Current candidate graphite grades for the core structures of NGNP include grades NBG-17, NBG-18, PCEA and IG-430. Both NBG-17 and NBG-18 are manufactured using pitch coke, and are vibrationally molded. These medium grain products are produced by SGL Carbon SAS (France). Tayo Tanso (Japan) produces IG-430 which is a petroleum coke, isostatically molded, nuclear grade graphite. And PCEA is a medium grain, extruded graphite produced by UCAR Carbon Co. (USA) from petroleum coke. An experimental program has been initiated to develop physical and mechanical properties data for these current candidate graphites. The results will be judged against the requirements for nuclear grade graphites set forth in ASTM standard D 7219-05 "Standard Specification for Isotropic and Near-isotropic Nuclear Graphites". Physical properties data including thermal conductivity and coefficient of thermal expansion, and mechanical properties data including tensile, compressive and flexural strengths will be obtained using the established test methods covered in D-7219 and ASTM C 781-02 "Standard Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear Reactors". Various factors known to effect the properties of graphites will be investigated. These include specimen size, spatial location within a graphite billet, specimen orientation (ag and wg) within a billet, and billet-to-billet variations. The current status of the materials characterization program is reported herein. To date billets of the four graphite grades have been procured, and detailed cut up plans for obtaining the various specimens have been prepared. Particular attention has been given to the traceability of each specimen to its spatial location and orientation within a billet.

  8. Development and operation of research-scale IIIV nanowire growth reactors

    E-Print Network [OSTI]

    Petta, Jason

    ,300,000.18­20 Unfortu- nately, the cost of commercial reactors presents a high barrier of entry into semiconductor,a A. M. Bergman, and J. R. Pettab Department of Physics, Princeton University, Princeton, New Jersey at the nanometer scale. However, the costs associated with commercial nanowire growth reactors are prohibitive

  9. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    SciTech Connect (OSTI)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  10. Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–2013

    SciTech Connect (OSTI)

    Idaho National Laboratory

    2009-12-01

    Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement and leadership on nuclear safety and security issues.

  11. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    SciTech Connect (OSTI)

    Corwin, William R; Burchell, Timothy D; Katoh, Yutai; McGreevy, Timothy E; Nanstad, Randy K; Ren, Weiju; Snead, Lance Lewis; Wilson, Dane F

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOE's structural materials research activities being conducted to support VHTR development. By far, the largest portion of material's R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water rea

  12. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect (OSTI)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.

  13. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect (OSTI)

    Gougar, Hans D.

    2014-10-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.

  14. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    SciTech Connect (OSTI)

    None

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  15. Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor

    E-Print Network [OSTI]

    Bean, Malcolm K.

    2011-08-01

    Molten salt is a promising coolant candidate for Advanced High Temperature Reactor (AHTR) Gen-IV designs. The low neutron absorption, high thermal capacity, chemical inertness, and high boiling point at low pressure of ...

  16. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    SciTech Connect (OSTI)

    Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

    2012-09-30

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

  17. Aerial of Nuclear Science Reactor 

    E-Print Network [OSTI]

    Unknown

    2011-08-17

    Small graphite-moderated and gas-cooled reactors have been around since the beginning of the atomic age. Though their existence in the past has been associated with nuclear weapons programs, they are capable of being used in civilian power programs...

  18. Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II)

    E-Print Network [OSTI]

    Connaway, Heather M. (Heather Moira)

    2012-01-01

    The MIT Research Reactor (MITR-II) is currently undergoing analysis for the planned conversion from high enriched uranium (HEU) to low enriched uranium (LEU), as part of a global effort to minimize the availability of ...

  19. Two Methods for Converting a Heavy-Water Research Reactor to Use Low-Enriched-Uranium Fuel to Improve Proliferation Resistance After Startup

    E-Print Network [OSTI]

    Kemp, R. Scott

    This article demonstrates the feasibility of converting a heavy-water research reactor from natural to low-enriched uranium in order to slow the production of weapon-usable plutonium, even if the core cannot be physically ...

  20. Application of the Isotope Ratio Method to a Boiling Water Reactor

    SciTech Connect (OSTI)

    Frank, Douglas P.; Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Meriwether, George H.; Mitchell, Mark R.; Reid, Bruce D.

    2010-08-11

    The isotope ratio method is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods. All reactor materials contain trace elemental impurities at parts per million levels, and the isotopes of these elements are transmuted by neutron irradiation in a predictable manner. While measuring the change in a particular isotope’s concentration is possible, it is difficult to correlate to energy production because the initial concentration of that element may not be accurately known. However, if the ratio of two isotopes of the same element can be measured, the energy production can then be determined without knowing the absolute concentration of that impurity since the initial natural ratio is known. This is the fundamental principle underlying the isotope ratio method. Extremely sensitive mass-spectrometric methods are currently available that allow accurate measurements of the impurity isotope ratios in samples. Additionally, “indicator” elements with stable activation products have been identified so that their post-irradiation isotope ratios remain constant. This method has been successfully demonstrated on graphite-moderated reactors. Graphite reactors are particularly well-suited to such analyses since the graphite moderator is resident in the fueled region of the core for the entire period of operation. Applying this method to other reactor types is more difficult since the resident portions of the reactor available for sampling are either outside the fueled region of the core or structural components of individual fuel assemblies. The goal of this research is to show that the isotope ratio method can produce meaningful results for light water-moderated power reactors. In this work, we use the isotope ratio method to estimate the energy production in a boiling water reactor fuel bundle based on measurements taken from the corresponding fuel assembly channel. Our preliminary results are in good agreement with the actual operating history of the reactor during the cycle that the fuel bundle was resident in the core.

  1. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    SciTech Connect (OSTI)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  2. Effect of graphite properties in thermal analysis of CHTR: A parametric study

    SciTech Connect (OSTI)

    Kaushik, Ankur; Basak, Abhishek; Dulera, I. V.; Vijayan, P. K.

    2013-06-12

    Compact High Temperature Reactor (CHTR) is a {sup 233}U-Thorium fuelled, lead-bismuth cooled reactor. The CHTR core mainly consists of graphite and beryllium oxide (BeO). The CHTR core consists of nineteen prismatic beryllium oxide (BeO) moderator blocks. These 19 blocks contain centrally located graphite fuel tubes. The BeO moderator blocks are surrounded by reflector blocks (partially graphite and partially BeO). The nuclear heat from the core is removed passively by natural circulation of the coolant between top and bottom plenums, upward through the fuel tubes and returning through the downcomer tubes at the periphery. The temperature gradient in fuel tubes, downcomer tubes and BeO is very high and therefore, to take care of the differential thermal expansion, gaps are provided in the core between the tubes and other core components. These gaps affect the heat transfer through the core in radial direction. In addition, there is a large variation in thermal properties of graphite which in turn affects the thermal behaviour of the core in various operating conditions. The fuel of CHTR is TRISO coated particle fuel. These particles are packed in with graphite powder as matrix and made into cylindrical compacts these compacts are packed in the bores of fuel tube. In this study, the effect of the thermal conductivity variation of the graphite on the temperature distribution of the core and density variation of the matrix graphite material in fuel compact on the maximum fuel kernel temperature is studied along with the overall role of graphite properties variation in heat transfer.

  3. Statistical Comparison of the Baseline Mechanical Properties of NBG-18 and PCEA Graphite

    SciTech Connect (OSTI)

    Mark C. Carroll; David T. Rohrbaugh

    2013-08-01

    High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR), a graphite-moderated, helium-cooled design that is capable of producing process heat for power generation and for industrial process that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is endeavoring to minimize the conservative estimates of as-manufactured mechanical and physical properties by providing comprehensive data that captures the level of variation in measured values. In addition to providing a comprehensive comparison between these values in different nuclear grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons and variations between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between the two grades of graphite that were initially favored in the two main VHTR designs. NBG-18, a medium-grain pitch coke graphite from SGL formed via vibration molding, was the favored structural material in the pebble-bed configuration, while PCEA, a smaller grain, petroleum coke, extruded graphite from GrafTech was favored for the prismatic configuration. An analysis of the comparison between these two grades will include not only the differences in fundamental and statistically-significant individual strength levels, but also the differences in variability in properties within each of the grades that will ultimately provide the basis for the prediction of in-service performance. The comparative performance of the different types of nuclear grade graphites will continue to evolve as thousands more specimens are fully characterized from the numerous grades of graphite being evaluated.

  4. Nuclear reactor shield including magnesium oxide

    DOE Patents [OSTI]

    Rouse, Carl A. (Del Mar, CA); Simnad, Massoud T. (La Jolla, CA)

    1981-01-01

    An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

  5. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    SciTech Connect (OSTI)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design

  6. Micro Joining of Aluminum Graphite Composites 

    E-Print Network [OSTI]

    Velamati, Manasa

    2012-07-16

    Advanced aluminum graphite composites have unique thermal properties due to opposing coefficients of thermal expansion of aluminum and graphite. The thermal and mechanical properties of such composites are anisotropic due ...

  7. Systems and methods for forming defects on graphitic materials and curing radiation-damaged graphitic materials

    SciTech Connect (OSTI)

    Ryu, Sunmin; Brus, Louis E.; Steigerwald, Michael L.; Liu, Haitao

    2012-09-25

    Systems and methods are disclosed herein for forming defects on graphitic materials. The methods for forming defects include applying a radiation reactive material on a graphitic material, irradiating the applied radiation reactive material to produce a reactive species, and permitting the reactive species to react with the graphitic material to form defects. Additionally, disclosed are methods for removing defects on graphitic materials.

  8. FY-09 Report: Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim

    2009-12-01

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is performing research and development that focuses on key phenomena important during potential scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Gen-IV very high temperature reactor (VHTR). Phenomena Identification and Ranking Studies to date have identified that an air ingress event following on the heels of a VHTR depressurization is a very important incident. Consequently, the development of advanced air ingress-related models and verification and validation data are a very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air will enter the core through the break, leading to oxidation of the in-core graphite structure and fuel. If this accident occurs, the oxidation will accelerate heat-up of the bottom reflector and the reactor core and will eventually cause the release of fission products. The potential collapse of the core bottom structures causing the release of CO and fission products is one of the concerns. Therefore, experimental validation with the analytical model and computational fluid dynamic (CFD) model developed in this study is very important. Estimating the proper safety margin will require experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model. It will also require effective strategies to mitigate the effects of oxidation. The results from this research will provide crucial inputs to the INL NGNP/VHTR Methods Research and Development project. The second year of this three-year project (FY-08 to FY-10) was focused on (a) the analytical, CFD, and experimental study of air ingress caused by density-driven, stratified, countercurrent flow; (b) advanced graphite oxidation experiments and modeling; (c) experimental study of burn-off in the core bottom structures, (d) implementation of advanced graphite oxidation models into the GAMMA code, and (f) air ingress and oxidation mitigation analyses of the whole air-ingress scenario.

  9. Composition and method for brazing graphite to graphite

    DOE Patents [OSTI]

    Taylor, Albert J. (Ten Mile, TN); Dykes, Norman L. (Oak Ridge, TN)

    1984-01-01

    The present invention is directed to a brazing material for joining graphite structures that can be used at temperatures up to about 2800.degree. C. The brazing material formed of a paste-like composition of hafnium carbide and uranium oxide with a thermosetting resin. The uranium oxide is converted to uranium dicarbide during the brazing operation and then the hafnium carbide and uranium dicarbide form a liquid phase at a temperature about 2600.degree. C. with the uranium diffusing and vaporizing from the joint area as the temperature is increased to about 2800.degree. C. so as to provide a brazed joint consisting essentially of hafnium carbide. This brazing temperature for hafnium carbide is considerably less than the eutectic temperature of hafnium carbide of about 3150.degree. C. The brazing composition also incorporates the thermosetting resin so that during the brazing operation the graphite structures may be temporarily bonded together by thermosetting the resin so that machining of the structures to final dimensions may be completed prior to the completion of the brazing operation. The resulting brazed joint is chemically and thermally compatible with the graphite structures joined thereby and also provides a joint of sufficient integrity so as to at least correspond with the strength and other properties of the graphite.

  10. Light Emission from Graphite Surfaces during Beam Bombardment, Observation and Consequences for use of Graphite in Divertors

    E-Print Network [OSTI]

    Light Emission from Graphite Surfaces during Beam Bombardment, Observation and Consequences for use of Graphite in Divertors

  11. Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements

    SciTech Connect (OSTI)

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

    2011-09-01

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

  12. Advanced reactor safety research. Quarterly report, April-June 1982. Volume 22

    SciTech Connect (OSTI)

    None

    1983-10-01

    Overall objective of this work is to provide NRC a comprehensive data base essential to (1) defining key safety issues, (2) understanding risk-significant accident sequences, (3) developing and verifying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the United States without appropriate consideration being given to their effects on health and safety. This report describes progress in a number of activities dealing with current safety issues relevant to both light water and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents, and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  13. Research and Development Assessments for Prometheus eavy Ion and Laser Driven Inertial Fusion Energy Reactor Designs

    E-Print Network [OSTI]

    Abdou, Mohamed

    -ion and laser drivers. INTRODUCTION Two commercial central station electric power plants have been conceptually designed and analyzed in the Prometheus[11study led by McDonnell Douglas Aerospace. These plants use reactors, (2) provide programmatic-decision makers with a list of important R&D tasks that need

  14. LOCA simulation in the national research universal reactor program: postirradiation examination results for the third materials experiment (MT-3)

    SciTech Connect (OSTI)

    Rausch, W.N.

    1984-04-01

    A series of in-reactor experiments were conducted using full-length 32-rod pressurized water reactor (PWR) fuel bundles as part of the Loss-of-Coolant Accident (LOCA) Simulation Program. The third materials experiment (MT-3) was the sixth in the series of thermal-hydraulic and materials deformation/rutpure experiments conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. The main objective of the experiment was to evaluate ballooning and rupture during active two-phase cooling in the temperature range from 1400 to 1500/sup 0/F (1030 to 1090 K). The 12 test rods in the center of the 32-rod bundle were initially pressurized to 550 psi (3.8 MPa) to insure rupture in the correct temperature range. All 12 of the rods ruptured, with an average peak bundle strain of approx. 55%. The UKAEA also funded destructive postirradiation examination (PIE) of several of the ruptured rods from the MT-3 experiment. This report describes the work performed and presents the PIE results. Information obtained during the PIE included cladding thickness measurements metallography, and particle size analysis of the cracked and broken fuel pellets.

  15. Research Update: Atmospheric pressure spatial atomic layer deposition of ZnO thin films: Reactors, doping, and devices

    E-Print Network [OSTI]

    Hoye, Robert L. Z.; Muñoz-Rojas, David; Nelson, Shelby F.; Illiberi, Andrea; Poodt, Paul; Roozeboom, Fred; MacManus-Driscoll, Judith L.

    2015-04-02

    given to describe this system, but in this research update, we standardize its name to the Cambridge University Close Proximity (CUCP) reactor (design details are given in Ref. 24). An illustration of the CUCP gas manifold is given in Fig. 2... times, as indicated. Reprinted with permission from D. Muñoz-Rojas, H. Sun, D. C. Iza, J. Weickert, L. Chen, H. Wang, L. Schmidt-Mende, J. L. MacManus-Driscoll, Prog. Photovoltaics: Res. Appl. 21, 393 (2013). Copyright 2013 Wiley-VCH. adherent...

  16. Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics

    SciTech Connect (OSTI)

    Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1993-03-01

    This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

  17. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    SciTech Connect (OSTI)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N. [Institution Project center ITER, Moscow (Russian Federation)

    2014-08-21

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and ?–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  18. Joint Statement of Intent Concerning the Arak Heavy Water Reactor Research

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties - WAPA Public CommentInverted Attic9: JohnofReactor Modernization

  19. Composition and method for brazing graphite to graphite

    DOE Patents [OSTI]

    Taylor, A.J.; Dykes, N.L.

    1982-08-10

    A brazing material is described for joining graphite structures that can be used up to 2800/sup 0/C. The brazing material is formed of a paste-like composition of hafnium carbide and uranium oxide with a thermosetting resin. The uranium oxide is converted to uranium dicarbide during the brazing operation and then the hafnium carbide and uranium dicarbide form a liquid phase at a temperature about 2600/sup 0/C with the uranium diffusing and vaporizing from the joint area as the temperature is increased to about 2800/sup 0/C so as to provide a brazed joint consisting essentially of hafnium carbide. The resulting brazed joint is chemically and thermally compatible with the graphite structures.

  20. Neutron Reference Benchmark Field Specifications: ACRR Polyethylene-Lead-Graphite (PLG) Bucket Environment (ACRR-PLG-CC-32-CL).

    SciTech Connect (OSTI)

    Vega, Richard Manuel; Parm, Edward J.; Griffin, Patrick J.; Vehar, David W.

    2015-07-01

    This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity with the Polyethylene-Lead-Graphite (PLG) bucket, reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 37 integral dosimetry measurements in the neutron field are reported.

  1. Theoretical analysis of the subcritical experiments performed in the IPEN/MB-01 research reactor facility

    SciTech Connect (OSTI)

    Lee, S. M.; Dos Santos, A. [Inst. de Pesquisas Energeticas e Nucleares, Cidade Universitaria, Av. Lineu Prestes, 2242, 05508-000 Sao Paulo - SP (Brazil)

    2012-07-01

    The theoretical analysis of the subcritical experiments performed at the IPEN/MB-01 reactor employing the coupled NJOY/AMPX-II/TORT systems was successfully accomplished. All the analysis was performed employing ENDF/B-VII.0. The theoretical approach follows all the steps of the subcritical model of Gandini and Salvatores. The theory/experiment comparison reveals that the calculated subcritical reactivity is in a very good agreement to the experimental values. The subcritical index ({xi}) shows some discrepancies although in this particular case some work still have to be made to model in a better way the neutron source present in the experiments. (authors)

  2. Research on direct containment heating and pressurized melt expulsion from the reactor coolant system

    SciTech Connect (OSTI)

    Tarbell, W.W.; Pilch, M.; Brockmann, J.E.; Powers, D.A.

    1988-01-01

    The expulsion of high temperature core debris from the reactor cavity into the containment atmosphere has recently been identified as an important potential contributor to containment failure in the event of a severe accident. Experiments and analyses have shown that failure of the reactor vessel while the primary system is pressurized can result in the rapid discharge of molten core debris into the cavity. Gas from the blowdown of the coolant system may then entrain the debris as fine particulate that may be carried out of the cavity region. Containment loading can result from the combustion of hydrogen produced by the interaction of the debris with steam from the primary system and from thermal and chemical energy transferred from the debris to the atmosphere is directed towards identifying and quantifying the phenomena associated with the pressurized discharge of the core debris and the direct containment heating processes. Experiments are being performed to provide the information needed to develop phenomenological models for use in system level code predictions. Emphasis has been primarily on the use of scaled cavities (ranging from 1:10 to 1:50 linear scale) and the quantification of the extent of the material dispersed. Information has been obtained on the physics of the jet behavior, the entrainment of the debris, debris characteristics (e.g., size and number distributions), debris-gas heat transfer and chemistry, aerosol generation, and the influence of water. Models and codes are reviewed and discussed and representative calculations are presented.

  3. Reactor Safety Research Programs. Quarterly report, July-September 1984. Volume 3. [PWR; BWR

    SciTech Connect (OSTI)

    Edler, S.K.

    1985-02-01

    This document summarizes work performed by Pacific Northwest Laboratory from July 1 through September 30, 1984, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission. Results from an instrumented fuel assembly irradiation program being performed at Halden, Norway, are reported. Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Fuel assemblies and analytical support are being provided for experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory, Idaho Falls, Idaho. High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior. Thermal-hydraulic models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Severe fuel damage accident tests are being conducted in the NRU Reactor, Chalk River, Canada.

  4. Reactor safety research programs. Quarterly report, January-March 1984. Vol. 1. [PWR; BWR

    SciTech Connect (OSTI)

    Edler, S.K.

    1984-06-01

    This document summarizes work performed by Pacific Northwest Laboratory from January 1 through March 31, 1984, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission. Results from an instrumented fuel assembly irradiation program being performed at Halden, Norway, are reported. Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions. Experimental data on analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Fuel assemblies and analytical support are being provided for experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory, Idaho Falls, Idaho. High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior. Thermal-hydraulic models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Severe fuel damage accident tests are being conducted at the NRU reactor, Chalk River, Canada.

  5. Designing a TAC thermometer from a VHTR graphite structure

    SciTech Connect (OSTI)

    Smith, James A. Kotter, Dale; Garrett, Steven L.; Ali, Randall A.

    2015-03-31

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. Very High Temperature Reactors are pushing the in core temperatures even higher. A unique sensing approach will be discussed to address the necessary high temperature measurements. Thermoacoustic thermometry exploits high temperatures and uses materials that are immune to the effects of ionizing radiation to create a temperature sensor that is self-powered and wireless. In addition, the form-factor for the Thermoacoustic Thermometer (TACT) can be designed to be integrated within common in-pile structures. There are no physical moving parts required for TACT and the sensor is self-powered, as it uses the nuclear fuel for its heat source. TACT data will be presented from a laboratory prototype mimicking the design necessary for a VHTR graphite structure.

  6. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01

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

  7. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28

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

  8. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    SciTech Connect (OSTI)

    Not Available

    1982-06-01

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments.

  9. Graphite in Science and Nuclear Technology

    E-Print Network [OSTI]

    Zhmurikov, Evgenij

    2015-01-01

    This review is devoted to the application of the graphite and graphite composites in the science and technology. The structure and an electrical properties, the technological aspects of producing the high-strength artificial graphite and dynamics of its destruction are considered. These type of graphite are traditionally used in the nuclear industry, so author concentrates on the actual problems of the application and testing of the graphite materials in the modern science and technology. Translated from chapters 1 of the the monograph (by Zhmurikov E.I., Bubnenkov I.A., Pokrovsky A.S. et al. Graphite in Science and Nuclear Technique// eprint arXiv:1307.1869, 07/2013 (BC 2013arXiv1307.1869Z).

  10. Graphite in Science and Nuclear Technology

    E-Print Network [OSTI]

    Evgenij Zhmurikov

    2015-08-12

    This review is devoted to the application of the graphite and graphite composites in the science and technology. The structure and an electrical properties, the technological aspects of producing the high-strength artificial graphite and dynamics of its destruction are considered. These type of graphite are traditionally used in the nuclear industry, so author concentrates on the actual problems of the application and testing of the graphite materials in the modern science and technology. Translated from chapters 1 of the the monograph (by Zhmurikov E.I., Bubnenkov I.A., Pokrovsky A.S. et al. Graphite in Science and Nuclear Technique// eprint arXiv:1307.1869, 07/2013 (BC 2013arXiv1307.1869Z).

  11. Decommissioning the Romanian Water-Cooled Water-Moderated Research Reactor: New Environmental Perspective on the Management of Radioactive Waste

    SciTech Connect (OSTI)

    Barariu, G.; Giumanca, R. [Romanian Authority for Nuclear Activity (RAAN), Subsidiary of Technology and Engineering for Nuclear Objectives (SITON), 111 Atomistilor St., Bucuresti-Magurele, Ilfov (Romania)

    2006-07-01

    Pre-feasibility and feasibility studies were performed for decommissioning of the water-cooled water-moderated research reactor (WWER) located in Bucharest - Magurele, Romania. Using these studies as a starting point, the preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as for the rehabilitation of the existing Radioactive Waste Treatment Plant and for the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and ecological reconstruction of the grounds need to be provided for, in accordance with national and international regulations. In accordance with IAEA recommendations at the time, the pre-feasibility study proposed three stages of decommissioning. However, since then new ideas have surfaced with regard to decommissioning. Thus, taking into account the current IAEA ideology, the feasibility study proposes that decommissioning of the WWER be done in one stage to an unrestricted clearance level of the reactor building in an Immediate Dismantling option. Different options and the corresponding derived preferred option for waste management are discussed taking into account safety measures, but also considering technical, logistical and economic factors. For this purpose, possible types of waste created during each decommissioning stage are reviewed. An approximate inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the recommended international basic safety standards identified in the previous phase of the project. The existing Radioactive Waste Treatment Plant (RWTP) from the Horia Hulubei Institute for Nuclear Physics and Engineering (IFIN-HH), which has been in service with no significant upgrade since 1974, will need refurbishing due to deterioration, as well as upgrading in order to ensure the plant complies with current safety standards. This plant will also need to be adapted to treat wastes generated by WWER dismantling. The Baita-Bihor National Radioactive Waste Disposal Facility consists of two galleries in an abandoned uranium mine located in the central-western part of the Bihor Mountains in Transylvania. The galleries lie at a depth of 840 m. The facility requires a considerable overhaul. Several steps recommended for the upgrade of the facility are explored. Environmental concerns have lately become a crucial part of the radioactive waste management strategy. As such, all decisions must be made with great regard for land utilization around nuclear objectives. (authors)

  12. Design of a low enrichment, enhanced fast flux core for the Massachusetts Institute of Technology Research Reactor

    E-Print Network [OSTI]

    Ellis, Tyler Shawn

    2009-01-01

    Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating ...

  13. FINAL REPORT on Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim; Hee C. NO; Nam Z. Cho

    2011-01-01

    The U.S. Department of Energy is performing research and development that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Generation IV very high temperature reactor (VHTR). Phenomena Identification and Ranking studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important. Consequently, the development of advanced air ingress-related models and verification & validation are of very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air ingress will occur through the break, leading to oxidation of the in-core graphite structure and fuel. This study indicates that depending on the location and the size of the pipe break, the air ingress phenomena are different. In an effort to estimate the proper safety margin, experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model are required. It will also require effective strategies to mitigate the effects of oxidation, eventually. This 3-year project (FY 2008–FY 2010) is focused on various issues related to the VHTR air-ingress accident, including (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the core bottom structures, (d) structural tests of the oxidized core bottom structures, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i) verification and validation of the coupled models.

  14. Preliminary Study of Bypass Flow in Prismatic Core of Very High Temperature Reactor Using Small-Scale Model 

    E-Print Network [OSTI]

    Kanjanakijkasem, Worasit 1975-

    2012-11-29

    energy are: gas-cooled fast reactor (GFR); lead-cooled fast reactor (LFR); molten salt reactor (MSR); sodium-cooled fast reactor (SFR); supercritical water-cooled reactor (SCWR); and very high-temperature gas-cooled reactor (VHTR) [3]. These reactor...-temperature gas-cooled reactor (VHTR) is a graphite-moderated, helium-cooled, thermal neutron spectrum reactor with a once-through uranium fuel cycle [4]. The reactor core can be either a prismatic block or a pebble-bed core. The VHTR system is designed to be a...

  15. PIA - 10th International Nuclear Graphite Specialists Meeting...

    Energy Savers [EERE]

    PIA - 10th International Nuclear Graphite Specialists Meeting registration web site PIA - 10th International Nuclear Graphite Specialists Meeting registration web site PIA - 10th...

  16. PIA - 10th International Nuclear Graphite Specialists Meeting...

    Energy Savers [EERE]

    10th International Nuclear Graphite Specialists Meeting registration web site PIA - 10th International Nuclear Graphite Specialists Meeting registration web site PIA - 10th...

  17. PLASMA PHYSICS AND FUSION TECHNOLOGY; GRAPHITE; CREEP; PHYSICAL...

    Office of Scientific and Technical Information (OSTI)

    creep of graphite) Kennedy, C.R. 36 MATERIALS SCIENCE; 70 PLASMA PHYSICS AND FUSION TECHNOLOGY; GRAPHITE; CREEP; PHYSICAL RADIATION EFFECTS; JAPAN; MEETINGS; TRAVEL; ASIA; CARBON;...

  18. Characterization of structural defects in nuclear graphite IG...

    Office of Scientific and Technical Information (OSTI)

    Characterization of structural defects in nuclear graphite IG-110 and NBG-18 Citation Details In-Document Search Title: Characterization of structural defects in nuclear graphite...

  19. Graphite Waste Tank Cleanup and Decontamination under the Marcoule UP1 D and D Program - 13166

    SciTech Connect (OSTI)

    Thomasset, Philippe [AREVA D and D BU, Marcoule (France)] [AREVA D and D BU, Marcoule (France); Chabeuf, Jean-Michel [AREVA D and D BU, La Hague (France)] [AREVA D and D BU, La Hague (France); Thiebaut, Valerie [CEA/DEN/DAPD/CPUP, Marcoule (France)] [CEA/DEN/DAPD/CPUP, Marcoule (France); Chambon, Frederic [AREVA FEDERAL SERVICES, Columbia, MD (United States)] [AREVA FEDERAL SERVICES, Columbia, MD (United States)

    2013-07-01

    The UP1 plant in Marcoule reprocessed nearly 20,000 tons of used natural uranium gas cooled reactor fuel coming from the first generation of civil nuclear reactors in France. During more than 40 years, the decladding operations produced thousands of tons of processed waste, mainly magnesium and graphite fragments. In the absence of a French repository for the graphite waste, the graphite sludge content of the storage pits had to be retrieved and transferred into a newer and safer pit. After an extensive R and D program, the equipment and process necessary for retrieval operations were designed, built and tested. The innovative process is mainly based on the use of two pumps (one to capture and the other one to transfer the sludge) working one after the other and a robotic arm mounted on a telescopic mast. A dedicated process was also set up for the removal of the biggest fragments. The retrieval of the most irradiating fragments was a challenge. Today, the first pit is totally empty and its stainless steel walls have been decontaminated using gels. In the second pit, the sludge retrieval and transfer operations have been almost completed. Most of the non-pumpable graphite fragments has been removed and transferred to a new storage pit. After more than 6 years of operations in sludge retrieval, a lot of experience was acquired from which important 'lessons learned' could be shared. (authors)

  20. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    SciTech Connect (OSTI)

    Monteleone, S.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  1. Design, construction and evaluation of a facility for the simulation of fast reactor blankets

    E-Print Network [OSTI]

    Forbes, Ian Alexander

    1970-01-01

    A facility has been designed and constructed at the MIT Reactor for the experimental investigation of typical LMFBR breeding blankets. A large converter assembly, consisting of a 20-cm-thick layer of graphite followed by ...

  2. A RESEARCH ON SEAMLESS PLATFORM CHANGE OF REACTOR PROTECTION SYSTEM FROM PLC TO FPGA

    E-Print Network [OSTI]

    1. INTRODUCTION A safety grade PLC is an industrial digital computer used to develop safety Hwayang-dong, Gwangjin-gu, Seoul, 143-701, Republic of Korea 2 Korea Atomic Energy Research Institute, Man. The 'FBDtoVerilog' mechanically transforms FBDs into behaviorally-equivalent [38] Verilog programs, and all V

  3. Nuclear Fission Reactor Safety Research in FP7 and future perspectives

    E-Print Network [OSTI]

    Garbil, Roger

    2014-01-01

    The European Union (?U) has defined in the Europe 2020 strategy and 2050 Energy Roadmap its long-term vision for establishing a secure, sustainable and competitive energy system and setting up legally binding targets by 2020 for reducing greenhouse emissions, by increasing energy efficiency and the share of renewable energy sources while including a significant share from nuclear fission. Nuclear energy can enable the further reduction in harmful emissions and can contribute to the EU’s competitive energy system, security of supply and independence from fossil fuels. Nuclear fission is a valuable option for those 14 EU countries that promote its use as part of their national energy mix. The European Group on Ethics in Science and New Technologies (EGE) adopted its Opinion No.27 ‘An ethical framework for assessing research, production and use of energy’ and proposed an integrated ethics approach for the research, production and use of energy in the EU, seeking equilibrium among four criteria – access ...

  4. Superconductivity in graphite intercalation compounds

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Smith, Robert P. [Univ. of Cambridge (United Kingdom); Dean, Mark P. M. [Brookhaven National Lab. (BNL), Upton, NY (United States); Weller, Thomas E. [Univ. College of London (United Kingdom); Howard, Christopher A. [Univ. College of London (United Kingdom); Rahnejat, Kaveh C. [Univ. College of London (United Kingdom); Saxena, Siddharth S. [Univ. of Cambridge (United Kingdom); Ellerby, Mark [Univ. College of London (United Kingdom)

    2015-07-01

    The field of superconductivity in the class of materials known as graphite intercalation compounds has a history dating back to the 1960s This paper recontextualizes the field in light of the discovery of superconductivity in CaC? and YbC? in 2005. In what follows, we outline the crystal structure and electronic structure of these and related compounds. We go on to experiments addressing the superconducting energy gap, lattice dynamics, pressure dependence, and how this relates to theoretical studies. The bulk of the evidence strongly supports a BCS superconducting state. However, important questions remain regarding which electronic states and phonon modes are most important for superconductivity and whether current theoretical techniques can fully describe the dependence of the superconducting transition temperature on pressure and chemical composition.

  5. Superconductivity in graphite intercalation compounds

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Smith, Robert P.; Weller, Thomas E.; Howard, Christopher A.; Dean, Mark P. M.; Rahnejat, Kaveh C.; Saxena, Siddharth S.; Ellerby, Mark

    2015-02-26

    This study examines the field of superconductivity in the class of materials known as graphite intercalation compounds which has a history dating back to the 1960s. This paper recontextualizes the field in light of the discovery of superconductivity in CaC? and YbC? in 2005. In what follows, we outline the crystal structure and electronic structure of these and related compounds. We go on to experiments addressing the superconducting energy gap, lattice dynamics, pressure dependence, and how this relates to theoretical studies. The bulk of the evidence strongly supports a BCS superconducting state. However, important questions remain regarding which electronic statesmore »and phonon modes are most important for superconductivity and whether current theoretical techniques can fully describe the dependence of the superconducting transition temperature on pressure and chemical composition.« less

  6. Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

    SciTech Connect (OSTI)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

    On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

  7. Method for molding threads in graphite panels

    DOE Patents [OSTI]

    Short, W.W.; Spencer, C.

    1994-11-29

    A graphite panel with a hole having a damaged thread is repaired by drilling the hole to remove all of the thread and making a new hole of larger diameter. A bolt with a lubricated thread is placed in the new hole and the hole is packed with graphite cement to fill the hole and the thread on the bolt. The graphite cement is cured, and the bolt is unscrewed therefrom to leave a thread in the cement which is at least as strong as that of the original thread. 8 figures.

  8. Method for molding threads in graphite panels

    DOE Patents [OSTI]

    Short, William W. (Livermore, CA); Spencer, Cecil (Silverton, OR)

    1994-01-01

    A graphite panel (10) with a hole (11) having a damaged thread (12) is repaired by drilling the hole (11) to remove all of the thread and make a new hole (13) of larger diameter. A bolt (14) with a lubricated thread (17) is placed in the new hole (13) and the hole (13) is packed with graphite cement (16) to fill the hole and the thread on the bolt. The graphite cement (16) is cured, and the bolt is unscrewed therefrom to leave a thread (20) in the cement (16) which is at least as strong as that of the original thread (12).

  9. Superhydrophilic graphite surfaces and water-dispersible graphite colloids by electrochemical exfoliation

    SciTech Connect (OSTI)

    Li, Yueh-Feng [Department of Chemical and Materials Engineering, National Central University, Jhongli, 320 Taiwan (China)] [Department of Chemical and Materials Engineering, National Central University, Jhongli, 320 Taiwan (China); Chen, Shih-Ming; Lai, Wei-Hao [Materials and Chemical Research Laboratories, Industrial Technology Research Institute, Chutung, Hsinchu, 31040 Taiwan (China)] [Materials and Chemical Research Laboratories, Industrial Technology Research Institute, Chutung, Hsinchu, 31040 Taiwan (China); Sheng, Yu-Jane [Department of Chemical Engineering, National Taiwan University, Taipei, 106 Taiwan (China)] [Department of Chemical Engineering, National Taiwan University, Taipei, 106 Taiwan (China); Tsao, Heng-Kwong [Department of Chemical and Materials Engineering, Department of Physics, National Central University, Jhongli, 320 Taiwan (China)] [Department of Chemical and Materials Engineering, Department of Physics, National Central University, Jhongli, 320 Taiwan (China)

    2013-08-14

    Superhydrophilic graphite surfaces and water-dispersible graphite colloids are obtained by electrochemical exfoliation with hydrophobic graphite electrodes. Such counterintuitive characteristics are caused by partial oxidation and investigated by examining both graphite electrodes and exfoliated particles after electrolysis. The extent of surface oxidation can be explored through contact angle measurement, scanning electron microscope, electrical sheet resistance, x-ray photoelectron spectroscopy, zeta-potential analyzer, thermogravimetric analysis, UV-visible, and Raman spectroscopy. The degree of wettability of the graphite anode can be altered by the electrolytic current and time. The water contact angle declines generally with increasing the electrolytic current or time. After a sufficient time, the graphite anode becomes superhydrophilic and its hydrophobicity can be recovered by peeling with adhesive tape. This consequence reveals that the anodic graphite is oxidized by oxygen bubbles but the oxidation just occurs at the outer layers of the graphite sheet. Moreover, the characteristics of oxidation revealed by UV peak shift, peak ratio between D and G bands, and negative zeta-potential indicate the presence of graphite oxide on the outer shell of the exfoliated colloids. However, thermogravimetric analysis for the extent of decomposition of oxygen functional groups verifies that the amount of oxygen groups is significantly less than that of graphite oxide prepared via Hummer method. The structure of this partially oxidized graphite may consist of a graphite core covered with an oxidized shell. The properties of the exfoliated colloids are also influenced by pH of the electrolytic solution. As pH is increased, the extent of oxidation descends and the thickness of oxidized shell decreases. Those results reveal that the degree of oxidation of exfoliated nanoparticles can be manipulated simply by controlling pH.

  10. Implementation of MPC and A systems at the VNIIEF research (reactor) site

    SciTech Connect (OSTI)

    Skripka, G.; Koloyartsev, V.; Mikijchuk, N. [Rossijskij Federal`nyj Yadernyj Tsentr, Arzamas (Russian Federation). Nauchno-Issledovatel`skij Inst. Ehksperimental`noj Fiziki] [and others

    1996-09-01

    As part of the US-Russian Lab-to-Lab program for strengthening nuclear material protection, control, and accounting (MPC and A), an integrated MPC and A system is being installed at a large site within the VNIIEF complex to upgrade safeguards on the nuclear material stored and used at this site. In addition to storage facilities, the site houses a number of critical facilities at which nuclear physics research is conducted. The design of the MPC and A system is based on the test bed work presented earlier and provides the functions of nondestructive measurements for plutonium and highly enriched uranium, item control, personnel access control, radiation portal monitoring, and computed on-line accounting. The system controls, monitors, and accounts for nuclear material and people as the material moves through three material balance areas, a measurement room, and a number of control points. It also assists with physical inventory taking. The instrumentation used to implement these functions will include US commercial equipment as well as Russian-designed and -fabricated items.

  11. Immobilization of Rocky Flats Graphite Fines Residue

    SciTech Connect (OSTI)

    Rudisill, T.S.

    1999-04-06

    The development of the immobilization process for graphite fines has proceeded through a series of experimental programs. The experimental procedures and results from each series of experiments are discussed in this report.

  12. Developing fuel management capabilities based on coupled Monte Carlo depletion in support of the MIT Research Reactor (MITR) conversion

    E-Print Network [OSTI]

    Romano, Paul K. (Paul Kollath)

    2009-01-01

    Pursuant to a 1986 NRC ruling, the MIT Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. Prior studies have shown that the MITR will be able to ...

  13. Graphite matrix materials for nuclear waste isolation

    SciTech Connect (OSTI)

    Morgan, W.C.

    1981-06-01

    At low temperatures, graphites are chemically inert to all but the strongest oxidizing agents. The raw materials from which artificial graphites are produced are plentiful and inexpensive. Morover, the physical properties of artificial graphites can be varied over a very wide range by the choice of raw materials and manufacturing processes. Manufacturing processes are reviewed herein, with primary emphasis on those processes which might be used to produce a graphite matrix for the waste forms. The approach, recommended herein, involves the low-temperature compaction of a finely ground powder produced from graphitized petroleum coke. The resultant compacts should have fairly good strength, low permeability to both liquids and gases, and anisotropic physical properties. In particular, the anisotropy of the thermal expansion coefficients and the thermal conductivity should be advantageous for this application. With two possible exceptions, the graphite matrix appears to be superior to the metal alloy matrices which have been recommended in prior studies. The two possible exceptions are the requirements on strength and permeability; both requirements will be strongly influenced by the containment design, including the choice of materials and the waste form, of the multibarrier package. Various methods for increasing the strength, and for decreasing the permeability of the matrix, are reviewed and discussed in the sections in Incorporation of Other Materials and Elimination of Porosity. However, it would be premature to recommend a particular process until the overall multi-barrier design is better defined. It is recommended that increased emphasis be placed on further development of the low-temperature compacted graphite matrix concept.

  14. Method of producing exfoliated graphite, flexible graphite, and nano-scaled graphene platelets

    DOE Patents [OSTI]

    Zhamu, Aruna (Centerville, OH); Shi, Jinjun (Columbus, OH); Guo, Jiusheng (Centerville, OH); Jang, Bor Z. (Centerville, OH)

    2010-11-02

    The present invention provides a method of exfoliating a layered material (e.g., graphite and graphite oxide) to produce nano-scaled platelets having a thickness smaller than 100 nm, typically smaller than 10 nm. The method comprises (a) dispersing particles of graphite, graphite oxide, or a non-graphite laminar compound in a liquid medium containing therein a surfactant or dispersing agent to obtain a stable suspension or slurry; and (b) exposing the suspension or slurry to ultrasonic waves at an energy level for a sufficient length of time to produce separated nano-scaled platelets. The nano-scaled platelets are candidate reinforcement fillers for polymer nanocomposites. Nano-scaled graphene platelets are much lower-cost alternatives to carbon nano-tubes or carbon nano-fibers.

  15. Comparative Analysis of Carbon Plasma in Arc and RF Reactors

    SciTech Connect (OSTI)

    Todorovic-Markovic, B.; Markovic, Z.; Mohai, I.; Szepvolgyi, J.

    2004-12-01

    Results on studies of molecular spectra emitted in the initial stages of fullerene formation during the processing of graphite powder in induction RF reactor and evaporation of graphite electrodes in arc reactor are presented in this paper. It was found that C2 radicals were dominant molecular species in both plasmas. C2 radicals have an important role in the process of fullerene synthesis. The rotational-vibrational temperatures of C2 and CN species were calculated by fitting the experimental spectra to the simulated ones. The results of optical emission study of C2 radicals generated in carbon arc plasma have shown that rotational temperature of C2 species depends on carbon concentration and current intensity significantly. The optical emission study of induction RF plasma and SEM analysis of graphite powder before and after plasma treatment have shown that evaporation of the processed graphite powder depends on feed rate and composition of gas phase significantly. Based on the obtained results, it was concluded that in the plasma region CN radicals could be formed by the reaction of C2 species with atomic nitrogen at smaller loads. At larger feed rate of graphite powder, CN species were produced by surface reaction of the hot carbon particles with nitrogen atoms. The presence of nitrogen in induction RF plasma reduces the fullerene yield significantly. The fullerene yield obtained in two different reactors was: 13% in arc reactor and 4.1% in induction RF reactor. However, the fullerene production rate was higher in induction RF reactor-6.4 g/h versus 1.7 g/h in arc reactor.

  16. Method for producing thin graphite flakes with large aspect ratios

    DOE Patents [OSTI]

    Bunnell, L. Roy (Kennewick, WA)

    1993-01-01

    A method for making graphite flakes of high aspect ratio by the steps of providing a strong concentrated acid and heating the graphite in the presence of the acid for a time and at a temperature effective to intercalate the acid in the graphite; heating the intercalated graphite at a rate and to a temperature effective to exfoliate the graphite in discrete layers; subjecting the graphite layers to ultrasonic energy, mechanical shear forces, or freezing in an amount effective to separate the layes into discrete flakes.

  17. Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina)

    E-Print Network [OSTI]

    Gratta, Giorgio

    Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina) Sandia of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being

  18. Nuclear reactor control

    DOE Patents [OSTI]

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

    1982-01-01

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

  19. Immobilization of Rocky Flats graphite fines residues

    SciTech Connect (OSTI)

    Rudisill, T.S.; Marra, J.C.; Peeler, D.K.

    1999-07-01

    The Savannah River Technology Center (SRTC) is developing an immobilization process for graphite fines residues generated during nuclear materials production activities at the Rocky Flats Environmental Technology Site (Rocky Flats). The continued storage of this material has been identified as an item of concern. The residue was generated during the cleaning of graphite casting molds and potentially contains reactive plutonium metal. The average residue composition is 73 wt% graphite, 15 wt% calcium fluoride (CaF{sub 2}), and 12 wt% plutonium oxide (PuO{sub 2}). Approximately 950 kg of this material are currently stored at Rocky Flats. The strategy of the immobilization process is to microencapsulate the residue by mixing with a sodium borosilicate (NBS) glass frit and heating at nominally 700 C. The resulting waste form would be sent to the Waste Isolation Pilot Plant (WIPP) for disposal. Since the PuO{sub 2} concentration in the residue averages 12 wt%, the immobilization process was required to meet the intent of safeguards termination criteria by limiting plutonium recoverability based on a test developed by Rocky Flats. The test required a plutonium recovery of less than 4 g/kg of waste form when a sample was leached using a nitric acid/CaF{sub 2} dissolution flowsheet. Immobilization experiments were performed using simulated graphite fines with cerium oxide (CeO{sub 2}) as a surrogate for PuO{sub 2} and with actual graphite fines residues. Small-scale surrogate experiments demonstrated that a 4:1 frit to residue ratio was adequate to prevent recovery of greater than 4 g/kg of cerium from simulated waste forms. Additional experiments investigated the impact of varying concentrations of CaF{sub 2} and the temperature/heating time cycle on the cerium recovery. Optimal processing conditions developed during these experiments were subsequently demonstrated at full-scale with surrogate materials and on a smaller scale using actual graphite fines.

  20. Piezoreslstive graphite/polyimide thin films for micromachining applications

    E-Print Network [OSTI]

    Piezoreslstive graphite/polyimide thin films for micromachining applications A. Bruno Frazier) In this work, graphite/polyimide composite thin films are introduced and characterized for micromachining tetracarboxylic dianhydride+xydianiline/metaphenylene diamine polyimide matrix. The resultant material represents

  1. NOVEL SALTS OF GRAPHITE AND A BORON NITRIDE SALT

    E-Print Network [OSTI]

    Bartlett, Neil

    2011-01-01

    ~ i\\f'{y AND DOCUMENTS SECTION NOVEL SALTS OF GRAPHITE ANDA BORON NITRIDE SALT Neil Bartlett, R. N. Biagioni, B. W.privately owned rights. Novel Salts of Graphite and a Boron

  2. Imparting Electrical Conductivity into Asphalt Composites Using Graphite 

    E-Print Network [OSTI]

    Baranikumar, Aishwarya

    2013-07-09

    publications showed the potential of graphite in mitigating the sudden transition. The study presented herein investigates possibility of precisely controlling the electrical conductivity of asphalt concrete only by adding filler size graphite powder. Nine...

  3. First Direct Evidence of Dirac Fermions in Graphite

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

    in Graphite Print The recent surge of interest in the electronic properties of graphene-that is, isolated layers of graphite just one atomic layer thick-has largely been...

  4. Systems and methods for dismantling a nuclear reactor

    SciTech Connect (OSTI)

    Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

    2014-10-28

    Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

  5. Effect of Vinylene Carbonate on Graphite Anode Cycling Efficiency

    E-Print Network [OSTI]

    Ridgway, Paul

    2010-01-01

    graphite formulations in particular, are the current standard for battery anodes in electric vehicle lithium-ion batteries (

  6. Carbon 41 (2003) 11751180 Thermomechanical behavior of a graphite foam

    E-Print Network [OSTI]

    Chung, Deborah D.L.

    2003-01-01

    vehicle radiators, satellite panels, heat exchangers for the low density of 0.5 g/cm , the specific), 0.15 for flexible graphite and 0.22 for PTFE. The loss tangent of the graphite foam decreased strain in the graphite foam upon heating and subsequent cooling, such that the thermal expansion

  7. Method of making segmented pyrolytic graphite sputtering targets

    DOE Patents [OSTI]

    McKernan, M.A.; Alford, C.S.; Makowiecki, D.M.; Chen, C.W.

    1994-02-08

    Anisotropic pyrolytic graphite wafers are oriented and bonded together such that the graphite's high thermal conductivity planes are maximized along the back surface of the segmented pyrolytic graphite target to allow for optimum heat conduction away from the sputter target's sputtering surface and to allow for maximum energy transmission from the target's sputtering surface. 2 figures.

  8. Phasefield Modeling of Graphite Single Particles And Porous Electrodes

    E-Print Network [OSTI]

    Chen, Long-Qing

    Abstract Graphite is the most commonly used anode material in lithiumion batteries because of excellent Phasefield Modeling of Graphite Single Particles And Porous Electrodes Singlelayer regular of the layered graphite material. We use the same model to fit both singleparticle experimental data

  9. Method of making segmented pyrolytic graphite sputtering targets

    DOE Patents [OSTI]

    McKernan, Mark A. (Livermore, CA); Alford, Craig S. (Tracy, CA); Makowiecki, Daniel M. (Livermore, CA); Chen, Chih-Wen (Livermore, CA)

    1994-01-01

    Anisotropic pyrolytic graphite wafers are oriented and bonded together such that the graphite's high thermal conductivity planes are maximized along the back surface of the segmented pyrolytic graphite target to allow for optimum heat conduction away from the sputter target's sputtering surface and to allow for maximum energy transmission from the target's sputtering surface.

  10. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    SciTech Connect (OSTI)

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  11. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  12. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  13. The use of U/sub 3/Si/sub 2/ dispersed in aluminum in plate-type fuel elements for research and test reactors

    SciTech Connect (OSTI)

    Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

    1987-10-01

    A high-density fuel based on U/sub 3/Si/sub 2/ dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U/sub 3/Si/sub 2/ fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U/sub 3/Si/sub 2/ particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U/sub 3/Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U/sub 3/Si/sub 2/-aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m/sup 3/ is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs.

  14. Twenty-First Water Reactor Safety Information Meeting. Volume 3, Primary system integrity; Aging research, products and applications; Structural and seismic engineering; Seismology and geology: Proceedings

    SciTech Connect (OSTI)

    Monteleone, S.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25-27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database.

  15. Research on the HYLIFE liquid-first-wall concept for future laser-fusion reactors. Final report No. 5

    SciTech Connect (OSTI)

    Hoffman, M.A.

    1980-09-01

    It has been proposed to protect the structural walls of a future laser fusion reactor with a curtain or fluid-wall of liquid lithium jets. As part of the investigation of this concept, experiments have been performed on planar sheet water jets issuing vertically downward from slit nozzles. The nozzles were subjected to transverse forced harmonic excitation to simulate the vibrational environment of the laser fusion reactor, and experiments were run at both 1 atm and at lower ambient pressures. Linear temporal stability theory is shown to predict the onset of the unstable regime and the initial spatial growth rates quite well for the cases where the amplitudes of the nozzle vibration are not too large and the waveform is nearly sinusoidal. In addition, both the linear theory and a simplified trajectory theory are shown to predict the initial wave envelope amplitudes very well. For larger amplitude nozzle excitation, the waveform becomes highly nonlinear and non-sinusoidal and can resemble a sawtooth waveform in some cases; these latter experimental results can only be partially explained by existing theories at the present time.

  16. The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases

    SciTech Connect (OSTI)

    Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

    2012-10-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest status and plans are presented.

  17. Atomic resolution images of graphite in air

    SciTech Connect (OSTI)

    Grigg, D.A.; Shedd, G.M.; Griffis, D.; Russell, P.E.

    1988-12-01

    One sample used for proof of operation for atomic resolution in STM is highly oriented pyrolytic graphite (HOPG). This sample has been imaged with many different STM`s obtaining similar results. Atomic resolution images of HOPG have now been obtained using an STM designed and built at the Precision Engineering Center. This paper discusses the theoretical predictions and experimental results obtained in imaging of HOPG.

  18. Electron oxidation of graphite by fluorospecies

    SciTech Connect (OSTI)

    Rosenthal, G.L.

    1984-09-01

    The fluoride-ion affinity (A/sub F/sup -//) of phosphorus pentafluoride was determined to be 100 kcal/mole from the heats of reaction of the Lewis bases SF/sub 4/ and ClO/sub 2/F with PF/sub 5/ near room temperature. The fluoride-ion affinity of boron trifluoride was determined to be 92 kcal/mole from the heat of reaction of ClO/sub 2/F with BF/sub 3/. The crystal structure of ClO/sub 2/BF/sub 4/ was determined and a precise lattice energy was calculated from this structure and used to determined A/sub F/sup -//. Both PF/sub 5/ and BF/sub 3/ were found to react with graphite in the presence of fluorine gas to yield a variety of non-stoichiometric compounds. The fluoride-ion affinity of silicon tetrafluoride is not known, but it does not react with graphite and F/sub 2/ except at high pressures. These and previous results suggested a threshold in oxidizing power of intercalating species below which the oxidative intercalation reaction would not occur. The reduction of C/sub x/PF/sub 6/ by PF/sub 3/ proved that the reaction is thermodynamically controlled to some extent. The displacement of PF/sub 5/ in C/sub x/PF/sub 6/ by BF/sub 3/ (with a smaller A/sub F/sup -//) suggested that two BF/sub 3/ molecules may have a larger fluoride-ion affinity than one PF/sub 5/ and that B/sub 2/F/sub 7//sup -/ may be a stable anion in graphite. Conductivity studies of PF/sub x/ and BF/sub y/ salts showed that a large drop in conductivity when the reaction reaches first stage is due in the most part to direct fluorination of carbon in graphite.

  19. HTGR Dust Safety Issues and Needs for Research and Development

    SciTech Connect (OSTI)

    Paul W. Humrickhouse

    2011-06-01

    This report presents a summary of high temperature gas-cooled reactor dust safety issues. It draws upon a literature review and the proceedings of the Very High Temperature Reactor Dust Assessment Meeting held in Rockville, MD in March 2011 to identify and prioritize the phenomena and issues that characterize the effect of carbonaceous dust on high temperature reactor safety. It reflects the work and input of approximately 40 participants from the U.S. Department of Energy and its National Labs, the U.S. Nuclear Regulatory Commission, industry, academia, and international nuclear research organizations on the topics of dust generation and characterization, transport, fission product interactions, and chemical reactions. The meeting was organized by the Idaho National Laboratory under the auspices of the Next Generation Nuclear Plant Project, with support from the U.S. Nuclear Regulatory Commission. Information gleaned from the report and related meetings will be used to enhance the fuel, graphite, and methods technical program plans that guide research and development under the Next Generation Nuclear Plant Project. Based on meeting discussions and presentations, major research and development needs include: generating adsorption isotherms for fission products that display an affinity for dust, investigating the formation and properties of carbonaceous crust on the inside of high temperature reactor coolant pipes, and confirming the predominant source of dust as abrasion between fuel spheres and the fuel handling system.

  20. Nondestructive Evaluation of Nuclear-Grade Graphite

    SciTech Connect (OSTI)

    Dennis C. Kunerth; Timothy R. McJunkin

    2011-07-01

    Nondestructive Evaluation of Nuclear Grade Graphite Dennis C. Kunerth and Timothy R. McJunkin Idaho National Laboratory Idaho Falls, ID, 83415 This paper discusses the nondestructive evaluation of nuclear grade graphite performed at the Idaho National Laboratory. Graphite is a composite material highly dependent on the base material and manufacturing methods. As a result, material variations are expected within individual billets as well billet to billet and lot to lot. Several methods of evaluating the material have been explored. Particular technologies each provide a subset of information about the material. This paper focuses on techniques that are applicable to in-service inspection of nuclear energy plant components. Eddy current examination of the available surfaces provides information on potential near surface structural defects and although limited, ultrasonics can be utilized in conventional volumetric inspection. Material condition (e.g. micro-cracking and porosity induced by radiation and stress) can be derived from backscatter or acousto-ultrasound (AU) methods. Novel approaches utilizing phased array ultrasonics have been attempted to expand the abilities of AU techniques. By combining variable placement of apertures, angle and depth of focus, the techniques provide the potential to obtain parameters at various depths in the material. Initial results of the study and possible procedures for application of the techniques are discussed.

  1. Immobilization of Rocky Flats Graphite Fines Residues

    SciTech Connect (OSTI)

    Rudisill, T. S.

    1998-11-06

    The Savannah River Technology Center (SRTC) is developing an immobilization process for graphite fines residues generated during nuclear materials production activities at the Rocky Flats Environmental Technology Site (Rocky Flats). The continued storage of this material has been identified as an item of concern. The residue was generated during the cleaning of graphite casting molds and potentially contains reactive plutonium metal. The average residue composition is 73 wt percent graphite, 15 wt percent calcium fluoride (CaF2), and 12 wt percent plutonium oxide (PuO2). Approximately 950 kilograms of this material are currently stored at Rocky Flats. The strategy of the immobilization process is to microencapsulate the residue by mixing with a sodium borosilicate (NBS) glass frit and heating at nominally 700 degrees C. The resulting waste form would be sent to the Waste Isolation Pilot Plant (WIPP) for disposal. Since the PuO2 concentration in the residue averages 12 wt percent, the immobilization process was required to meet the intent of safeguards termination criteria by limiting plutonium recoverability based on a test developed by Rocky Flats. The test required a plutonium recovery of less than 4 g/kg of waste form when a sample was leached using a nitric acid/CaF2 dissolution flowsheet. Immobilization experiments were performed using simulated graphite fines with cerium oxide (CeO2) as a surrogate for PuO2 and with actual graphite fines residues. Small-scale surrogate experiments demonstrated that a 4:1 frit to residue ratio was adequate to prevent recovery of greater than 4 g/kg of cerium from simulated waste forms. Additional experiments investigated the impact of varying concentrations of CaF2 and the temperature/heating time cycle on the cerium recovery. Optimal processing conditions developed during these experiments were subsequently demonstrated at full-scale with surrogate materials and on a smaller scale using actual graphite fines.In general, the recovery of cerium from the full-scale waste forms was higher than for smaller scale experiments. The presence of CaF2 also caused a dramatic increase in cerium recovery not seen in the small-scale experiments. However, the results from experiments with actual graphite fines were encouraging. A 4:1 frit to residue ratio, a temperature of 700 degrees C, and a 2 hr heating time produced waste forms with plutonium recoveries of 4 plus/minus 1 g/kg. With an increase in the frit to residue ratio, waste forms fabricated at this scale should meet the Rocky Flats product specification. The scale-up of the waste form fabrication process to nominally 3 kg is expected to require a 5:1 to 6:1 frit to residue ratio and maintaining the waste form centerline temperature at 700 degrees C for 2 hr.

  2. Forming gas treatment of lithium ion battery anode graphite powders

    DOE Patents [OSTI]

    Contescu, Cristian Ion; Gallego, Nidia C; Howe, Jane Y; Meyer, III, Harry M; Payzant, Edward Andrew; Wood, III, David L; Yoon, Sang Young

    2014-09-16

    The invention provides a method of making a battery anode in which a quantity of graphite powder is provided. The temperature of the graphite powder is raised from a starting temperature to a first temperature between 1000 and 2000.degree. C. during a first heating period. The graphite powder is then cooled to a final temperature during a cool down period. The graphite powder is contacted with a forming gas during at least one of the first heating period and the cool down period. The forming gas includes H.sub.2 and an inert gas.

  3. First Direct Evidence of Dirac Fermions in Graphite

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

    physics, including the novel quantum Hall effect in graphene, the magnetic-field-driven metal-insulator-like transition in graphite, superfluidity in 3He, and the exotic...

  4. Experimental thermal conductivity and contact conductance of graphite composites 

    E-Print Network [OSTI]

    Jackson, Marian Christine

    1998-01-01

    Graphite fiber organic matrix composites were reviewed ics. for potential heat sink applications in the electronics packaging determined the effective transverse and longitudinal thermal industry. This experimental ...

  5. Establish and Expand Commercial Production of Graphite Anode...

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

    Expand Commercial Production of Graphite Anode Materials for High Performance Lithium-ion Batteries 2011 DOE Hydrogen and Fuel Cells Program, and Vehicle Technologies Program...

  6. Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

    SciTech Connect (OSTI)

    Bruce G. Schnitzler

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.

  7. Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

    SciTech Connect (OSTI)

    Trammell, Michael P [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Qualls, A L [ORNL; Harrison, Thomas J [ORNL

    2013-01-01

    ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

  8. Radiation Characterization Summary: ACRR Polyethylene-Lead-Graphite (PLG) Bucket Located in the Central Cavity on the 32-Inch Pedestal at the Core Centerline (ACRR-PLG-CC-32-cl).

    SciTech Connect (OSTI)

    Parma, Edward J.,; Vehar, David W.; Lippert, Lance L.; Griffin, Patrick J.; Naranjo, Gerald E.; Luker, Spencer M.

    2015-06-01

    This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the polyethylene-lead-graphite (PLG) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-PLG-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma-ray fluence profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples. Acknowledgements The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work. Also thanks to David Ames for his assistance in running MCNP on the Sandia parallel machines.

  9. Next Generation Nuclear Plant Research and Development Program Plan

    SciTech Connect (OSTI)

    2005-01-01

    The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: (1) Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission (2) Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: (1) High temperature gas reactor fuels behavior; (2) High temperature materials qualification; (3) Design methods development and validation; (4) Hydrogen production technologies; and (5) Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented in Section 4. The DOE-funded hydrogen production [DOE 2004] and energy conversion technologies programs are described elsewhere.

  10. Next Generation Nuclear Plant Research and Development Program Plan

    SciTech Connect (OSTI)

    P. E. MacDonald

    2005-01-01

    The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission Demonstrate safe and economical nuclearassisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented in Section 4. The DOE-funded hydrogen production [DOE 2004] and energy conversion technologies programs are described elsewhere.

  11. Carbon 40 (2002) 22852289 Flexible graphite as a heating element

    E-Print Network [OSTI]

    Chung, Deborah D.L.

    2002-01-01

    of aircraft [20,21] and the the plane of the sheet [5,6]. Due to the graphite layers heating of floors, pipes, is corrosion-resistant, does not need to be encased flexible graphite is electrically and thermally conductive

  12. Compact Reactor

    SciTech Connect (OSTI)

    Williams, Pharis E. [Williams Research, P.O. Box 554, Los Alamos, NM87544 (United States)

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  13. Reactor Technology | Nuclear Science | ORNL

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

    Reactor Technology Advanced Reactor Concepts Advanced Instrumentation & Controls Light Water Reactor Sustainability Safety and Regulatory Technology Small Modular Reactors Nuclear...

  14. Lithium Ceramic Blankets for Russian Fusion Reactors and Influence of Breeding Operation Mode on Parameters of Reactor Tritium Systems

    SciTech Connect (OSTI)

    Kapyshev, Victor K.; Chernetsov, Mikhail Yu.; Zhevotov, Sergej I.; Kersnovskij, Alexandr Yu.; Kolbasov, Boris N.; Kovalenko, Victor G.; Paltusov, Nikolaj P.; Sernyaev, Georgeij A.; Sterebkov, Juri S.; Zyryanov, Alexej P.

    2005-07-15

    Russian controlled fusion program supposes development of a DEMO reactor design and participation in ITER Project. A solid breeder blanket of DEMO contains a ceramic lithium orthosilicate breeder and a beryllium multiplier. Test modules of the blanket are developed within the scope of ITER activities. Experimental models of module tritium breeding zones (TBZ), materials and fabrication technology of the TBZ, tritium reactor systems to analyse and process gas released from lithium ceramics are being developed. Two models of tritium breeding and neutron multiplying elements of the TBZ have been designed, manufactured and tested in IVV-2M nuclear reactor. Initial results of the in-pile experiments and outcome of lithium ceramics irradiation in a water-graphite nuclear reactor are considered to be a data base for development of the test modules and initial requirements for DEMO tritium system design. Influence of the tritium release parameters and hydrogen concentration in a purge gas on parameters of reactor system are discussed.

  15. Phoenix: A Reactor Burnup Code With Uncertainty Quantification 

    E-Print Network [OSTI]

    Spence, Grant R

    2014-12-15

    , and will continue to be, an area of great interest in nuclear research. Several methods have been developed to simulate reactor burnup; however, quantifying the uncertainty in reactor burnup simulations is in its relative infancy. This research developed a...

  16. Reactor and Nuclear Systems Division (RNSD)

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

    RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation...

  17. O2 Reduction on Graphite and Nitrogen-Doped Graphite: Experiment and Theory Reyimjan A. Sidik and Alfred B. Anderson*

    E-Print Network [OSTI]

    Popov, Branko N.

    toward NOx reduction.7,8 Others are oxidation catalysts.9 In the case of NOx oxidation and O2 reductionO2 Reduction on Graphite and Nitrogen-Doped Graphite: Experiment and Theory Reyimjan A. Sidik for reduction of approximately 0.5 V (SHE) compared to the onset potential of 0.2 V observed for untreated

  18. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL...

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

    Problem Integration Virtual Reactor Integration (VRI) Bridging the gap between research and engineering. Chemistry Mesh Motion Quality Improvement Multi- resolution...

  19. Summer 2014 Undergraduate Research Experience in Physics: Materials and Modeling

    E-Print Network [OSTI]

    ) nuclear reactor found on a college campus. The MU Research Reactor (MURR) provides advanced research experiments, theory, and modeling and span a variety of research projects (see http

  20. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul (Pittsburgh, PA)

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  1. Thermonuclear Reflect AB-Reactor

    E-Print Network [OSTI]

    Alexander Bolonkin

    2008-03-26

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

  2. Pulsed Repetition Rate Nanosecond Laser Heating and Ablation of the Tokamak Graphite Tile Deposited Layers

    E-Print Network [OSTI]

    Pulsed Repetition Rate Nanosecond Laser Heating and Ablation of the Tokamak Graphite Tile Deposited Layers

  3. Challenges in the Development of Advanced Reactors

    SciTech Connect (OSTI)

    P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

    2012-08-01

    Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

  4. Stabilized Spheromak Fusion Reactors

    SciTech Connect (OSTI)

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  5. EVALUATION OF ZERO-POWER, ELEVATED-TEMPERATURE MEASUREMENTS AT JAPAN’S HIGH TEMPERATURE ENGINEERING TEST REACTOR

    SciTech Connect (OSTI)

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2011-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The experimental benchmark performed and currently evaluated in this report pertains to the data available for two zero-power, warm-critical measurements with the fully-loaded HTTR core. Six isothermal temperature coefficients for the fully-loaded core from approximately 340 to 740 K have also been evaluated. These experiments were performed as part of the power-up tests (References 1 and 2). Evaluation of the start-up core physics tests specific to the fully-loaded core (HTTR-GCR-RESR-001) and annular start-up core loadings (HTTR-GCR-RESR-002) have been previously evaluated.

  6. Catalytic reactor

    DOE Patents [OSTI]

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

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  7. Sodium Reactor Experiment decommissioning. Final report

    SciTech Connect (OSTI)

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

    1983-08-15

    The Sodium Reactor Experiment (SRE) located at the Rockwell International Field Laboratories northwest of Los Angeles was developed to demonstrate a sodium-cooled, graphite-moderated reactor for civilian use. The reactor reached full power in May 1958 and provided 37 GWh to the Southern California Edison Company grid before it was shut down in 1967. Decommissioning of the SRE began in 1974 with the objective of removing all significant radioactivity from the site and releasing the facility for unrestricted use. Planning documentation was prepared to describe in detail the equipment and techniques development and the decommissioning work scope. A plasma-arc manipulator was developed for remotely dissecting the highly radioactive reactor vessels. Other important developments included techniques for using explosives to cut reactor vessel internal piping, clamps, and brackets; decontaminating porous concrete surfaces; and disposing of massive equipment and structures. The documentation defined the decommissioning in an SRE dismantling plan, in activity requirements for elements of the decommissioning work scope, and in detailed procedures for each major task.

  8. Mechanics of fatigue damage in titanium-graphite hybrid laminates

    E-Print Network [OSTI]

    Burianek, Dennis Arthur

    2001-01-01

    Titanium-graphite hybrid laminates are being developed for high-temperature aerospace applications. Experimental observations have indicated that cracks in the titanium facesheets initiate at free edges as well as in areas ...

  9. Dry synthesis of lithium intercalated graphite powders and carbon fibers

    SciTech Connect (OSTI)

    Sacci, Robert L [ORNL; Adamczyk, Leslie A [ORNL; Veith, Gabriel M [ORNL; Dudney, Nancy J [ORNL

    2014-01-01

    Herein we describe the direct synthesis of lithium intercalated graphite by heating under vacuum or ball milling under pressurized Ar(g). Both methods allow for stoichometric control of Li-C ratio in batter-grade graphites and carbon fibers prior formation of a solid electrolyte interphase. The products' surface chemistries, as probed by XPS, suggest that LiC6 are extremely reactive with trace amounts of moisture or oxygen. The open circuit potential and SEM data show that the reactivity of the lithiated battery-grade graphite and the carbon fiber can be related to the density of edge/defect sites on the surfaces. Preliminary results of spontaneous SEI formation on Li-graphite in electrolyte are also given.

  10. Disposal options for burner ash from spent graphite fuel. Final study report November 1993

    SciTech Connect (OSTI)

    Pinto, A.P.

    1994-08-01

    Three major disposal alternatives are being considered for Fort St. Vrain Reactor (FSVR) and Peach Bottom Reactor (PBR) spent fuels: direct disposal of packaged, intact spent fuel elements; (2) removal of compacts to separate fuel into high-level waste (HLW) and low-level waste (LLW); and (3) physical/chemical processing to reduce waste volumes and produce stable waste forms. For the third alternative, combustion of fuel matrix graphite and fuel particle carbon coatings is a preferred technique for head-end processing as well as for volume reduction and chemical pretreatment prior to final fixation, packaging, and disposal of radioactive residuals (fissile and fertile materials together with fission and activation products) in a final repository. This report presents the results of a scoping study of alternate means for processing and/or disposal of fissile-bearing particles and ash remaining after combustion of FSVR and PBR spent graphite fuels. Candidate spent fuel ash (SFA) waste forms in decreasing order of estimated technical feasibility include glass-ceramics (GCs), polycrystalline ceramic assemblages (PCAs), and homogeneous amorphous glass. Candidate SFA waste form production processes in increasing order of estimated effort and cost for implementation are: low-density GCs via fuel grinding and simultaneous combustion and waste form production in a slagging cyclone combustor (SCC); glass or low-density GCs via fluidized bed SFA production followed by conventional melting of SFA and frit; PCAs via fluidized bed SFA production followed by hot isostatic pressing (HIPing) of SFA/frit mixtures; and high-density GCs via fluidized bed SFA production followed by HIPing of Calcine/Frit/SFA mixtures.

  11. Gaseous fission product management for molten salt reactors and vented fuel systems

    SciTech Connect (OSTI)

    Messenger, S. J.; Forsberg, C.; Massie, M.

    2012-07-01

    Fission gas disposal is one of the unresolved difficulties for Molten Salt Reactors (MSRs) and advanced reactors with vented fuel systems. As these systems operate, they produce many radioactive isotopes of xenon and krypton (e.g. {sup 135}Xe t{sub 1/2} = 9.14 hours and {sup 85}Kr t{sub 1/2}= 10.73 years). Removing these gases proves vital to the success of such reactor designs for two reasons. First, the gases act as large neutron sinks which decrease reactivity and must be counterbalanced by increasing fuel loading. Second, for MSRs, inert fission product gases naturally separate quickly from high temperature salts, thus creating high vapor pressure which poses safety concerns. For advanced reactors with solid vented fuel, the gases are allowed to escape into an off-gas system and thus must be managed. Because of time delays in transport of fission product gases in vented fuel systems, some of the shorter-lived radionuclides will decay away thereby reducing the fission gas source term relative to an MSR. To calculate the fission gas source term of a typical molten salt reactor, we modeled a 1000 MWe graphite moderated thorium MSR similar to that detailed in Mathieu et al. [1]. The fuel salt used in these calculations was LiF (78 mole percent) - (HN)F 4 (22 mole percent) with a heavy nuclide composition of 3.86% {sup 233}U and 96.14% {sup 232}Th by mass. Before we can remove the fission product gases produced by this reactor configuration, we must first develop an appropriate storage mechanism. The gases could be stored in pressurized containers but then one must be concerned about bottle failure. Methods to trap noble gases in matrices are expensive and complex. Alternatively, there are direct storage/disposal options: direct injection into the Earth or injecting a grout-based product into the Earth. Advances in drilling technologies, hydro fracture technologies, and methods for the sequestration of carbon dioxide from fossil fuel plants are creating new options for disposal of fission gas wastes. In each option, lithostatic pressure, a kilometer or more underground, eliminates the pressure driving force for noble gas release and dissolves any untrapped gas in deep groundwater or into incorporated solid waste forms. The options, challenges, and potential for these methods to dispose of gaseous fission products are described. With this research, we hope to help both MSRs and other advanced reactors come one step closer to commercialization. (authors)

  12. Compression induced delamination in a unidirectional graphite/epoxy composite 

    E-Print Network [OSTI]

    Earley, John W.

    1981-01-01

    December 1981 Major Subject: Civil Engineering COMPRESSION INDUCED DELAMINATION IN A UNIDIRECTIONAL GRAPHITE/EPOXY COMPOSITE A Thesis by JOHN W. EARLEY Approved as to style and content by: (K. L. Jerina, Chairman) (R, A. Schape y', MemP ) W. L.... Bradley, Me er Il. 8 0 1d, O. pa t~tH d December 1981 ABSTRACT Compression Induced Delamination in a Unidirectional Graphite/Epoxy Composite (December 1981) John William Earley, B. S. Aeronautical Engineering California Polytechnic State University...

  13. Slow, stable delamination in graphite/epoxy composites 

    E-Print Network [OSTI]

    Razi, Hamid

    1982-01-01

    SLOB, STABLE DELAFIINATION IN GRAPHITE/EPOXY COMPOSITES A Thesis by HAMID RA2I Submitted to the Graduate College of Texas ASM University in partial fulfillment of the reouirement for the degree of MASTER OF SCIENCE December 1982 Major... Subject: Mechanical Engineering SLOW, STABLE DELAMINATION IN GRAPHITE/EPOXY COMPOSITES A Thesis by HAMID RAZI Approved as to style and content by: (R. A. Schapery, hair (J. R. Wa ton, Member) (W. L. Bradley, Membe . R. Hopkins, ead of Department...

  14. An investigation of damage accumulation in graphite/epoxy laminates 

    E-Print Network [OSTI]

    Norvell, Robert Gerald

    1985-01-01

    AN INVESTIGATION OF DAMAGE ACCUMULATION IN GRAPHITE/EPOXY LAMINATES A Thesis by ROBERT GERALD NORVELL Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE... August 1985 Major Subject: Aerospace Engineering AN INVESTIGATION OF DAMAGE ACCUMULATION IN GRAPHITE/EPOXY LAMINATES A Thesis by ROBERT GERALD NORVELL Approved as to style and content by: David H. Allen (Co-Chair of C mmitt. ) Richard A. Schap...

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would likeUniverse (Journal Article) |FinalIndustrialFollowingFor yourMatteo ManeraSavannah

  16. Graphite electrode arc melter demonstration Phase 2 test results

    SciTech Connect (OSTI)

    Soelberg, N.R.; Chambers, A.G.; Anderson, G.L.; O`Connor, W.K.; Oden, L.L.; Turner, P.C.

    1996-06-01

    Several U.S. Department of Energy organizations and the U.S. Bureau of Mines have been collaboratively conducting mixed waste treatment process demonstration testing on the near full-scale graphite electrode submerged arc melter system at the Bureau`s Albany (Oregon) Research Center. An initial test series successfully demonstrated arc melter capability for treating surrogate incinerator ash of buried mixed wastes with soil. The conceptual treatment process for that test series assumed that buried waste would be retrieved and incinerated, and that the incinerator ash would be vitrified in an arc melter. This report presents results from a recently completed second series of tests, undertaken to determine the ability of the arc melter system to stably process a wide range of {open_quotes}as-received{close_quotes} heterogeneous solid mixed wastes containing high levels of organics, representative of the wastes buried and stored at the Idaho National Engineering Laboratory (INEL). The Phase 2 demonstration test results indicate that an arc melter system is capable of directly processing these wastes and could enable elimination of an up-front incineration step in the conceptual treatment process.

  17. Neutronic reactor

    DOE Patents [OSTI]

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

    1983-01-01

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

  18. Electrical and thermal properties of graphite/polyaniline composites

    SciTech Connect (OSTI)

    Bourdo, Shawn E., E-mail: sxbourdo@ualr.edu [Center for Integrative Nanotechnology Sciences, University of Arkansas at Little Rock, 2801 South University Avenue, Little Rock, AR 72204 (United States); Warford, Brock A.; Viswanathan, Tito [Department of Chemistry, University of Arkansas at Little Rock, 2801 South University Avenue, Little Rock, AR 72204 (United States)] [Department of Chemistry, University of Arkansas at Little Rock, 2801 South University Avenue, Little Rock, AR 72204 (United States)

    2012-12-15

    A composite of a carbon allotrope (graphite) and an inherently conducting polymer, polyaniline (PANI), has been prepared that exhibits an electrical conductivity greater than either of the two components. An almost 2-fold increase in the bulk conductivity occurs when only a small mass fraction of polyaniline exists in the composite (91% graphite/ 9% polyaniline, by mass). This increase in dc electrical conductivity is curious since in most cases a composite material will exhibit a conductivity somewhere between the two individual components, unless a modification to the electronic nature of the material occurs. In order to elucidate the fundamental electrical properties of the composite we have performed variable temperature conductivity measurements to better understand the nature of conduction in these materials. The results from these studies suggest a change in the mechanism of conduction as the amount of polyaniline is increased in the composite. Along with superior electrical properties, the composites exhibit an increase in thermal stability as compared to the graphite. - Graphical abstract: (Left) Room temperature electrical conductivity of G-PANI composites at different mass ratios. (Right) Electrical conductivity of G-PANI composites at temperatures from 5 K to 300 K. Highlights: Black-Right-Pointing-Pointer Composites of graphite and polyaniline have been synthesized with unique electrical and thermal properties. Black-Right-Pointing-Pointer Certain G-PANI composites are more conductive and more thermally stable than graphite alone. Black-Right-Pointing-Pointer G-PANI composites exhibit a larger conductivity ratio with respect to temperature than graphite alone.

  19. Characterisation of graphite using boron as a marker element

    SciTech Connect (OSTI)

    Kamble, Granthali S.; Pandey, Shailaja; Thakur, Neha; Kumar, Sanjukta A.; Venkatesh, K.; Kumar, Sangita D.; Kameswaran, R.; Reddy, A. V. R.

    2013-06-12

    Graphite has many industrial applications. Two of the most important applications are as electrodes in industries and as moderator in nuclear industry. Determination of the Boron Equivalent of the impurity elements in graphite is the most important parameter for certifying the grade of graphite electrode [1]. The use of a suitable method with low limits of determination of boron is therefore necessary. A method has been standardised in Analytical Chemistry Division, BARC for determining trace amounts of boron in graphite electrodes. It involves controlled dissolution of graphite sample powder and measurement of boron by Inductively Coupled Plasma Mass Spectrometer (ICP-MS) using matrix matched standards. The method detection limit is 1 {mu}g g{sup -1}. The method Relative Standard Deviation was 5%. The method was verified by spike recovery experiments. Recoveries were found to be within 100{+-}2% in the concentration range of 1 to 100 {mu}g g{sup -1}. The developed method has been adopted for the compositional characterization of several graphite electrode samples.

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    increase for a typical sodium fast reactor fuel rod geometryof the new Russian sodium fast reactor BN-800 [111]. Thethe strong focus of sodium fast reactor research to avoid

  1. Fuel and core testing plan for a target fueled isotope production reactor.

    SciTech Connect (OSTI)

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

    2010-12-01

    In recent years there has been an unstable supply of the critical diagnostic medical isotope 99Tc. Several concepts and designs have been proposed to produce 99Mo the parent nuclide of 99Tc, at a commercial scale sufficient to stabilize the world supply. This work lays out a testing and experiment plan for a proposed 2 MW open pool reactor fueled by Low Enriched Uranium (LEU) 99Mo targets. The experiments and tests necessary to support licensing of the reactor design are described and how these experiments and tests will help establish the safe operating envelop for a medical isotope production reactor is discussed. The experiments and tests will facilitate a focused and efficient licensing process in order to bring on line a needed production reactor dedicated to supplying medical isotopes. The Target Fuel Isotope Reactor (TFIR) design calls for an active core region that is approximately 40 cm in diameter and 40 cm in fuel height. It contains up to 150 cylindrical, 1-cm diameter, LEU oxide fuel pins clad with Zircaloy (zirconium alloy), in an annular hexagonal array on a {approx}2.0 cm pitch surrounded, radially, by a graphite or a Be reflector. The reactor is similar to U.S. university reactors in power, hardware, and safety/control systems. Fuel/target pin fabrication is based on existing light water reactor fuel fabrication processes. However, as part of licensing process, experiments must be conducted to confirm analytical predictions of steady-state power and accident conditions. The experiment and test plan will be conducted in phases and will utilize existing facilities at the U.S. Department of Energy's Sandia National Laboratories. The first phase is to validate the predicted reactor core neutronics at delayed critical, zero power and very low power. This will be accomplished by using the Sandia Critical Experiment (CX) platform. A full scale TFIR core will be built in the CX and delayed critical measurements will be taken. For low power experiments, fuel pins can be removed after the experiment and using Sandia's metrology lab, relative power profiles (radially and axially) can be determined. In addition to validating neutronic analyses, confirming heat transfer properties of the target/fuel pins and core will be conducted. Fuel/target pin power limits can be verified with out-of-pile (electrical heating) thermal-hydraulic experiments. This will yield data on the heat flux across the Zircaloy clad and establish safety margin and operating limits. Using Sandia's Annular Core Research Reactor (ACRR) a 4 MW TRIGA type research reactor, target/fuel pins can be driven to desired fission power levels for long durations. Post experiment inspection of the pins can be conducted in the Auxiliary Hot Cell Facility to observe changes in the mechanical properties of the LEU matrix and burn-up effects. Transient tests can also be conducted at the ACRR to observe target/fuel pin performance during accident conditions. Target/fuel pins will be placed in double experiment containment and driven by pulsing the ACRR until target/fuel failure is observed. This will allow for extrapolation of analytical work to confirm safety margins.

  2. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

    SciTech Connect (OSTI)

    Philip E. MacDonald

    2005-01-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks: Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking Task 3. Plant Engineering and Reactor Safety Analysis. moderator rods. materials.

  3. Research

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

    Research Research Isotopes produced at Los Alamos National Laboratory are saving lives, advancing cutting-edge research and keeping the U.S. safe. Research thorium test foil A...

  4. Investigation on the Core Bypass Flow in a Very High Temperature Reactor

    SciTech Connect (OSTI)

    Hassan, Yassin

    2013-10-22

    Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: • Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. • Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. • Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. • Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the high side, the quantity of cooling flow through the core may be considerably less than the nominal design value, causing some regions of the core to operate at temperatures in excess of the design values. These effects are postulated to lead to localized hot regions in the core that must be considered when evaluating the VHTR operational and accident scenarios.

  5. Method for wetting a boron alloy to graphite

    DOE Patents [OSTI]

    Storms, E.K.

    1987-08-21

    A method is provided for wetting a graphite substrate and spreading a a boron alloy over the substrate. The wetted substrate may be in the form of a needle for an effective ion emission source. The method may also be used to wet a graphite substrate for subsequent joining with another graphite substrate or other metal, or to form a protective coating over a graphite substrate. A noneutectic alloy of boron is formed with a metal selected from the group consisting of nickel (Ni), palladium (Pd), and platinum (Pt) with excess boron, i.e., and atomic percentage of boron effective to precipitate boron at a wetting temperature of less than the liquid-phase boundary temperature of the alloy. The alloy is applied to the substrate and the graphite substrate is then heated to the wetting temperature and maintained at the wetting temperature for a time effective for the alloy to wet and spread over the substrate. The excess boron is evenly dispersed in the alloy and is readily available to promote the wetting and spreading action of the alloy. 1 fig.

  6. Structure of processes in flow reactor and closed reactor: Flow reactor

    E-Print Network [OSTI]

    Greifswald, Ernst-Moritz-Arndt-Universität

    Structure of processes in flow reactor and closed reactor: Flow reactor Closed reactor Active Zone -- chemical quasi- equilibria, similarity principles and macroscopic kinetics", in: Lectures on Plasma Physics

  7. Effect of Vinylene Carbonate on Graphite Anode Cycling Efficiency

    SciTech Connect (OSTI)

    Ridgway, Paul; Zheng, Honghe; Liu, Gao; Song, Xiangun; Ross, Philip; Battaglia, Vincent

    2009-05-05

    Vinylene Carbonate (VC) was added to the electrolyte in graphite-lithium half-cells. We report its effect on the coulombic efficiency (as capacity shift) of graphite electrodes under various formation cycling conditions. Cyclic voltammetry on glassy carbon showed that VC passivates the electrode against electrolyte reduction. The dQ/dV plots of the first lithiation of the graphite suggest that VC alters the SEI layer, and that by varying the cell formation rate, the initial ratio of ethylene carbonate to VC in the SEI layer can be controlled. VC was found to decrease first cycle efficiency and reversible capacity (in ongoing cycling) when used to excess. However, experiments with VC additive used with various formation rates did not show any decrease in capacity shift.

  8. Graphite Foams for Lithium-Ion Battery Current Collectors

    SciTech Connect (OSTI)

    Dudney, Nancy J [ORNL; Tiegs, Terry N [ORNL; Kiggans, Jim [ORNL; Jang, Young-Il [ORNL; Klett, James William [ORNL

    2007-01-01

    Graphite open-cell foams, with their very high electronic and thermal conductivities, may serve as high surface area and corrosion resistant current collectors for lithium-ion batteries. As a proof of principle, cathodes were prepared by sintering carbon-coated LiFePO4 particles into the porous graphite foams. Cycling these cathodes in a liquid electrolyte cell showed promising performance even for materials and coatings that have not been optimized. The specific capacity is not limited by the foam structure, but by the cycling performance of the coated LiFePO4 particles. Upon extended cycling for more than 100 deep cycles, no loss of capacity is observed for rates of C/2 or less. The uncoated graphite foams will slowly intercalate lithium reversibly at potentials less than 0.2 volts versus lithium.

  9. Graphitization of polymer surfaces by scanning ion irradiation

    SciTech Connect (OSTI)

    Koval, Yuri [Department of Physics, Universität Erlangen-Nürnberg, Erwin-Rommel-Str. 1, 91058 Erlangen (Germany)

    2014-10-20

    Graphitization of polymer surfaces was performed by low-energy Ar{sup +} and He{sup +} ion irradiation. A method of scanning irradiation was implemented. It was found that by scanning ion irradiation, a significantly higher electrical conductivity in the graphitized layers can be achieved in comparison with a conventional broad-beam irradiation. The enhancement of the conductance becomes more pronounced for narrower and better collimated ion beams. In order to analyze these results in more detail, the temperature dependence of conductance of the irradiated samples was investigated. The results of measurements are discussed in terms of weak localization corrections to conductance in disordered metals. The observed effects can be explained by enlargement of graphitic patches, which was achieved with the scanning ion irradiation method.

  10. Chemical and Physical Modification of Graphitic Materials by Oxidative Processes and Solvent Intercalation 

    E-Print Network [OSTI]

    Werke, Carrie Beth

    2014-08-27

    is by chemical functionalization, such as oxidation. This work looks at two different oxidation techniques for graphite; UV/O3 exposure and biased AFM lithography for broad and local oxidation, respectively. For the supported graphitic samples including graphene...

  11. Graphit-ceramic RF Faraday-thermal shield and plasma limiter

    DOE Patents [OSTI]

    Hwang, David L. (Princeton Junction, NJ); Hosea, Joel C. (Princeton, NJ)

    1989-01-01

    The present invention is directed to a process of brazing a ceramic mater to graphite. In particular, the brazing procedure is directed to the production of a novel brazed ceramic graphite product useful as a Faraday shield.

  12. Low-Cost Graphite and Olivine-Based Materials for Li-Ion Batteries...

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

    Graphite and Olivine-Based Materials for Li-Ion Batteries Low-Cost Graphite and Olivine-Based Materials for Li-Ion Batteries Presentation from the U.S. DOE Office of Vehicle...

  13. Analysis of a graphite foam-NaCl latent heat storage system for...

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

    Analysis of a graphite foam-NaCl latent heat storage system for supercritical CO2 power cycles for concentrated solar power Title Analysis of a graphite foam-NaCl latent heat...

  14. Reactor accelerator coupling experiments: a feasability study 

    E-Print Network [OSTI]

    Woddi Venkat Krishna, Taraknath

    2006-08-16

    ) ______________________________ William E. Burchill (Head of Department) ______________________________ Marvin L. Adams (Member) ______________________________ Ronald J. Ellis (Member) May 2005 Major Subject: Nuclear Engineering iii ABSTRACT Reactor... (Commissariat ? l??nergie Atomique), EDF (Electricit? de France), and FRAMATOME. This experimental program takes place at the research reactor MASURCA (MAquette de SURg?n?rateur ? CAdarache) of the CEA Nuclear Center of Cadarache. It is devoted to the study...

  15. Gelcasting polymeric precursors for producing net-shaped graphites

    DOE Patents [OSTI]

    Klett, James W.; Janney, Mark A.

    2005-02-15

    The present invention discloses a method for molding complex and intricately shaped high density monolithic carbon, carbon-carbon, graphite, and thermoplastic composites using gelcasting technology. The method comprising a polymeric carbon precursor, a solvent, a dispersant, an anti-foaming agent, a monomer system, and an initiator system. The components are combined to form a suspension which is poured into a mold and heat-treated to form a thermoplastic part. The thermoplastic part can then be further densified and heat-treated to produce a high density carbon or graphite composite. The present invention also discloses the products derived from this method.

  16. Microfabrication of freestanding metal structures released from graphite substrates.

    SciTech Connect (OSTI)

    Makarova, O. V.; Tang, C.-M.; Mancini, D. C.; Moldovan, N.; Divan, R.; Ryding, D. G.; Lee, R. H.

    2002-02-22

    A sacrificial layer is usually used to release electroformed microstructures. Because of the chemistry applied to the sacrificial layer, only a limited number of metals can be used for electroforming. A novel method to fabricate freestanding electroformed copper structures is presented. A graphite substrate allows the release of the metal part, by abrasive removal of the graphite after electroforming. Results on fabrication of high-aspect-ratio freestanding copper grids are presented; these can be used as x-ray collimator in medical imaging to reduce scattered radiation. This process has potential application to the fabrication of injection molds and microparts on pick-and-place carriers for microelectromechanical systems (MEMS).

  17. Delamination fracture toughness of a unidirectional graphite/epoxy composite 

    E-Print Network [OSTI]

    Hulsey, Roy Charles

    1980-01-01

    DELAMINATION FRACTURE TOUGHNESS OF A UNIDIRECTIONAL GRAPHITE/EPOXY COMPOS ITE A Thesis by ROY CHARLES HULSEY Submitted to the Graduate College of Texas A8M University in partial fulfillment of the requirement for the degree of MASTER... OF SCIENCE December 1980 Major Subject: Mechanical Engineering DELAMINA. ION FRACTURE TOUGHNESS OF A UNIDIRECTIONAL GRAPHITE/EPOXY COMPOSITE A Thesis by ROY CHARLES HULSEY Approved as to sty1e and content by: +alter L. Brad1ey, C airman TR. A. S p...

  18. Early Damage Mechanisms in Nuclear Grade Graphite under Irradiation

    SciTech Connect (OSTI)

    Eapen, Dr. Jacob [North Carolina State University] [North Carolina State University; Krishna, Dr Ram [North Carolina State University] [North Carolina State University; Burchell, Timothy D [ORNL] [ORNL; Murty, Prof K.L. [North Carolina State University] [North Carolina State University

    2014-01-01

    Using Raman and X-ray photoelectron spectroscopy,we delineate the bond and defect structures in nuclear block graphite (NBG-18) under neutron and ion irradiation. The strengthening of the defect (D) peak in the Raman spectra under irradiation is attributed to an increase in the topological, sp2-hybridized defects. Using transmission electron microscopy, we provide evidence for prismatic dislocations as well as a number of basal dislocations dissociating into Shockley partials. The non-vanishing D peak in the Raman spectra, together with a generous number of dislocations, even at low irradiation doses, indicates a dislocation-mediated amorphization process in graphite.

  19. Gelcasting polymeric precursors for producing net-shaped graphites

    DOE Patents [OSTI]

    Klett, James W. (Knoxville, TN); Janney, Mark A. (Knoxville, TN)

    2002-01-01

    The present invention discloses a method for molding complex and intricately shaped high density monolithic carbon, carbon-carbon, graphite, and thermoplastic composites using gelcasting technology. The method comprising a polymeric carbon precursor, a solvent, a dispersant, an anti-foaming agent, a monomer system, and an initiator system. The components are combined to form a suspension which is poured into a mold and heat-treated to form a thermoplastic part. The thermoplastic part can then be further densified and heat-treated to produce a high density carbon or graphite composite. The present invention also discloses the products derived from this method.

  20. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

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

    2011-03-01

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

  1. Alternative Passive Decay-Heat Systems for the Advanced High-Temperature Reactor

    SciTech Connect (OSTI)

    Forsberg, Charles W.

    2006-07-01

    The Advanced High-Temperature Reactor (AHTR) is a low-pressure, liquid-salt-cooled high-temperature reactor for the production of electricity and hydrogen. The high-temperature (950 deg C) variant is defined as the liquid-salt-cooled very high-temperature reactor (LS-VHTR). The AHTR has the same safety goals and uses the same graphite-matrix coated particle fuel as do modular high-temperature gas-cooled reactors. However, the large AHTR power output [2400 to 4000 MW(t)] implies the need for a different type of passive decay-heat removal system. Because the AHTR is a low-pressure, liquid-cooled reactor like sodium-cooled reactors, similar types of decay-heat-removal systems can be used. Three classes of passive decay heat removal systems have been identified: the reactor vessel auxiliary cooling system which is similar to that proposed for the General Electric S-PRISM sodium-cooled fast reactor; the direct reactor auxiliary cooling system, which is similar to that used in the Experimental Breeder Reactor-II; and a new pool reactor auxiliary cooling system. These options are described and compared. (author)

  2. Performance of Lithium Ion Cell Anode Graphites Under Various Cycling Conditions

    E-Print Network [OSTI]

    Ridgway, Paul

    2010-01-01

    graphite formulations in particular, are the current standard for lithium-ion anodes for electric vehicle batteries(

  3. Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2010, Prepared for the Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 2012

    SciTech Connect (OSTI)

    D. E. Lewis D. A. Hagemeyer Y. U. McCormick

    2012-07-07

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission’s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 2010 annual reports submitted by five of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Because there are no geologic repositories for high-level waste currently licensed and no NRC-licensed low-level waste disposal facilities currently in operation, only five categories will be considered in this report. The annual reports submitted by these licensees consist of radiation exposure records for each monitored individual. These records are analyzed for trends and presented in this report in terms of collective dose and the distribution of dose among the monitored individuals. Annual reports for 2010 were received from a total of 190 NRC licensees. The summation of reports submitted by the 190 licensees indicated that 192,424 individuals were monitored, 81,961 of whom received a measurable dose. When adjusted for transient workers who worked at more than one licensee during the year, there were actually 142,471 monitored individuals and 62,782 who received a measurable dose. The collective dose incurred by these individuals was 10,617 person-rem, which represents a 12% decrease from the 2009 value. This decrease was primarily due to the decrease in collective dose at commercial nuclear power reactors, as well as a decrease in the collective dose for most of the other categories of NRC licensees. The number of individuals receiving a measurable dose also decreased, resulting in an average measurable dose of 0.13 rem for 2010. The average measurable dose is defined as the total effective dose equivalent (TEDE) divided by the number of individuals receiving a measurable dose. In calendar year 2010, the average annual collective dose per reactor for light water reactor (LWR) licensees was 83 person-rem. This represents a 14% decrease from the value reported for 2009 (96 person-rem). The decrease in collective dose for commercial nuclear power reactors was due to an 11% decrease in total outage hours in 2010. During outages, activities involving increased radiation exposure such as refueling and maintenance are performed while the reactor is not in operation. The average annual collective dose per reactor for boiling water reactors (BWRs) was 137 personrem for 35 BWRs, and 55 person-rem for 69 pressurized water reactors (PWRs). Analyses of transient individual data indicate that 29,333 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient individuals by multiple licensees. The adjustment to account for transient individuals has been specifically noted in footnotes in the figures and tables for commercial nuclear power reactors. In 2010, the average measurable dose per individual for all licensees calculated from reported data was 0.13 rem. Although the average measurable dose per individual from data submitted by licensees was 0.13 rem, a corrected dose distribution resulted in an average measurable dose per individual of 0.17 rem.

  4. Status of ASME Section III Task Group on Graphite Support Core Structures

    SciTech Connect (OSTI)

    Robert L. Bratton; Tim D. Burchell

    2005-08-01

    This report outlines the roadmap that the ASME Project Team on Graphite Core Supports is pursuing to establish design codes for unirradiated and irradiated graphite core components during its first year of operation. It discusses the deficiencies in the proposed Section III, Division 2, Subsection CE graphite design code and the different approaches the Project Team has taken to address those deficiencies.

  5. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    SciTech Connect (OSTI)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next Generation Safeguards Initiative (NGSI).

  6. Reference worldwide model for antineutrinos from reactors

    E-Print Network [OSTI]

    Marica Baldoncini; Ivan Callegari; Giovanni Fiorentini; Fabio Mantovani; Barbara Ricci; Virginia Strati; Gerti Xhixha

    2015-02-16

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency (IAEA). We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO+) and proposed (Juno, RENO-50, LENA and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation and detection processes are estimated using a Monte Carlo based approach, which provides an overall site dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes and we estimate a 2.4% increase of the unoscillated event rate in the geoneutrino energy window due to the storage of spent nuclear fuels in the cooling pools. We predict that the research reactors contribute to less than 0.2% to the commercial reactor signal in the investigated 14 sites. We perform a multitemporal analysis of the expected reactor signal over a time lapse of 10 years using reactor operational records collected in a comprehensive database published at www.fe.infn.it/antineutrino.

  7. Stability analysis of graphite crystal lattice with moment interactions

    E-Print Network [OSTI]

    Krivtsov, Anton M.

    of graphite lattice is considered. The model is based on usage of moment interactions. Carbon atom transversal stiffness to provide the stability of the hexagonal lattice. Proc. of XXXIV Summer School between atoms is r0. The angles between bonds and are shown on Fig.1. Let us present internal energy

  8. High Capacity Graphite Anodes for Li-Ion battery applications

    E-Print Network [OSTI]

    Popov, Branko N.

    High Capacity Graphite Anodes for Li-Ion battery applications using Tin microencapsulation Basker range 1.6V to 0.01V at 0.05 mV/s Physical characterization SEM, EDAX and XRD #12;SEM images of Bare

  9. Progress in Developing Finite Element Models Replicating Flexural Graphite Testing

    SciTech Connect (OSTI)

    Robert Bratton

    2010-06-01

    This report documents the status of flexural strength evaluations from current ASTM procedures and of developing finite element models predicting the probability of failure. This work is covered under QLD REC-00030. Flexural testing procedures of the American Society for Testing and Materials (ASTM) assume a linear elastic material that has the same moduli for tension and compression. Contrary to this assumption, graphite is known to have different moduli for tension and compression. A finite element model was developed and demonstrated that accounts for the difference in moduli tension and compression. Brittle materials such as graphite exhibit significant scatter in tensile strength, so probabilistic design approaches must be used when designing components fabricated from brittle materials. ASTM procedures predicting probability of failure in ceramics were compared to methods from the current version of the ASME graphite core components rules predicting probability of failure. Using the ASTM procedures yields failure curves at lower applied forces than the ASME rules. A journal paper was published in the Journal of Nuclear Engineering and Design exploring the statistical models of fracture in graphite.

  10. Oxygen adsorption on graphite and nanotubes P. Giannozzi,a)

    E-Print Network [OSTI]

    Giannozzi, Paolo

    Oxygen adsorption on graphite and nanotubes P. Giannozzi,a) R. Car, and G. Scoles Chemistry Department and Princeton Materials Institute, Princeton University, Princeton, New Jersey 08544 Received 29 in molecular form,6 with an estimated binding energy E 0.19 eV. This would be consistent with the well

  11. Process for the fabrication of aluminum metallized pyrolytic graphite sputtering targets

    DOE Patents [OSTI]

    Makowiecki, Daniel M. (Livermore, CA); Ramsey, Philip B. (Livermore, CA); Juntz, Robert S. (Hayward, CA)

    1995-01-01

    An improved method for fabricating pyrolytic graphite sputtering targets with superior heat transfer ability, longer life, and maximum energy transmission. Anisotropic pyrolytic graphite is contoured and/or segmented to match the erosion profile of the sputter target and then oriented such that the graphite's high thermal conductivity planes are in maximum contact with a thermally conductive metal backing. The graphite contact surface is metallized, using high rate physical vapor deposition (HRPVD), with an aluminum coating and the thermally conductive metal backing is joined to the metallized graphite target by one of four low-temperature bonding methods; liquid-metal casting, powder metallurgy compaction, eutectic brazing, and laser welding.

  12. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    sensitivities and uncertainties in full-scale multi physics Pressurized Water Reactors (PWR) simulations Research & Development VERA COBRA-TF Dakota Hydra-TH Insilico MAMBA MPACT...

  13. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21

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

  14. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  15. EFFECTS OF GRAPHITE SURFACE ROUGHNESS ON BYPASS FLOW COMPUTATIONS FOR AN HTGR

    SciTech Connect (OSTI)

    Rich Johnson; Yu-Hsin Tung; Hiroyuki Sato

    2011-07-01

    Bypass flow in a prismatic high temperature gas reactor (HTGR) occurs between graphite blocks as they sit side by side in the core. Bypass flow is not intentionally designed to occur in the reactor, but is present because of tolerances in manufacture, imperfect installation and expansion and shrinkage of the blocks from heating and irradiation. It is desired to increase the knowledge of the effects of such flow, which has been estimated to be as much as 20% of the total helium coolant flow. Computational fluid dynamic (CFD) simulations can provide estimates of the scale and impacts of bypass flow. Previous CFD calculations have examined the effects of bypass gap width, level and distribution of heat generation and effects of shrinkage. The present contribution examines the effects of graphite surface roughness on the bypass flow for different relative roughness factors on three gap widths. Such calculations should be validated using specific bypass flow measurements. While such experiments are currently underway for the specific reference prismatic HTGR design for the next generation nuclear plant (NGNP) program of the U. S. Dept. of Energy, the data are not yet available. To enhance confidence in the present calculations, wall shear stress and heat transfer results for several turbulence models and their associated wall treatments are first compared for flow in a single tube that is representative of a coolant channel in the prismatic HTGR core. The results are compared to published correlations for wall shear stress and Nusselt number in turbulent pipe flow. Turbulence models that perform well are then used to make bypass flow calculations in a symmetric onetwelfth sector of a prismatic block that includes bypass flow. The comparison of shear stress and Nusselt number results with published correlations constitutes a partial validation of the CFD model. Calculations are also compared to ones made previously using a different CFD code. Results indicate that increasing surface roughness increases the maximum fuel and helium temperatures as do increases in gap width. However, maximum coolant temperature variation due to increased gap width is not changed by surface roughness.

  16. ELECTRICAL CONDUCTIVITY OF A GRAPHITE BASED COMPOSITE AS AFFECTED BY THE DEGREE OF MIXEDNESS OF GRAPHITE IN

    E-Print Network [OSTI]

    filled polymers and the mechanisms involved in the formation of conductive composites of polymerELECTRICAL CONDUCTIVITY OF A GRAPHITE BASED COMPOSITE AS AFFECTED BY THE DEGREE OF MIXEDNESS of the volume conductivity of the composite. Background Electrically conductive nonmetallic materials are used

  17. Research

    E-Print Network [OSTI]

    My research interests. Numerical method of stochastic partial differential equations; Uncertainty Quantification; High-order numerical method; Domain ...

  18. Research

    E-Print Network [OSTI]

    author

    Research Interests. Mathematical biology: Computational modelling of biological systems, experimental design and control of cellular processes. Applied math: ...

  19. An Investigation of the Effect of Graphite Degradation on the Irreversible Capacity in Lithium-ion Cells

    E-Print Network [OSTI]

    Hardwick, Laurence

    2008-01-01

    graphite anodes suffer severe surface structural damage upon prolonged cycling in rechargeable lithium-ion batteries.

  20. A vectorized heat transfer model for solid reactor cores

    SciTech Connect (OSTI)

    Rider, W.J.; Cappiello, M.W.; Liles, D.R.

    1990-01-01

    The new generation of nuclear reactors includes designs that are significantly different from light water reactors. Among these new reactor designs is the Modular High-Temperature Gas-Cooled Reactor (MHTGR). In addition, nuclear thermal rockets share a number of similarities with terrestrial HTGRs and would be amenable to similar types of analyses. In these reactors, the heat transfer in the solid core mass is of primary interest in design and safety assessment. One significant safety feature of these reactors is the capability to withstand a loss of pressure and forced cooling in the primary system and still maintain peak fuel temperatures below the safe threshold for retaining the fission products. To accurately assess the performance of gas-cooled reactors during these types of transients, a Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions. Also, HERA has been used to analyze a depressurized loss of forced cooling transient in a HTGR with a very detailed three-dimensional input model. The results compare favorably with other means of analysis and provide further validation of the models and methods. 18 refs., 11 figs.

  1. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16

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

  2. Articulated limiter blade for a tokamak fusion reactor

    DOE Patents [OSTI]

    Doll, D.W.

    1982-10-21

    A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

  3. Articulated limiter blade for a tokamak fusion reactor

    DOE Patents [OSTI]

    Doll, David W. (San Diego, CA)

    1985-01-01

    A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

  4. Slurry Molding Technologies for Novel Carbon and Graphite Materials

    SciTech Connect (OSTI)

    Burchell, T.D.

    2004-06-30

    The Oak Ridge National Laboratory (ORNL) has developed a slurry molding technology for the manufacture of porous, high surface area, carbon fiber composites molecular sieves, and carbon-carbon composite preforms. Potentially, this technology could be applied to the manufacture of a host of novel carbon materials including porous adsorbent carbons, low-pressure drop adsorbent carbon composites, ultra-fine-grained graphite, and carbon fiber reinforced graphite. New opportunities for high surface carbon fiber composite molecular sieve (CFCMS) materials are now emerging. Many of these opportunities are driven by increasingly harsh environmental pressures. Traditional granular activated carbon (GAC) is not suitable for many of these applications because of the difficulties encountered with attrition and in forming ''structures'' which have the necessary mechanical and physical properties. In addition, the electrical desorption of adsorbed species is not possible with GAC due to its low bulk electrical conductivity. Activated carbon fibers have been found to be useful in some applications. Work by ORNL has shown, for example, that CFCMS materials are capable of adsorbing various gases and desorbing them under electrical stimulation. For some applications these fibers have to be formed into a structure that can offer the desired mechanical integrity and pressure drop characteristics. To date, the work by ORNL has focused on the use of a single manufacturer's isotropic pitch fibers which, when activated, may be cost prohibitive for many applications. Fine-grained graphite is attractive for many applications including the chemical processing industry where their unique combination of properties--including high strength and chemical inertness, are particularly attractive. However, a lack of toughness can limit their utility in certain applications. The use of ultra-fine powders in conjunction with slurry molding and hot pressing offers the possibility of higher strength graphite. Moreover, the inclusion of carbon fibers may provide a toughening mechanism, resulting in tougher, stronger graphite at an attractive cost. The objective of this work was to further develop the ORNL slurry molding technology and apply it to the following tasks: (1) the development of low cost, high surface area CFCMS materials and structures; (2) the development of ultra-fine-grained graphite; and (3) to identify suitable applications for the materials developed in (1) and (2). The work was conducted jointly by SGL and ORNL.

  5. Catalytic graphitization of carbon aerogels by transition metals

    SciTech Connect (OSTI)

    Maldonado-Hodar, F.J.; Moreno-Castilla, C.; Rivera-Utrilla, J.; Hanzawa, Y.; Yamada, Y.

    2000-05-02

    Carbon aerogels and Cr-, Fe-, Co-, and Ni-containing carbon aerogels were obtained by pyrolysis, at temperatures between 500 and 1,800 C, of the corresponding aerogels prepared by the sol-gel method from polymerization of resorcinol with formaldehyde. All samples were characterized by mercury porosimetry, nitrogen adsorption, X-ray diffraction (XRD), high-resolution transmission electron microscopy (HRTEM), and Raman spectroscopy. Results obtained show that carbon aerogels are, essentially, macroporous materials that maintain large pore volumes even after pyrolysis at 1,800 C. For pyrolysis at temperatures higher than 1,000 C, the presence of the transition metals produced graphitized areas with three-dimensional stacking order, as shown by HRTEM, XRD, and Raman spectroscopy. HRTEM also showed that the metal-carbon containing aerogels were formed by polyhedral structures. Cr and Fe seem to be the best catalysts for graphitization of carbon aerogels.

  6. Carbon K-Edge XANES Spectromicroscopy of Natural Graphite

    SciTech Connect (OSTI)

    Brandes,J.; Cody, G.; Rumble, D.; Haberstroh, P.; Wirick, S.; Gelinas, Y.; Morais-Cabral, J.

    2008-01-01

    The black carbon continuum is composed of a series of carbon-rich components derived from combustion or metamorphism and characterized by contrasting environmental behavior and susceptibility to oxidation. In this work, we present a micro-scale density fractionation method that allows isolating the small quantities of soot-like and graphitic material usually found in natural samples. Organic carbon and {delta}{sup 13}C mass balance calculations were used to quantify the relative contributions of the two fractions to thermally-stable organic matter from a series of aquatic sediments. Varying proportions of soot-like and graphitic material were found in these samples, with large variations in {delta}{sup 13}C signatures suggesting important differences in their origin and/or dynamics in the environment.

  7. Development of a system model for advanced small modular reactors.

    SciTech Connect (OSTI)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandia's concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  8. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    SciTech Connect (OSTI)

    Bernard A. Toseland, Ph.D.

    1999-03-01

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors. The past three months of research have been focused on two major areas of bubble column hydrodynamics: (1) pressure and temperature effects on gas holdup and (2) region transition using a sparger as a gas distributor.

  9. Polymer graphite composite anodes for Li-ion batteries

    E-Print Network [OSTI]

    Popov, Branko N.

    Polymer graphite composite anodes for Li-ion batteries Basker Veeraraghavan, Bala Haran, Ralph analysis #12;TGA analysis of polymer composite SFG10 samples -0.0 150.0 300.0 450.0 600.0 750.0 900-discharge curves of polymer composite SFG10 samples 0 200 400 600 800 Specific Capacity (mAh/g) 0.0 1.0 2.0 3.0 4

  10. Cryogenic Thermal Expansion of Y-12 Graphite Fuel Elements

    SciTech Connect (OSTI)

    Eash, D. T.

    2013-07-08

    Thermal expansion measurements betwccn 20°K and 300°K were made on segments of three uranium-loaded Y-12 uncoated graphite fuel elements. The thermal expansion of these fuel elements over this temperature range is represented by the equation: {Delta}L/L = -39.42 x 10{sup -5} + 1.10 x 10{sup -7} T + 6.47 x 10{sup -9} T{sup 2} - 8.30 x 10{sup -12} T{sup 3}.

  11. Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations

    SciTech Connect (OSTI)

    Margaret A. Marshall; John D. Bess

    2013-02-01

    A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O2 fuelled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  12. Graphite having improved thermal stress resistance and method of preparation

    DOE Patents [OSTI]

    Kennedy, Charles R. (Oak Ridge, TN)

    1980-01-01

    An improved method for fabricating a graphite article comprises the steps of impregnating a coke article by first heating the coke article in contact with a thermoplastic pitch at a temperature within the range of 250.degree.-300.degree. C. at a pressure within the range of 200-2000 psig for at least 4-10 hours and then heating said article at a temperature within the range of 450.degree.-485.degree. C. at a pressure of 200-2000 psig for about 16-24 hours to provide an impregnated article; heating the impregnated article for sufficient time to carbonize the impregnant to provide a second coke article, and graphitizing the second coke article. A graphite having improved thermal stress resistance results when the coke to be impregnated contains 1-3 wt.% sulfur and no added puffing inhibitors. An additional improvement in thermal stress resistance is achieved when the second coke article is heated above about 1400.degree. C. at a rate of at least 10.degree. C./minute to a temperature above the puffing temperature.

  13. CHARACTERIZATION OF GRAPHITE SLEEVES FROM BUGEY 1 EDF PLANT FOR PERMANENT DISPOSAL--MEASUREMENT AND CALCULATION OF SCALING FACTORS FOR DIFFICULT-TO-MEASURE NUCLIDES

    SciTech Connect (OSTI)

    PONCET, Bernard R.

    2003-02-27

    Electricite De France's Bugey-1 reactor, with graphite moderator, was shutdown permanently in 1994. The natural uranium elements are encased in graphite sleeves to facilitate handling. 2,000 m3 of concrete containers, containing non conditioned graphite sleeves, must be characterized and conditioned before shipment to the national repository site called ''Centre de l'Aube''. The characterization work consists in quantifying Difficult-To-Measure nuclides (DTM) by the use of Scaling Factors (SF), which use Co-60 as tracer. Bugey developed an industrial method for the gamma counting of each package to perform easily and rapidly the measurement of the Co-60 content. Depending upon the DTM radionuclide, Co-60 scaling factors are determined, or by measurement on graphite samples (case of C-14, Cl-36, Ni-63, H-3), either by using a calculation technique which is based upon the impurities present in the graphite sleeves. This method is applied for the other pure beta emitters all DTM radionucli des : Ag-108m, Be-10, Ca-41, Cd-109, Cd-113m, Co-57, Cs-135, Cs-137, Eu-155, Fe-55, Gd-153, Mo-93, Nb- 93m, Nb-94, Ni-59, Pd-107, Pm-147, Sm-151, Sn-119m, Sn-121m, Sn-126, Sr-90, Tc-99, V-49 and Zr-93. Calculations use six sleeve history cases : 1 year at 50% power, 2 years at 50 % power, 3 years at 50 % power, 4 years at 50 % power, 1 year at 100 % power and 2 years at 100 % power. The DTM nuclides have been calculated from impurity concentrations for each of these six cases, and the greatest scaling factor has been kept. The calculation is based upon two impurity sets: First impurity set : a reverse activation calculation provides us with the best estimate value of impurities calculated from the measured mean gamma spectrum and from measured scaling factors. It consists in solving a system of simultaneous equations for the impurities as a function of the mean gamma radioactive spectrum and of the measured scaling factors. The concerned calculated impurities are Co, Cl, Li, Ag, Cs, Eu, Fe, Ni, Sb, Sc, Zn and Sn. Second impurity set: The other impurities which were not available by this reverse calculation are originated from the greatest value, which has been measured in the graphite and sometimes by using the detection limit. This method allows us to avoid some detection limit problems and statistical weaknesses. It gets better, cheaper and faster characterization by mixing easy gamma spectrum measurement and simple linear calculation.

  14. UCLA program in reactor studies: The ARIES tokamak reactor study

    SciTech Connect (OSTI)

    Not Available

    1991-01-01

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

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

    SciTech Connect (OSTI)

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions between the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very effective in removing decay heat. By removing the limit on the decay heat removal capability due to the limited available surface area as in a RVACS, the reactor power and power density can be significantly increased, without losing the passive heat removal feature. This paper will introduce the concept of using DRACS to enhance VHTR passive safety and economics. Three design options will be discussed, depending on the cooling pipe locations. Analysis results from a lumped volume based model and CFD simulations will be presented.

  16. The effect of 150?m expandable graphite on char expansion of intumescent fire retardant coating

    SciTech Connect (OSTI)

    Ullah, Sami, E-mail: samichemist1@gmail.com; Shariff, A. M., E-mail: azmish@petronas.com.my, E-mail: azmibustam@petronas.com.my; Bustam, M. A., E-mail: azmish@petronas.com.my, E-mail: azmibustam@petronas.com.my [Research Center for Carbon Dioxide Capture, Department of Chemical Engineering, Universiti Techologi PETRONAS, Bandar Sri Iskandar, Tronoh 31750 Perak (Malaysia); Ahmad, Faiz, E-mail: faizahmadster@gmail.com [Department of Mechanical Engineering, Universiti Techologi PETRONAS, Bandar Sri Iskandar, Tronoh 31750 Perak (Malaysia)

    2014-10-24

    Intumescent is defined as the swelling of certain substances to insulate the underlying substrate when they are heated. In this research work the effect of 150?m expandable graphite (EG) was studied on char expansion, char morphology and char composition of intumescent coating formulations (ICFs). To study the expansion and thermal properties of the coating, nine different formulations were prepared. The coatings were tested at 500 °C for one hour and physically were found very stable and well bound with the steel substrate. The morphology was studied by Scanning Electron Microscopy (SEM). The char composition was analysed by X-ray Diffraction (XRD) and Fourier transform infrared spectroscopy (FTIR) techniques. EG above than 10.8wt% expands the char abruptly with uniform network structure and affect the outer surface of the char.

  17. Development of a thin wall graphite/polyimide composite pressure vessel with a large aspect ratio

    SciTech Connect (OSTI)

    LaJeunesse, C.A.; Trinh, K.V.; Bartel, J.J.; Spingarn, J.R.

    1988-10-01

    A graphite/polyimide composite pressure vessel has been developed to contain corrosive gas mixtures up to 400 psig and 300/degree/F. The goal of this research project was to develop a lightweight, thin-walled, composite pressure vessel with an l/d aspect ratio of 42 capable of containing hydrogen fluoride (HF) and chlorine trifluoride (ClF/sub 3/). The vessel was to have a crown radius approaching infinity, a desired knuckle radius approaching zero, and a desired wall thickness of 0.023 in. In this paper the problems encountered and the iterative solutions in addition to the design, analysis, and fabrication of the vessel are presented. 3 refs., 14 figs., 2 tabs.

  18. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  19. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  20. Process for the fabrication of aluminum metallized pyrolytic graphite sputtering targets

    DOE Patents [OSTI]

    Makowiecki, D.M.; Ramsey, P.B.; Juntz, R.S.

    1995-07-04

    An improved method is disclosed for fabricating pyrolytic graphite sputtering targets with superior heat transfer ability, longer life, and maximum energy transmission. Anisotropic pyrolytic graphite is contoured and/or segmented to match the erosion profile of the sputter target and then oriented such that the graphite`s high thermal conductivity planes are in maximum contact with a thermally conductive metal backing. The graphite contact surface is metallized, using high rate physical vapor deposition (HRPVD), with an aluminum coating and the thermally conductive metal backing is joined to the metallized graphite target by one of four low-temperature bonding methods; liquid-metal casting, powder metallurgy compaction, eutectic brazing, and laser welding. 11 figs.

  1. Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report

    SciTech Connect (OSTI)

    Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

    2006-09-01

    Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (i) chemical reactivity of the ceramics with fission products palladium and rhodium, (ii) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (iii) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin’s 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature up to 800ºC.

  2. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  3. Improved vortex reactor system

    DOE Patents [OSTI]

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

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  4. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-03-02

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  5. Environmental and health effects review for obscurant graphite flakes. Final report, 1991 July--1993 May

    SciTech Connect (OSTI)

    Driver, C.J.; Ligotke, M.W.; Landis, W.G.; Downs, J.L.; Tiller, B.L.; Moore, E.B. Jr.; Cataldo, D.A.

    1993-07-01

    The health and environmental effects of obscurant graphite flakes were reviewed and compared to predicted levels of graphite flake material in the field during typical testing and training scenarios. Graphite flake dispersion and deposition for simulated mechanical and pyrotechnic releases were determined using a modified Gaussian atmospheric plume-dispersion model. The potential for wind resuspension of graphite flakes is controlled by weathering processes and incorporation rates in soil. Chemically, graphite flakes pose little risk to aquatic or terrestrial systems. Mechanical damage to plants and invertebrate and vertebrate organisms from the flakes is also minimal. In humans, the pathological and physiological response to inhaled graphite flake is similar to that induced by nuisance dusts and cause only transient pulmonary changes. Repeated exposure to very high concentrations (such as those near the source generator) may overwhelm the clearance mechanisms of the lung and result in pulmonary damage from the retained particles in unprotected individuals. However, these lesions either resolve with time or are of limited severity. Health effects of mixed aerosols of mixed aerosols of graphite and fog oil are similar to those produced by graphite flakes alone. Environmental impacts of fog oil-coated graphite flakes are not well known.

  6. Surface treated natural graphite as anode material for high-power Li-ion battery applications.

    SciTech Connect (OSTI)

    Liu, J.; Vissers, D. R.; Amine, K.; Barsukov, I. V.; Henry, F.; Doniger, J.; Chemical Engineering; Superior Graphite Co.

    2006-01-01

    High power application of Li-ion battery in hybrid electrical vehicles requires low cost and safe cell materials. Among the various carbon anode materials used in lithium ion batteries, natural graphite shows the most promise with advantages in performance and cost. However, natural graphite is not compatible with propylene carbonate (PC)-based electrolytes, which have a lower melting point and improved safety characteristics. The problem with it is that the molecules of propylene carbonate intercalate with Li+ into graphite, and that frequently leads to the exfoliation of the graphite matrix.

  7. Raman spectroscopy of graphite in high magnetic fields: Electron-phonon coupling and magnetophonon resonance

    SciTech Connect (OSTI)

    Kim, Younghee; Smirnov, Dmitry; Kalugin, Nikolai G.; Lombardo, Antonio; Ferrari, Andrea C.

    2013-12-04

    The magneto-Raman measurements of graphite were performed in a back-scattering Faraday geometry at temperature 10 K in magnetic fields up to 45 T. The experimental data reveal the rich structure of Raman-active excitations dominated by K-point massive electrons. At high magnetic fields the graphite E{sub 2g} Raman line shows complex multi- component behavior interpreted as magnetophonon resonance coupled electron-phonon modes at graphite’s K-point. Also we found the clear signature of the fundamental, strongly dumped, n=0 magnetophonon resonance associated with H point massless holes.

  8. Research

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration wouldMassR&D100 Winners *ReindustrializationEnergyWindNO.RequirementsResearch Research

  9. Reactor vessel support system

    DOE Patents [OSTI]

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

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  10. Spinning fluids reactor

    DOE Patents [OSTI]

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

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  11. Reactor water cleanup system

    DOE Patents [OSTI]

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

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  12. Reactor water cleanup system

    DOE Patents [OSTI]

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

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  13. Enhanced performance of graphite anode materials by AlF3 coating for lithium-ion batteries

    SciTech Connect (OSTI)

    Ding, Fei; Xu, Wu; Choi, Daiwon; Wang, Wei; Li, Xiaolin; Engelhard, Mark H.; Chen, Xilin; Yang, Zhenguo; Zhang, Jiguang

    2012-04-27

    In order to form the stable surface film and to further enhance the long-term cycling stability of the graphite anodes of lithium-ion batteries, the surface of graphite powders has been modified by AlF3 coating through chemical precipitation method. The AlF3-coated graphite shows no evident changes in the bulk structure and a thin AlF3-coating layer of about 2 nm thick is found to uniformly cover the graphite particles with 2 wt% AlF3 content. However, it delivers a higher initial discharge capacity and largely improved rate performances compared to the pristine graphite. Remarkably, AlF3 coated graphite demonstrated a much better cycle life. After 300 cycles, AlF3 coated graphite and uncoated graphite show capacity retention of 92% and 81%, respectively. XPS measurement shows that a more conductive solid electrode interface (SEI) layer was formed on AlF3 coated graphite as compared to uncoated graphite. SEM monograph also reveals that the AlF3-coated graphite particles have a much more stable surface morphology after long-term cycling. Therefore, the improved electrochemical performance of AlF3 coated graphite can be attributed to a more stable and conductive SEI formed on coated graphite anode during cycling process.

  14. Method of fabricating silicon carbide coatings on graphite surfaces

    DOE Patents [OSTI]

    Varacalle, D.J. Jr.; Herman, H.; Burchell, T.D.

    1994-07-26

    The vacuum plasma spray process produces well-bonded, dense, stress-free coatings for a variety of materials on a wide range of substrates. The process is used in many industries to provide for the excellent wear, corrosion resistance, and high temperature behavior of the fabricated coatings. In this application, silicon metal is deposited on graphite. This invention discloses the optimum processing parameters for as-sprayed coating qualities. The method also discloses the effect of thermal cycling on silicon samples in an inert helium atmosphere at about 1,600 C which transforms the coating to silicon carbide. 3 figs.

  15. Stacking fault induced tunnel barrier in platelet graphite nanofiber

    SciTech Connect (OSTI)

    Lan, Yann-Wen, E-mail: chiidong@phys.sinica.edu.tw, E-mail: ywlan@phys.sinica.edu.tw; Chang, Yuan-Chih; Chang, Chia-Seng; Chen, Chii-Dong, E-mail: chiidong@phys.sinica.edu.tw, E-mail: ywlan@phys.sinica.edu.tw [Institute of Physics, Academia Sinica, Taipei 115, Taiwan (China); Chang, Wen-Hao [Institute of Physics, Academia Sinica, Taipei 115, Taiwan (China); Graduate Institute of Opto-Mechatronics, National Chung Cheng University, Chia-Yi 62102, Taiwan (China); Li, Yuan-Yao [Graduate Institute of Opto-Mechatronics, National Chung Cheng University, Chia-Yi 62102, Taiwan (China)

    2014-09-08

    A correlation study using image inspection and electrical characterization of platelet graphite nanofiber devices is conducted. Close transmission electron microscopy and diffraction pattern inspection reveal layers with inflection angles appearing in otherwise perfectly stacked graphene platelets, separating nanofibers into two domains. Electrical measurement gives a stability diagram consisting of alternating small-large Coulomb blockade diamonds, suggesting that there are two charging islands coupled together through a tunnel junction. Based on these two findings, we propose that a stacking fault can behave as a tunnel barrier for conducting electrons and is responsible for the observed double-island single electron transistor characteristics.

  16. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect (OSTI)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

  17. Next Generation Nuclear Plant Materials Research and Development Program Plan

    SciTech Connect (OSTI)

    G. O. Hayner; E.L. Shaber

    2004-09-01

    The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years.

  18. LBB application in the US operating and advanced reactors

    SciTech Connect (OSTI)

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

    1997-04-01

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

  19. Simulation of Fischer-Tropsch Fixed-Bed Reactor in Different Reaction Media 

    E-Print Network [OSTI]

    Bani Nasser, Laial Ahmad

    2013-11-04

    reactor), while simultaneously overcoming their limitations. This work focuses on modeling the behavior in the reactor bed (‘macro-scale’ assessment) and then zooming into the catalyst pellet itself (‘micro-scale’ assessment). The aim of this research...

  20. Research

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home RoomPreservation of Fe(II) by Carbon-RichProtonAboutNuclearPrincipal InvestigatorsResearch

  1. Heterogeneous Recycling in Fast Reactors

    SciTech Connect (OSTI)

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  2. Graphite fountain: Modeling of growth on transition metals under a thermal Jongpil Ye and Rodney S. Ruoff

    E-Print Network [OSTI]

    ://dx.doi.org/10.1063/1.4812730] I. INTRODUCTION Graphite has long been in high demand in the steel, automotive storage devices.1,2 Commercial graphite is generally synthesized by mixing and thermally annealing

  3. Analysis of Granular Flow in a Pebble-Bed Nuclear Reactor

    E-Print Network [OSTI]

    Chris H. Rycroft; Gary S. Grest; James W. Landry; Martin Z. Bazant

    2006-02-16

    Pebble-bed nuclear reactor technology, which is currently being revived around the world, raises fundamental questions about dense granular flow in silos. A typical reactor core is composed of graphite fuel pebbles, which drain very slowly in a continuous refueling process. Pebble flow is poorly understood and not easily accessible to experiments, and yet it has a major impact on reactor physics. To address this problem, we perform full-scale, discrete-element simulations in realistic geometries, with up to 440,000 frictional, viscoelastic 6cm-diameter spheres draining in a cylindrical vessel of diameter 3.5m and height 10m with bottom funnels angled at 30 degrees or 60 degrees. We also simulate a bidisperse core with a dynamic central column of smaller graphite moderator pebbles and show that little mixing occurs down to a 1:2 diameter ratio. We analyze the mean velocity, diffusion and mixing, local ordering and porosity (from Voronoi volumes), the residence-time distribution, and the effects of wall friction and discuss implications for reactor design and the basic physics of granular flow.

  4. Study of polypyrrole graphite composite as anode material for secondary lithium-ion batteries

    E-Print Network [OSTI]

    Popov, Branko N.

    Study of polypyrrole graphite composite as anode material for secondary lithium-ion batteries; Irreversible capacity; Anode material; Lithium-ion batteries 1. Introduction To ensure long cycle life for the Li-ion battery. Of various carbon materials that have been tried, graphite is favored because it (i

  5. Synthesis of graphene-based nanosheets via chemical reduction of exfoliated graphite oxide

    E-Print Network [OSTI]

    for the preparation of graphene sheets from graphite. After numerous failed attempts to create graphene-based sheetsSynthesis of graphene-based nanosheets via chemical reduction of exfoliated graphite oxide Sasha of a colloidal suspension of exfoliated graphene oxide sheets in water with hydrazine hydrate results

  6. Measurements of chemical erosion of ATJ graphite by low energy D

    E-Print Network [OSTI]

    Abstract We report on initial results of chemical sputtering of ATJ graphite by impact of Dþ 2Measurements of chemical erosion of ATJ graphite by low energy Dþ 2 impact F.W. Meyer *, H. PACS: 34.50.Dy; 52.20.Hv; 52.40.Hf; 79.20.Àm; 79.20.Rf Keywords: Carbon-based materials; Chemical

  7. Laser vaporization clusters from metal loaded graphite used for SWNT generation

    E-Print Network [OSTI]

    Maruyama, Shigeo

    Laser vaporization clusters from metal loaded graphite used for SWNT generation Toshikazu Mukaea, The Univ. of Tokyo In order to study the generation mechanism of SWNTs, we have studied metal-carbon binary clusters generated by the laser vaporization of Ni/Co and Ni/Y doped graphite materials used

  8. Emission characteristics and dynamics of C2 from laser produced graphite plasma

    E-Print Network [OSTI]

    Harilal, S. S.

    Emission characteristics and dynamics of C2 from laser produced graphite plasma S. S. Harilal, Riju 1996; accepted for publication 20 December 1996 The emission features of laser ablated graphite plume diagnostic technique. Time resolved optical emission spectroscopy is employed to reveal the velocity

  9. Diffusion within a Layered, Graphite-Like, Spherical Electrode Theoretical Aspects

    E-Print Network [OSTI]

    Taylor, Philip L.

    Diffusion within a Layered, Graphite-Like, Spherical Electrode Theoretical Aspects Paul B. Antohi. Within this model, which mimics a layered, graphite-like spherical electrode, species can enter or leave further performance optimization of Li+ batteries have generated renewed interest into experimental

  10. Calendar aging of a Graphite/LiFePO4 cell M. Kassem,a

    E-Print Network [OSTI]

    Boyer, Edmond

    1 Calendar aging of a Graphite/LiFePO4 cell M. Kassem,a J. Bernard,b R. Revel,b S. Pélissier,c F Rochelle, France Abstract Graphite/LFP commercial cells are stored under 3 different conditions increase is 70% or less. 1. Introduction Lithium-ion batteries are now the dominant rechargeable systems

  11. Coulomb Oscillations and Hall Effect in Quasi-2D Graphite Quantum Dots

    E-Print Network [OSTI]

    McEuen, Paul L.

    Coulomb Oscillations and Hall Effect in Quasi-2D Graphite Quantum Dots J. Scott Bunch, Yuval Yaish-temperature electrical transport measurements on gated, quasi-2D graphite quantum dots. In devices with low contact of graphene, a zero band gap semiconductor with two linearly dispersing bands that touch at the corners

  12. Hydrogen storage material and process using graphite additive with metal-doped complex hydrides

    DOE Patents [OSTI]

    Zidan, Ragaiy (Aiken, SC); Ritter, James A. (Lexington, SC); Ebner, Armin D. (Lexington, SC); Wang, Jun (Columbia, SC); Holland, Charles E. (Cayce, SC)

    2008-06-10

    A hydrogen storage material having improved hydrogen absorbtion and desorption kinetics is provided by adding graphite to a complex hydride such as a metal-doped alanate, i.e., NaAlH.sub.4. The incorporation of graphite into the complex hydride significantly enhances the rate of hydrogen absorbtion and desorption and lowers the desorption temperature needed to release stored hydrogen.

  13. PHYSICAL REVIEW B 86, 064302 (2012) Polarons in highly doped atomically thin graphitic materials

    E-Print Network [OSTI]

    Hague, Jim

    2012-01-01

    can be mechanically exfoliated, including SnS2, CdI2, and MoS2, but these have the chalcogenidePHYSICAL REVIEW B 86, 064302 (2012) Polarons in highly doped atomically thin graphitic materials J are computed for highly doped graphene-on-substrate and other atomically thin graphitic systems using

  14. Chemical sputtering of ATJ graphite induced by low-energy D

    E-Print Network [OSTI]

    on the surface temperature and the hydrogen impact energy. Previ- ous hydrogen impact results obtained using pure sputtering processes that occur when hydrogen interacts with the low-density ATJ form of graphite ($1.7 gChemical sputtering of ATJ graphite induced by low-energy Dþ 2 bombardment L.I. Vergara *, F

  15. Micro -Thermonuclear AB-Reactors for Aerospace

    E-Print Network [OSTI]

    Alexander Bolonkin

    2007-01-08

    The author offers several innovations that he first suggested publicly early in 1983 for the AB multi-reflex engine, space propulsion, getting energy from plasma, etc. (see: A. Bolonkin, Non-Rocket Space Launch and Flight, Elsevier, London, 2006, Chapters 12, 3A). It is the micro-thermonuclear AB-Reactors. That is new micro-thermonuclear reactor with very small fuel pellet that uses plasma confinement generated by multi-reflection of laser beam or its own magnetic field. The Lawson criterion increases by hundreds of times. The author also suggests a new method of heating the power-making fuel pellet by outer electric current as well as new direct method of transformation of ion kinetic energy into harvestable electricity. These offered innovations dramatically decrease the size, weight and cost of thermonuclear reactor, installation, propulsion system and electric generator. Non-industrial countries can produce these researches and constructions. Currently, the author is researching the efficiency of these innovations for two types of the micro-thermonuclear reactors: multi-reflection reactor (ICF) and self-magnetic reactor (MCF).

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

    SciTech Connect (OSTI)

    GERBER MS

    2009-04-28

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

  17. Modeling for Anaerobic Fixed-Bed Biofilm Reactors

    SciTech Connect (OSTI)

    Liu, B. Y. M.; Pfeffer, J. T.

    1989-06-01

    The specific objectives of this research were: 1. to develop an equilibrium model for chemical aspects of anaerobic reactors; 2. to modify the equilibrium model for non-equilibrium conditions; 3. to incorporate the existing biofilm models into the models above to study the biological and chemical behavior of the fixed-film anaerobic reactors; 4. to experimentally verify the validity of these models; 5. to investigate the biomass-holding ability of difference packing materials for establishing reactor design criteria.

  18. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    SciTech Connect (OSTI)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  19. Effects of Stone-Wales and vacancy defects in atomic-scale friction on defective graphite

    SciTech Connect (OSTI)

    Sun, Xiao-Yu; Wu, RunNi; Xia, Re; Chu, Xi-Hua; Xu, Yuan-Jie

    2014-05-05

    Graphite is an excellent solid lubricant for surface coating, but its performance is significantly weakened by the vacancy or Stone-Wales (SW) defect. This study uses molecular dynamics simulations to explore the frictional behavior of a diamond tip sliding over a graphite which contains a single defect or stacked defects. Our results suggest that the friction on defective graphite shows a strong dependence on defect location and type. The 5-7-7-5 structure of SW defect results in an effectively negative slope of friction. For defective graphite containing a defect in the surface, adding a single vacancy in the interior layer will decrease the friction coefficients, while setting a SW defect in the interior layer may increase the friction coefficients. Our obtained results may provide useful information for understanding the atomic-scale friction properties of defective graphite.

  20. Improved vortex reactor system

    DOE Patents [OSTI]

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

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  1. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  2. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani (Karhula, FI)

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  3. Tokamak reactor first wall

    DOE Patents [OSTI]

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

    1984-11-20

    This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

  4. DECAY HEAT CONDITIONS OF CURRENT AND NEXT GENERATION REACTORS 

    E-Print Network [OSTI]

    Choe, JongSoo 1985-

    2012-05-04

    the accident in Japan last year. Thus, decay heat must be considered in reactor design for safety. The research focused on decay heat conditions of current and next generation reactors. US-APWR, ABWR, VHTR, and ABR were modeled and simulated using the program...

  5. Catalytic Dehydrogenation of Propane in Hydrogen Permselective Membrane Reactors

    E-Print Network [OSTI]

    Brinker, C. Jeffrey

    Catalytic Dehydrogenation of Propane in Hydrogen Permselective Membrane Reactors John P. Collins and Production, Amoco Research Center, 150 West Warrenville Road, Naperville, Illinois 60566-7011 Propane operated at liquid hourly space velocities (LHSVs) similar to those used in commercial reactors for propane

  6. Basic Engineering Research for D and D of R Reactor Storage Pond Sludge: Electrokinetics, Carbon Dioxide Extraction, and Supercritical Water Oxidation

    SciTech Connect (OSTI)

    Michael A. Matthews; David A. Bruce,; Thomas A. Davis; Mark C. Thies; John W. Weidner; Ralph E. White

    2002-04-01

    Large quantities of mixed low level waste (MLLW) that fall under the Toxic Substances Control Act (TSCA) exist and will continue to be generated during D and D operations at DOE sites across the country. The standard process for destruction of MLLW is incineration, which has an uncertain future. The extraction and destruction of PCBs from MLLW was the subject of this research Supercritical Fluid Extraction (SFE) with carbon dioxide with 5% ethanol as cosolvent and Supercritical Waster Oxidation (SCWO) were the processes studied in depth. The solid matrix for experimental extraction studies was Toxi-dry, a commonly used absorbent made from plant material. PCB surrogates were 1.2,4-trichlorobenzene (TCB) and 2-chlorobiphenyl (2CBP). Extraction pressures of 2,000 and 4,000 psi and temperatures of 40 and 80 C were studied. Higher extraction efficiencies were observed with cosolvent and at high temperature, but pressure little effect. SCWO treatment of the treatment of the PCB surrogates resulted in their destruction below detection limits.

  7. The Thorium Molten Salt Reactor : Moving on from the MSBR

    E-Print Network [OSTI]

    L. Mathieu; D. Heuer; R. Brissot; C. Le Brun; E. Liatard; J. M. Loiseaux; O. Méplan; E. Merle-Lucotte; A. Nuttin; J. Wilson; C. Garzenne; D. Lecarpentier; E. Walle; the GEDEPEON Collaboration

    2005-06-02

    A re-evaluation of the Molten Salt Breeder Reactor concept has revealed problems related to its safety and to the complexity of the reprocessing considered. A reflection is carried out anew in view of finding innovative solutions leading to the Thorium Molten Salt Reactor concept. Several main constraints are established and serve as guides to parametric evaluations. These then give an understanding of the influence of important core parameters on the reactor's operation. The aim of this paper is to discuss this vast research domain and to single out the Molten Salt Reactor configurations that deserve further evaluation.

  8. Direct printing and reduction of graphite oxide for flexible supercapacitors

    SciTech Connect (OSTI)

    Jung, Hanyung; Ve Cheah, Chang; Jeong, Namjo; Lee, Junghoon

    2014-08-04

    We report direct printing and photo-thermal reduction of graphite oxide (GO) to obtain a highly porous pattern of interdigitated electrodes, leading to a supercapacitor on a flexible substrate. Key parameters optimized include the amount of GO delivered, the suitable photo-thermal energy level for effective flash reduction, and the substrate properties for appropriate adhesion after reduction. Tests with supercapacitors based on the printed-reduced GO showed performance comparable with commercial supercapacitors: the energy densities were 1.06 and 0.87 mWh/cm{sup 3} in ionic and organic electrolytes, respectively. The versatility in the architecture and choice of substrate makes this material promising for smart power applications.

  9. A graphite-prism definition for Avogadro's "integer"

    E-Print Network [OSTI]

    P. Fraundorf; Melanie Lipp

    2015-08-05

    The new International System of Units may let us select an integer value for Avogadro's number. Some might prefer an integer that's divisible by 12, so that an integer number of $^{12}C$ atoms may be associated (to first order) with a gram's mass. For educational and practical reasons it may also help to choose a {\\em physically-meaningful} definition within measurement error of the current numeric value. Cubes of diamond face-centered-cubic Si and (much rarer) face-centered-cubic C have been proposed, but these structures do not have naturally-occurring facets (or numbers of atoms generally divisible by 12). We show here that graphite prisms formed by stacking $m$ hexagonal graphene sheets, with $m \\equiv 51,150,060$ carbon-12 atoms on each side, are a natural solution that may facilitate generation of precise molar standards as well.

  10. Method of producing exfoliated graphite composite compositions for fuel cell flow field plates

    DOE Patents [OSTI]

    Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

    2014-04-08

    A method of producing an electrically conductive composite composition, which is particularly useful for fuel cell bipolar plate applications. The method comprises: (a) providing a supply of expandable graphite powder; (b) providing a supply of a non-expandable powder component comprising a binder or matrix material; (c) blending the expandable graphite with the non-expandable powder component to form a powder mixture wherein the non-expandable powder component is in the amount of between 3% and 60% by weight based on the total weight of the powder mixture; (d) exposing the powder mixture to a temperature sufficient for exfoliating the expandable graphite to obtain a compressible mixture comprising expanded graphite worms and the non-expandable component; (e) compressing the compressible mixture at a pressure within the range of from about 5 psi to about 50,000 psi in predetermined directions into predetermined forms of cohered graphite composite compact; and (f) treating the so-formed cohered graphite composite to activate the binder or matrix material thereby promoting adhesion within the compact to produce the desired composite composition. Preferably, the non-expandable powder component further comprises an isotropy-promoting agent such as non-expandable graphite particles. Further preferably, step (e) comprises compressing the mixture in at least two directions. The method leads to composite plates with exceptionally high thickness-direction electrical conductivity.

  11. Design of an Alternative Coolant Inlet Flow Configuaration for the Modular Helium Reactor

    SciTech Connect (OSTI)

    SM Mohsin Reza; E. A. Harvego; Matt Richards; Arkal Shenoy; Kenneth Lee Peddicord

    2006-06-01

    The coolant outlet temperature for the Modular Helium Reactor (MHR) was increased to improve the overall efficiency of nuclear hydrogen production using either thermochemical or high temperature electrolysis (HTE) processes. The inlet temperature was also increased to keep about the same _T across the reactor core. Thermal hydraulic analyses of the current MHR design were performed with these updated temperatures to determine the impact of these highter temperatures on pressure drops, coolant flow rates and temperature profiles within the vessel and core regions. Due to these increased operating temperatures, the overall efficiency of hydrogen production processes increases but the steady state reactor vessel temperature is found to be well above the ASME code limits for current vessel materials. Using the RELAP5-3D/ATHENA computer code, an alternative configuration for the MHR coolant inlet flow path was evaluated in an attempt to reduce the reactor vessel temperatures. The coolant inlet flow was shifted from channel boxes located in the annular region between the reactor core barrel and the inner wall of the reactor vessel to a flow path through the outer permanent reflector. Considering the available thickness of graphite in the permanent outer reflector, the total flow area, the number of coolant holes and the coolant-hole diameter were varied to optimize the pressure drop, the coolant inlet velocity and the percentage of graphite removed from the core. The resulting thermal hydraulic analyses of the optimized design showed that peak vessel and fuel temperatures were within acceptable limits for both steady-state and transient operating conditions.

  12. Improved Prediction of the Temperature Feedback in TRISO-Fueled Reactors

    SciTech Connect (OSTI)

    Javier Ortensi; Abderrafi M. Ougouag

    2009-08-01

    The Doppler feedback mechanism is a major contributor to the passive safety of gas-cooled, graphite-moderated high temperature reactors that use fuel based on Tristructural-Isotropic coated particles. It follows that the correct prediction of the magnitude and time-dependence of this feedback effect is essential to the conduct of safety analyses for these reactors. We present a fuel conduction model for obtaining better estimates of the temperature feedback during moderate and fast transients. The fuel model has been incorporated in the CYNOD-THERMIX-KONVEK suite of coupled codes as a single TRISO particle within each calculation cell. The heat generation rate is scaled down from the neutronic solution and a Dirichlet boundary condition is imposed as the bulk graphite temperature from the thermal-hydraulic solution. This simplified approach yields similar results to those obtained with more complex methods, requiring multi-TRISO calculations within one control volume, but with much less computational effort. An analysis of the hypothetical total control ejection in the PBMR-400 design verifies the performance of the code during fast transients. In addition, the analysis of the earthquake-initiated event in the PBMR-400 design verifies the performance of the code during slow transients. These events clearly depict the improvement in the predictions of the fuel temperature, and consequently, of the power escalations. In addition, a brief study of the potential effects of particle layer de-bonding on the transient behavior of high temperature reactors is included. Although the formation of a gap occurs under special conditions its consequences on the dynamic behavior of the reactor should be analyzed. The presence of a gap in the fuel can cause some unusual reactor behavior during fast transients, but still the reactor shuts down due to the strong feedback effects.

  13. Synthesis and characterization of graphite-metal fluoride intercalation compounds

    SciTech Connect (OSTI)

    McQuillan, B.W.

    1981-04-01

    The compound C/sub x/AsF/sub 5/ was prepared and characterized by x-ray diffraction and x-ray absorption, which show the presence of As(III) and As(V), and the As-F bond distances are consistent with AsF/sub 3/ and AsF/sup -//sub 6/, C/sup +//sub 8/AsF/sup -//sub 6/ and C/sup +//sub 8/OsF/sup -//sub 6/ were synthesized. The C/sub x/AsF/sub 6/ and numerous standard arsenic-flourine compounds were studied by x-ray absorption. Magnetic susceptibility of C/sup +//sub 8/OsF/sup -//sub 6/ confirms the high degree of oxidation in this compound. X-ray absorption studies were begun to determine the species present within the graphite when BrF/sub 3/ or GeF/sub 4/ + F/sub 2/ are added. In the BrF/sub 3/ case, Br/sub 2/ is evolved and only Br(III) is present in the graphite. The binary phase diagram XeF/sub 2/:Xe/sup +//sub 5/AsF/sup -//sub 6/ was investigated by melting point determination and Raman spectroscopy. This mixture near 1.3:1 forms a kinetically stable glass at room temperature and is molten at 50/sup 0/C. Several new species or phases are observed in the Raman spectra. These species have been assigned tentative structures.

  14. Graphite fuels combustion off-gas treatment options

    SciTech Connect (OSTI)

    Kirkham, R.J.; Lords, R.E.

    1993-03-01

    Scenarios for burning bulk graphite and for burning crushed fuel particles from graphite spent nuclear fuels have been considered. Particulates can be removed with sintered metal filters. Subsequent cooling would then condense semi-volatile fission products into or onto a particulate. These particulates would be trapped by a second sintered metal filter or downstream packed bed. A packed bed scrub column can be used to eliminate most of the iodine-129 and tritium. A molecular sieve bed is proposed to collect the residual {sup 129}I and other tramp radionuclides downstream (Ruthenium, etc.). Krypton-85 can be recovered, if need be, either by cryogenics or by the KALC process (Krypton Adsorption in Liquid Carbon dioxide). Likewise carbon-14 in the form of carbon dioxide could be collected with a caustic or lime scrub solution and incorporated into a grout. Sulfur dioxide present will be well below regulatory concern level of 4.0 tons per year and most of it would be removed by the scrubber. Carbon monoxide emissions will depend on the choice of burner and start-up conditions. Should the system exceed the regulatory concern level, a catalytic converter in the final packed bed will be provided. Radon and its daughters have sufficiently short half-lives (less than two minutes). If necessary, an additional holdup bed can be added before the final HEPA filters or additional volume can be added to the molecular sieve bed to limit radon emissions. The calculated total effective dose equivalent at the Idaho National Engineering Laboratory boundary from a single release of all the {sup 3}, {sup 14}C, {sup 85}Kr, and {sup 129}I in the total fuel mass if 0.43 mrem/year.

  15. Nuclear Reactor Safeguards and Monitoring with Antineutrino Detectors

    E-Print Network [OSTI]

    Adam Bernstein; Yifang Wang; Giorgio Gratta; Todd West

    2001-08-01

    Cubic-meter-sized antineutrino detectors can be used to non-intrusively, robustly and automatically monitor and safeguard a wide variety of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors. Since the antineutrino spectra and relative yields of fissioning isotopes depend on the isotopic composition of the core, changes in composition can be observed without ever directly accessing the core itself. Information from a modest-sized antineutrino detector, coupled with the well-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being operated in an illegitimate way. A group at Sandia is currently constructing a one cubic meter antineutrino detector at the San Onofre reactor site in California to demonstrate these principles.

  16. Graphite-ceramic rf Faraday-thermal shield and plasma limiter

    DOE Patents [OSTI]

    Hwang, D.L.Q.; Hosea, J.C.

    1983-05-05

    The present invention is directed to a brazing procedure for joining a ceramic or glass material (e.g., Al/sub 2/O/sub 3/ or Macor) to graphite. In particular, the present invention is directed to a novel brazing procedure for the production of a brazed ceramic graphite product useful as a Faraday shield. The brazed ceramic graphite Faraday shield of the present invention may be used in Magnetic Fusion Devices (e.g., Princeton Large Torus Tokamak) or other high temperature resistant apparatus.

  17. High density-high purity graphite prepared by hot isostatic pressing in refractory metal containers

    DOE Patents [OSTI]

    Hoenig, Clarence L. (Livermore, CA)

    1994-01-01

    Porous graphite in solid form is hot isostatically pressed in a refractory metal container to produce a solid graphite monolith with a bulk density greater than or equal to 2.10 g/cc. The refractory metal container is formed of tantalum, niobium, tungsten, molybdenum or alloys thereof in the form of a canister or alternatively plasma sprayed, chemically vapor deposited, or coated by some other suitable means onto graphite. Hot isostatic pressing at 2200.degree. C. and 30 KSI (206.8 MPa) argon pressure for two hours produces a bulk density of 2.10 g/cc. Complex shapes can be made.

  18. High density-high purity graphite prepared by hot isostatic pressing in refractory metal containers

    DOE Patents [OSTI]

    Hoenig, C.L.

    1994-08-09

    Porous graphite in solid form is hot isostatically pressed in a refractory metal container to produce a solid graphite monolith with a bulk density greater than or equal to 2.10 g/cc. The refractory metal container is formed of tantalum, niobium, tungsten, molybdenum or alloys thereof in the form of a canister or alternatively plasma sprayed, chemically vapor deposited, or coated by some other suitable means onto graphite. Hot isostatic pressing at 2,200 C and 30 KSI (206.8 MPa) argon pressure for two hours produces a bulk density of 2.10 g/cc. Complex shapes can be made. 1 fig.

  19. Microchannel Reactor System for Catalytic Hydrogenation

    SciTech Connect (OSTI)

    2004-07-01

    This factsheet describes a research project whose goal is to design, fabricate, evaluate, and optimize a laboratory-scale microchannel reactor/heat exchanger system with thin-film or particulate catalysts for hydrogenation of o-nitroanisole and other nitro aromatic compounds, under moderate temperature and pressure.

  20. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    SciTech Connect (OSTI)

    Ruger, C.J.; Higgins, J.C.

    1993-11-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970`s and early 1980`s raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants.

  1. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M. (Plum Borough, PA)

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  2. Slurry reactor design studies

    SciTech Connect (OSTI)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  3. Production capabilities in US nuclear reactors for medical radioisotopes

    SciTech Connect (OSTI)

    Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr. [Oak Ridge National Lab., TN (United States); Schenter, R.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1992-11-01

    The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted.

  4. The HIgh Flux Isotope Reactor: Past, Present, and Future

    SciTech Connect (OSTI)

    Beierschmitt, Kelly J [ORNL; Farrar, Mike B [ORNL

    2009-01-01

    HFIR construction began in 1965 and completed in 1966. During the first 15 years of operation, the heavy actinide isotope production mission was dominant. HFIR is now positioned as one of the most versataile research reactors in the world.

  5. Computational neutronics analysis of TRIGA reactors during power pulsing

    E-Print Network [OSTI]

    Bean, Malcolm (Malcolm K.)

    2011-01-01

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

  6. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

    SciTech Connect (OSTI)

    Forsberg, C. W.; Terrani, Kurt A; Snead, Lance Lewis; Katoh, Yutai

    2012-01-01

    The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of {sup 7}LiF and BeF{sub 2} (FLiBe) possessing a boiling point above 1300 C and the figure of merit {rho}C{sub p} (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

  7. Fast Breeder Reactor studies

    SciTech Connect (OSTI)

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

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  8. Nuclear reactor reflector

    DOE Patents [OSTI]

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

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  9. Nuclear reactor reflector

    DOE Patents [OSTI]

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

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  10. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  11. Microfluidic electrochemical reactors

    DOE Patents [OSTI]

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

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  12. Research and Development Technology Development Roadmaps for the Next Generation Nuclear Plant Project

    SciTech Connect (OSTI)

    Ian McKirdy

    2011-07-01

    The U.S. Department of Energy (DOE) has selected the high temperature gas-cooled reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for process heat, hydrogen and electricity production. The reactor will be graphite moderated with helium as the primary coolant and may be either prismatic or pebble-bed. Although, final design features have not yet been determined. Research and Development (R&D) activities are proceeding on those known plant systems to mature the technology, codify the materials for specific applications, and demonstrate the component and system viability in NGNP relevant and integrated environments. Collectively these R&D activities serve to reduce the project risk and enhance the probability of on-budget, on-schedule completion and NRC licensing. As the design progresses, in more detail, toward final design and approval for construction, selected components, which have not been used in a similar application, in a relevant environment nor integrated with other components and systems, must be tested to demonstrate viability at reduced scales and simulations prior to full scale operation. This report and its R&D TDRMs present the path forward and its significance in assuring technical readiness to perform the desired function by: Choreographing the integration between design and R&D activities; and proving selected design components in relevant applications.

  13. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    SciTech Connect (OSTI)

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  14. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect (OSTI)

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  15. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect (OSTI)

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  16. Laboratory Directed Research and Development Program FY 2001

    E-Print Network [OSTI]

    Hansen, Todd; Levy, Karin

    2002-01-01

    compact, inexpensive, and safe to operate. The current options for Accelerator and Fusion Research Division neutron sources are reactors

  17. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    SciTech Connect (OSTI)

    Gauld, Ian C; Sly, Nicholas C; Michel-Sendis, Franco

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

  18. Improved Design of Nuclear Reactor Control System | U.S. DOE...

    Office of Science (SC) Website

    Improved Design of Nuclear Reactor Control System Nuclear Physics (NP) NP Home About Research Facilities Science Highlights Benefits of NP Applications of Nuclear Science...

  19. Loop formation in graphitic nanoribbon edges using furnace heating or Joule heating

    E-Print Network [OSTI]

    Jia, Xiaoting

    Here the authors report the use of either furnace heating or Joule heating to pacify the exposed graphene edges by loop formation in a novel graphitic nanoribbonmaterial, grown by chemical vapor deposition. The edge energy ...

  20. Graphitization of small diamond cluster --Molecular dynamics simulation A. Brdka a,, T.W. Zerda b

    E-Print Network [OSTI]

    Powles, Rebecca

    Elsevier B.V. All rights reserved. Keywords: Nanodiamond; Graphitization; Molecular dynamics simulation 1 of nanodiamond clus- ters (nD-c) [2] is the method that can provide carbon onions with uniform size in a large

  1. Electrochromic polyaniline/graphite oxide nanocomposites with endured electrochemical energy storage

    E-Print Network [OSTI]

    Guo, John Zhanhu

    films for electrochromic displays and electrochemical energy storage devices applications were. Introduction The development of sustainable and renewable energy storage resources with both high power densityElectrochromic polyaniline/graphite oxide nanocomposites with endured electrochemical energy

  2. The effects of marine microorganisms on the mechanical properties of graphite/epoxy composites 

    E-Print Network [OSTI]

    Puh, John Shui-Ming

    1997-01-01

    with two different lay-ups were conditioned in natural seawater and then tensile tested while simultaneously monitored for acoustic emission activity. Graphite/epoxy composite specimens were fabricated from prepreg tape and then conditioned for 4 and I I...

  3. Direct Observation of Optically Induced Transient Structures in Graphite Using Ultrafast Electron Crystallography

    E-Print Network [OSTI]

    initio density functional calculations, we trace the governing mechanism back to electronic structure changes in the electronic properties, direct de- termination of lattice structural dynamics from opticalDirect Observation of Optically Induced Transient Structures in Graphite Using Ultrafast Electron

  4. Claisen Rearrangement of Graphite Oxide: A Route to Covalently Functionalized Graphenes

    E-Print Network [OSTI]

    Collins, William R.

    On the GO: The basal plane allylic alcohol functionality of graphite oxide (GO) can be converted into N,N-dimethylamide groups through an Eschenmoser–Claisen sigmatropic rearrangement by using N,N-dimethylacetamide dimethyl ...

  5. Near E{sub F} Electronic Structure of Graphite from Photoemission and Inverse Photoemission Studies

    SciTech Connect (OSTI)

    Sekhar, B. R.; Kundu, R.; Mishra, P. [Institute of Physics, Sachivalaya Marg, Bhubaneswar 751005 (India); Maniraj, M.; Barman, S. R. [Surface Physics Laboratory, UGC-DAE Consortium for Scientific Research, Khandwa Road, Indore 452001, Madhya Pradesh (India)

    2011-10-20

    A comparative study of the electronic band structure of single crystal and highly oriented pyrolitic graphite is presented. We have used angle resolved photoelectron spectroscopy and angle resolved inverse photoelectron spectroscopy to map the occupied and unoccupied electronic states respectively.

  6. Thermodynamic and kinetic properties of the Li-graphite system from first-principles calculations

    E-Print Network [OSTI]

    Ceder, Gerbrand

    We present an ab initio study of the thermodynamics and kinetics of Li [subscript x]C[subscript 6], relevant for anode Li intercalation in rechargeable Li batteries. In graphite, the interlayer interactions are dominated ...

  7. Enhancing thermal conductivity of fluids with graphite nanoparticles and carbon nanotube

    DOE Patents [OSTI]

    Zhang, Zhiqiang (Lexington, KY); Lockwood, Frances E. (Georgetown, KY)

    2008-03-25

    A fluid media such as oil or water, and a selected effective amount of carbon nanomaterials necessary to enhance the thermal conductivity of the fluid. One of the preferred carbon nanomaterials is a high thermal conductivity graphite, exceeding that of the neat fluid to be dispersed therein in thermal conductivity, and ground, milled, or naturally prepared with mean particle size less than 500 nm, and preferably less than 200 nm, and most preferably less than 100 nm. The graphite is dispersed in the fluid by one or more of various methods, including ultrasonication, milling, and chemical dispersion. Carbon nanotubes with graphitic structure is another preferred source of carbon nanomaterial, although other carbon nanomaterials are acceptable. To confer long term stability, the use of one or more chemical dispersants is preferred. The thermal conductivity enhancement, compared to the fluid without carbon nanomaterial, is proportional to the amount of carbon nanomaterials (carbon nanotubes and/or graphite) added.

  8. Oxidation of hydrocarbons over ordered arrays of heteropolyacids and polyoxoanions on graphite

    DOE Patents [OSTI]

    Shaikh, S.N.; Ellis, P.E. Jr.; Lyons, J.E.

    1994-08-02

    Alkanes are catalytically oxidized using heteropolyacids or polyoxoanions deposited on a graphite surface. The heteropolyacids and polyoxoanions are framework-substituted with a different metal in place of a metal-oxygen unit.

  9. Oxidation of hydrocarbons over ordered arrays of heteropolyacids and polyoxoanions on graphite

    DOE Patents [OSTI]

    Shaikh, Shahid N. (Media, PA); Ellis, Jr., Paul E. (Downingtown, PA); Lyons, James E. (Wallingford, PA)

    1994-01-01

    Alkanes are catalytically oxidized using heteropolyacids (HPAs) or polyoxoanions (POAs) deposited on a graphite surface. The HPAs and POAs are framework-substituted with a different metal in place of a metal-oxygen unit.

  10. 2012 Environmental/Radiological Assistance Directory (ERAD) Presentati...

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

    Measurements in an Emergency: This is not a Drill; Stephen V. Musolino; Brookhaven National Laboratory November 2012; Brookhaven Graphite Research Reactor (BGRR) D&D...

  11. Reactor- Nuclear Science Center 

    E-Print Network [OSTI]

    Unknown

    2011-08-17

    A neutronic evaluation of two reactor benchmark problems was performed. The benchmark problems describe typical PWR uranium and plutonium (mixed oxide) fueled lattices. WIMSd4m, a neutron transport lattice code, was used to evaluate multigroup...

  12. P Reactor Grouting

    SciTech Connect (OSTI)

    None

    2010-01-01

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

  13. Hypothetical Reactor Accident Study

    E-Print Network [OSTI]

    POPULATIONS; IODINE 131; MELTDOWN; METEOROLOGY; NUCLEAR POWER PLANTS; P CODES; PWR TYPE REACTORS; RADIATION in a Typical BWR and in a typical PWR. Comparison with WASH-1400 by C F . Højerup 202 APPENDIX 3. Calculation

  14. Direct synthesis of sp-bonded carbon chains on graphite surface by femtosecond laser irradiation

    SciTech Connect (OSTI)

    Hu, A.; Rybachuk, M.; Lu, Q.-B.; Duley, W. W.

    2007-09-24

    Microscopic phase transformation from graphite to sp-bonded carbon chains (carbyne) and nanodiamond has been induced by femtosecond laser pulses on graphite surface. UV/surface enhanced Raman scattering spectra and x-ray photoelectron spectra displayed the local synthesis of carbyne in the melt zone while nanocrystalline diamond and trans-polyacetylene chains form in the edge area of gentle ablation. These results evidence possible direct 'writing' of variable chemical bonded carbons by femtosecond laser pulses for carbon-based applications.

  15. Mode II delamination fracture toughness of unidirectional graphite/epoxy composites 

    E-Print Network [OSTI]

    Corleto Mena, Carlos Roberto

    1986-01-01

    MODE II DELAMINATION FRACTURE TOUGHNESS OF UNIDIRECTIONAL GRAPHITE/EPOXY COMPOSITES A Thesis by CARLOS ROBERTO CORLETO MENA Submitted to the Graduate College of TEXAS ASM UNIVERSITY in partial fulfillment of the requirement for the degree... of MASTER OF SCIENCE August 1986 Major Subject: Mechanical Engineering MODE II DELAMINATION FRACTURE TOUGHNESS OF UNIDIRECTIONAL GRAPHITE/EPOXY COMPOSITES A Thesis by CARLOS ROBERTO CORLETO MENA Approved as to style and content by: Walter L. Bra...

  16. Residual thermal stresses in an unsymmetrical cross-ply graphite/epoxy laminate 

    E-Print Network [OSTI]

    Harper, Brian Douglas

    1980-01-01

    RESIDUAL THERMAL STRESSES IN AN UNSYMMETRICAL CROSS-PLY GRAPHITE/EPOXY LAMINATE A Thesis by BRIAN DOUGLAS HARPER Submitted to the Graduate College of Texas A&M University in parrial fulfillment of the requirement for the degree of MASTER... OF SCIENCE August 1980 Major Subject: Mechanical Engineering RESIDUAL THERMAL STRESSES IN AN UNSYMMETRICAL CROSS-PLY GRAPHITE/EPOXY LAMINATE A Thesis by BRIAN DOUGLAS HARPER Approved as to style and content by: r. Y. N itsman (Chair of Committee) Dr...

  17. Nonlinear viscoelastic characterization of AS-3502 graphite/epoxy composite material 

    E-Print Network [OSTI]

    Kerstetter, Michael Scott

    1980-01-01

    NONLINEAR VISCOELASTIC CHARACTERIZATION OF AS-3502 GRAPHITE/EPOXY COMPOSITE MATERIAL A Thesis MICHAEL SCOTT KERSTETTER Submitted to the Graduate College of Texas A8M University in partial fulfillment of the requirement for the degree... of MASTER OF SCIENCE December 1980 Major Subject: Aerospace Engineering NONLINEAR VISCOELASTIC CHARACTERIZATION OF AS-3502 GRAPHITE/EPOXY COMPOSITE MATERIAL A Thesis by MICHAEL SCOTT KERSTETTER Approved as to style and content by: Dr. Kenneth L...

  18. Molten metal reactors

    DOE Patents [OSTI]

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

    2013-11-05

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

  19. Three-dimensional EBSD characterization of thermo-mechanical fatigue crack morphology in compacted graphite iron

    SciTech Connect (OSTI)

    Pirgazi, Hadi, E-mail: Hadi.pirgazi@ugent.be [Department of Materials Science and Engineering, Ghent University, Technologiepark 903, 9052 Gent (Belgium); Ghodrat, Sepideh, E-mail: s.ghodrat@tudelft.nl [Department of Materials Science and Engineering, Delft University of Technology, Mekelweg 2, 2628 CD, Delft (Netherlands); Kestens, Leo A.I., E-mail: leo.kestens@ugent.be [Department of Materials Science and Engineering, Ghent University, Technologiepark 903, 9052 Gent (Belgium); Department of Materials Science and Engineering, Delft University of Technology, Mekelweg 2, 2628 CD, Delft (Netherlands)

    2014-04-01

    In cylinder heads made of compacted graphitic iron (CGI), heating and cooling cycles can lead to localized cracking due to thermo-mechanical fatigue (TMF). To meticulously characterize the complex crack path morphology of CGI under TMF condition, in relation to microstructural features and to find out how and by which mechanisms the cracks predominantly develop, three-dimensional electron back scattering diffraction (EBSD) was employed. Based on the precise quantitative microstructural analysis, it is found that graphite particles not only play a crucial role in the crack initiation, but also are of primary significance for crack propagation, i.e. crack growth is enhanced by the presence of graphite particles. Furthermore, the density of graphite particles on the fracture plane is more than double as high as in any other arbitrary plane of the structure. The obtained results did not indicate a particular crystallographic preference of fracture plane, i.e. the crystal plane parallel to the fracture plane was nearly of random orientation. - Highlights: • Crystallographic features of a thermo-mechanical fatigue (TMF) crack were studied. • Wide-field 3D EBSD is used to characterize the TMF crack morphology. • Data processing was applied on a large length scale of the order of millimeters. • Graphite density in the fracture plane is much higher than any other random plane. • It is revealed that crack growth is enhanced by the presence of graphite particles.

  20. F Reactor Inspection

    ScienceCinema (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-11-24

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

  1. F Reactor Inspection

    SciTech Connect (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29

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

  2. Risk Based Security Management at Research Reactors

    SciTech Connect (OSTI)

    Ek, David R.

    2015-09-01

    This presentation provides a background of what led to the international emphasis on nuclear security and describes how nuclear security is effectively implemented so as to preserve the societal benefits of nuclear and radioactive materials.

  3. Ames Laboratory Research Reactor Facility Ames, Iowa

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal Gas &SCE-SessionsSouth DakotaRobbins and Myers Co -VANaval ,, *' ; . Final Radiological

  4. Missouri University Research Reactor Site Fact Sheet

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal Gas &SCE-SessionsSouthReport for the Weldon Spring, Missouri,MSEReporty

  5. Decommissioning experience from the Experimental Breeder Reactor-II.

    SciTech Connect (OSTI)

    Henslee, S.P.; Rosenberg, K.E.

    2002-03-28

    Consistent with the intent of this International Atomic Energy Agency technical meeting, decommissioning operating experience and contributions to the preparation for the Coordinated Research Project from Experimental Breeder Reactor-II activities will be discussed. This paper will review aspects of the decommissioning activities of the Experimental Breeder Reactor-II, make recommendations for future decommissioning activities and reactor system designs and discuss relevant areas of potential research and development. The Experimental Breeder Reactor-II (EBR-II) was designed as a 62.5 MWt, metal fueled, pool reactor with a conventional 19 MWe power plant. The productive life of the EBR-II began with first operations in 1964. Demonstration of the fast reactor fuel cycle, serving as an irradiation facility, demonstration of fast reactor passive safety and lastly, was well on its way to close the fast breeder fuel cycle for the second time when the Integral Fast Reactor program was prematurely ended in October 1994 with the shutdown of the EBR-II. The shutdown of the EBR-II was dictated without an associated planning phase that would have provided a smooth transition to shutdown. Argonne National Laboratory and the U.S. Department of Energy arrived at a logical plan and sequence for closure activities. The decommissioning activities as described herein fall into in three distinct phases.

  6. Burning and graphitization of optically levitated nanodiamonds in vacuum

    E-Print Network [OSTI]

    Rahman, A T M A; Kim, M S; Bose, S; Morley, G W; Barker, P F

    2015-01-01

    A nitrogen-vacancy (NV$^-$) center in a nanodiamond, levitated in high vacuum, has recently been proposed as a probe for demonstrating mesoscopic center-of-mass superpositions \\cite{Scala2013, Zhang2013} and for testing quantum gravity \\cite{Albrecht2014}. Here, we study the behavior of optically levitated nanodiamonds containing NV$^-$ centers at sub-atmospheric pressures and show that while they burn in air, this can be prevented by replacing the air with nitrogen. However, in nitrogen the nanodiamonds graphitize below $\\approx 10$ mB. Exploiting the Brownian motion of a levitated nanodiamond, we extract its internal temperature ($T_i$) and find that it would be detrimental to the NV$^-$ center's spin coherence time \\cite{Toyli2012}. These values of $T_i$ make it clear that the diamond is not melting, contradicting a recent suggestion \\cite{Neukirch2015}. Additionally, using the measured damping rate of a levitated nanoparticle at a given pressure, we propose a new way of determining its size.

  7. Internal degrees of freedom and transport of benzene on graphite

    E-Print Network [OSTI]

    Astrid S. de Wijn

    2011-07-13

    In this paper, the chaotic internal degrees of freedom of a benzene molecule adsorbed on a graphite substrate, their interplay with thermal noise, and their effects on the diffusion and drift are investigated analytically by making use of the presence of two different time scales as well as by molecular-dynamics simulations. The effects of thermal noise are investigated, and it is found that noise does not significantly alter the dynamics of the internal degrees of freedom, yet affects the friction and diffusion of the center of mass. Qualitative and quantitative theoretical predictions for the friction and diffusion of the molecule on the substrate are made and are compared to molecular-dynamics simulations. Contributions to the friction and diffusion from the finite heat bath as well as the slow dynamics of the center of mass are formally identified. It is shown that the torsion in benzene, which dominates the nonlinear coupling, significantly affects the friction of the molecule on the surface. The results compare favorably with recent results from He/neutron spin echo experiments on this system. Based on the analytical and numerical results, some suggestions are made for experimental conditions under which the effects of internal degrees of freedom might be observable.

  8. Risk Management for Sodium Fast Reactors.

    SciTech Connect (OSTI)

    Denman, Matthew R; Groth, Katrina; Cardoni, Jeffrey N; Wheeler, Timothy A.

    2015-01-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

  9. Control Rod Malfunction at the NRAD Reactor

    SciTech Connect (OSTI)

    Thomas L. Maddock

    2010-05-01

    The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

  10. Corium Retention for High Power Reactors by An In-Vessel Core Catcher in Combination with External Reactor Vessel Cooling

    SciTech Connect (OSTI)

    Joy L. Rempe; D. L. Knudson; K. G. Condie; K. Y. Suh; F. -B. Cheung; S. -B. Kim

    2004-05-01

    If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel lower head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe Pressurized Water Reactor (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for in-vessel retention (IVR), resulted in the United States Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). Accordingly, IVR of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors. However, it is not clear that currently-proposed methods to achieve ERVC will provide sufficient heat removal for higher power reactors. A US–Korean International Nuclear Energy Research Initiative (INERI) project has been initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) will determine if IVR is feasible for reactors up to 1500 MWe. This paper summarizes results from the first year of this 3-year project.

  11. Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 39, No. 11, p. 11691181 (November 2002) Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M"

    E-Print Network [OSTI]

    Laughlin, Robert B.

    and accepted September 10, 2002) A metal fueled modular island core sodium cooled fast breeder reactor concept Design of a Modular Island Core Fast Breeder Reactor "RAPID-M" Mitsuru KAMBE Central Research Institute type reactors, performance, sodium cooled reactor, modular island core, inherent safety, in- tegrated

  12. Methanation assembly using multiple reactors

    DOE Patents [OSTI]

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

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

  13. Power Burst Facility (PBF) Reactor Reactor Decommissioning

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration wouldMass mapSpeedingProgram Guidelines This document outlinesPotentialReactor Decommissioning

  14. Effect of electrode density on cycle performance and irreversible capacity loss for natural graphite anode in lithium ion batteries

    SciTech Connect (OSTI)

    Shim, Joongpyo; Striebel, Kathryn A.

    2002-12-02

    The effect of electrode thickness and density for unpressed and pressed natural graphite electrodes were studied using electrochemical characterization. Pressing the graphite electrode decreases the reversible capacity and the irreversible capacity loss during formation. As electrode density increased, the capacity retention at high rate increased until 0.9g/cm{sup 3}, and then decreased. The cycle performances of the pressed graphite electrodes were more stable than the unpressed one. Pressing graphite electrode affected on its electrochemical characterization such as irreversible capacity loss, high rate cycling and cycle performance.

  15. Laminated exfoliated graphite composite-metal compositions for fuel cell flow field plate or bipolar plate applications

    DOE Patents [OSTI]

    Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

    2014-05-20

    An electrically conductive laminate composition for fuel cell flow field plate or bipolar plate applications. The laminate composition comprises at least a thin metal sheet having two opposed exterior surfaces and a first exfoliated graphite composite sheet bonded to the first of the two exterior surfaces of the metal sheet wherein the exfoliated graphite composite sheet comprises: (a) expanded or exfoliated graphite and (b) a binder or matrix material to bond the expanded graphite for forming a cohered sheet, wherein the binder or matrix material is between 3% and 60% by weight based on the total weight of the first exfoliated graphite composite sheet. Preferably, the first exfoliated graphite composite sheet further comprises particles of non-expandable graphite or carbon in the amount of between 3% and 60% by weight based on the total weight of the non-expandable particles and the expanded graphite. Further preferably, the laminate comprises a second exfoliated graphite composite sheet bonded to the second surface of the metal sheet to form a three-layer laminate. Surface flow channels and other desired geometric features can be built onto the exterior surfaces of the laminate to form a flow field plate or bipolar plate. The resulting laminate has an exceptionally high thickness-direction conductivity and excellent resistance to gas permeation.

  16. Core design and reactor physics of a breed and burn gas-cooled fast reactor

    E-Print Network [OSTI]

    Yarsky, Peter

    2005-01-01

    In order to fulfill the goals set forth by the Generation IV International Forum, the current NERI funded research has focused on the design of a Gas-cooled Fast Reactor (GFR) operating in a Breed and Burnm (B&B) fuel cycle ...

  17. Dynamic bed reactor

    DOE Patents [OSTI]

    Stormo, Keith E. (Moscow, ID)

    1996-07-02

    A dynamic bed reactor is disclosed in which a compressible open cell foam matrix is periodically compressed and expanded to move a liquid or fluid through the matrix. In preferred embodiments, the matrix contains an active material such as an enzyme, biological cell, chelating agent, oligonucleotide, adsorbent or other material that acts upon the liquid or fluid passing through the matrix. The active material may be physically immobilized in the matrix, or attached by covalent or ionic bonds. Microbeads, substantially all of which have diameters less than 50 microns, can be used to immobilize the active material in the matrix and further improve reactor efficiency. A particularly preferred matrix is made of open cell polyurethane foam, which adsorbs pollutants such as polychlorophenol or o-nitrophenol. The reactors of the present invention allow unidirectional non-laminar flow through the matrix, and promote intimate exposure of liquid reactants to active agents such as microorganisms immobilized in the matrix.

  18. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

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

    1987-01-01

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

  19. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  20. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    1986-01-01

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  1. Reactor for exothermic reactions

    DOE Patents [OSTI]

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

    1993-01-01

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

  2. Reactor for exothermic reactions

    DOE Patents [OSTI]

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

    1993-03-02

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

  3. Fast quench reactor method

    DOE Patents [OSTI]

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

    1999-08-10

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

  4. Perspectives on reactor safety

    SciTech Connect (OSTI)

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

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  5. Nuclear reactor apparatus

    DOE Patents [OSTI]

    Wade, Elman E. (Ruffs Dale, PA)

    1978-01-01

    A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

  6. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID); Berry, Ray A. (Idaho Falls, ID)

    1999-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

  7. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, Daniel L. (Princeton, NJ)

    1988-01-01

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

  8. Innovative design of uranium startup fast reactors

    E-Print Network [OSTI]

    Fei, Tingzhou

    2012-01-01

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

  9. NanoSIMS, TEM, and XANES studies of a unique presolar supernova graphite grain

    SciTech Connect (OSTI)

    Groopman, Evan; Bernatowicz, Thomas; Zinner, Ernst; Nittler, Larry R.

    2014-07-20

    We report on isotopic and microstructural investigations of a unique presolar supernova (SN) graphite grain, referred to as G6, isolated from the Orgueil CI chondrite. G6 contains complex heterogeneities in its isotopic composition and in its microstructure. Nano-scale secondary ion mass spectrometer isotope images of ultramicrotome sections reveal heterogeneities in its C, N, and O isotopic compositions, including anomalous shell-like structures. Transmission electron microscope studies reveal a nanocrystalline core surrounded by a turbostratic graphite mantle, the first reported nanocrystalline core from a low-density SN graphite grain. Electron diffraction analysis shows that the nanocrystalline core consists of randomly oriented 2-4 nm graphene particles, similar to those in cores of high-density (HD) presolar graphite grains from asymptotic giant branch stars. G6's core also exhibits evidence for planar stacking of these graphene nano-sheets with a domain size up to 4.5 nm, which was unobserved in the nanocrystalline cores of HD graphite grains. We also report on X-ray absorption near-edge structure measurements of G6. The complex isotopic- and micro-structure of G6 provides evidence for mixing and/or granular transport in SN ejecta.

  10. F Reactor Area Cleanup Complete

    Broader source: Energy.gov [DOE]

    RICHLAND, Wash. – U.S. Department of Energy (DOE) contractors have cleaned up the F Reactor Area, the first reactor area at the Hanford Site in southeastern Washington state to be fully remediated.

  11. Reactor operation environmental information document

    SciTech Connect (OSTI)

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

    1989-12-01

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

  12. Fuel Cycle Research and Development Presentation Title

    Energy Savers [EERE]

    Fuels Technical Lead Advanced Fuels Campaign Advanced LWR Fuels Pathway Lead Light Water Reactor Sustainability Program August 2013 Outline Overview of DOE SiC research ...

  13. International Nuclear Energy Research Initiative, Fiscal Year...

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

    Area: Reactor Concepts RD&D Project Start Date: January 2011 Project End Date: December 2013 38 | International Nuclear Energy Research Initiative (I-NERI) Fiscal Year 2011...

  14. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

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

  15. Instrumentation to Enhance Advanced Test Reactor Irradiations

    SciTech Connect (OSTI)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  16. (Irradiation embrittlement of reactor pressure vessels)

    SciTech Connect (OSTI)

    Corwin, W.R.

    1990-09-24

    The traveler served as a member of the two-man US Nuclear Regulatory Commission sponsored team who visited the Prometey Complex in Leningrad to assess the potential for expanded cooperative research concerning integrity of the primary pressure boundary in commercial light-water reactors. The emphasis was on irradiation embrittlement, structural analysis, and fracture mechanics research for reactor pressure vessels. At the irradiation seminar in Cologne, presentations were made by German, French, Finnish, Russian, and US delegations concerning many aspects of irradiation of pressure vessel steels. The traveler made presentations on mechanisms of irradiation embrittlement and on important aspects of the Heavy-Section Steel Irradiation Program results of irradiated fracture mechanics tests.

  17. A Wide Range Neutron Detector for Space Nuclear Reactor Applications

    SciTech Connect (OSTI)

    Nassif, Eduardo; Sismonda, Miguel; Matatagui, Emilio; Pretorius, Stephan

    2007-01-30

    We propose here a versatile and innovative solution for monitoring and controlling a space-based nuclear reactor that is based on technology already proved in ground based reactors. A Wide Range Neutron Detector (WRND) allows for a reduction in the complexity of space based nuclear instrumentation and control systems. A ground model, predecessor of the proposed system, has been installed and is operating at the OPAL (Open Pool Advanced Light Water Research Reactor) in Australia, providing long term functional data. A space compatible Engineering Qualification Model of the WRND has been developed, manufactured and verified satisfactorily by analysis, and is currently under environmental testing.

  18. Fossil fuel furnace reactor

    DOE Patents [OSTI]

    Parkinson, William J. (Los Alamos, NM)

    1987-01-01

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

  19. Nuclear Reactors and Technology

    SciTech Connect (OSTI)

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

    1992-01-01

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

  20. Thermal Reactor Safety

    SciTech Connect (OSTI)

    Not Available

    1980-06-01

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

  1. Cermet fuel reactors

    SciTech Connect (OSTI)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  2. Nuclear reactor building

    DOE Patents [OSTI]

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

    1994-01-01

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

  3. Nuclear reactor building

    DOE Patents [OSTI]

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

    1994-04-05

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

  4. NETL - Chemical Looping Reactor

    SciTech Connect (OSTI)

    2013-07-24

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

  5. Reactor Monitoring with Neutrinos

    E-Print Network [OSTI]

    M. Cribier

    2007-04-06

    The fundamental knowledge on neutrinos acquired in the recent years open the possibility of applied neutrino physics. Among it the automatic and non intrusive monitoring of nuclear reactor by its antineutrino signal could be very valuable to IAEA in charge of the control of nuclear power plants. Several efforts worldwide have already started.

  6. Reactor Monitoring with Neutrinos

    E-Print Network [OSTI]

    Cribier, Michel

    2011-01-01

    The fundamental knowledge on neutrinos acquired in the recent years open the possibility of applied neutrino physics. Among it the automatic and non intrusive monitoring of nuclear reactor by its antineutrino signal could be very valuable to IAEA in charge of the control of nuclear power plants. Several efforts worldwide have already started.

  7. Reactor component automatic grapple

    DOE Patents [OSTI]

    Greenaway, Paul R. (Bethel Park, PA)

    1982-01-01

    A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment.

  8. NETL - Chemical Looping Reactor

    ScienceCinema (OSTI)

    None

    2014-06-26

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

  9. A facile approach to prepare graphene via solvothermal reduction of graphite oxide

    SciTech Connect (OSTI)

    Yuan, Bihe [State Key Laboratory of Fire Science, University of Science and Technology of China, Hefei 230026 (China); Suzhou Key Laboratory of Urban Public Safety, Suzhou Institute for Advanced Study, University of Science and Technology of China, Suzhou 215123 (China); Bao, Chenlu [State Key Laboratory of Fire Science, University of Science and Technology of China, Hefei 230026 (China); Qian, Xiaodong; Wen, Panyue [State Key Laboratory of Fire Science, University of Science and Technology of China, Hefei 230026 (China); Suzhou Key Laboratory of Urban Public Safety, Suzhou Institute for Advanced Study, University of Science and Technology of China, Suzhou 215123 (China); Xing, Weiyi; Song, Lei [State Key Laboratory of Fire Science, University of Science and Technology of China, Hefei 230026 (China); Hu, Yuan, E-mail: yuanhu@ustc.edu.cn [State Key Laboratory of Fire Science, University of Science and Technology of China, Hefei 230026 (China); Suzhou Key Laboratory of Urban Public Safety, Suzhou Institute for Advanced Study, University of Science and Technology of China, Suzhou 215123 (China)

    2014-07-01

    Highlights: • Graphene was prepared via a novel and facile solvothermal reduction method for graphite oxide. • Most of the oxygen functional groups of graphite oxide were removed. • The reduced graphene oxide obtained was featured with bilayer nanosheets. - Abstract: In this work, a facile reduction strategy is reported for the fabrication of graphene. Graphite oxide (GO) is reduced via a novel solvothermal reaction in a mixed solution of acetone and sodium hypochlorite (NaClO). The structure, surface chemistry, morphology and thermal stability of the as-prepared reduced graphene oxide (RGO) are characterized by X-ray diffraction (XRD), Fourier transform infrared spectroscopy (FTIR), Raman spectroscopy, transmission electron microscopy (TEM), atomic force microscopy (AFM), thermogravimetric analysis (TGA) and X-ray photoelectron spectroscopy (XPS). The results indicate that most of the oxygenated groups in GO are effectively removed in this solvothermal reaction. The novel reduction method provides a simple, cost-effective and efficient strategy for the fabrication of graphene.

  10. Phase transformations of nano-sized cubic boron nitride to white graphene and white graphite

    SciTech Connect (OSTI)

    Dang, Hongli; Liu, Yingdi; Xue, Wenhua; Anderson, Ryan S.; Sewell, Cody R. [Department of Physics and Engineering Physics, The University of Tulsa, Tulsa, Oklahoma 74104 (United States); Xue, Sha; Crunkleton, Daniel W. [Department of Chemical Engineering, The University of Tulsa, Tulsa, Oklahoma 74104 (United States); Institute of Alternate Energy, The University of Tulsa, Tulsa, Oklahoma 74104 (United States); Shen, Yaogen [Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Kowloon (Hong Kong); Wang, Sanwu, E-mail: sanwu-wang@utulsa.edu [Department of Physics and Engineering Physics, The University of Tulsa, Tulsa, Oklahoma 74104 (United States); Institute of Alternate Energy, The University of Tulsa, Tulsa, Oklahoma 74104 (United States)

    2014-03-03

    We report quantum-mechanical investigations that predict the formation of white graphene and nano-sized white graphite from the first-order phase transformations of nano-sized boron nitride thin-films. The phase transformations from the nano-sized diamond-like structure, when the thickness d?>?1.4?nm, to the energetically more stable nano-sized white graphite involve low activation energies of less than 1.0?eV. On the other hand, the diamond-like structure transforms spontaneously to white graphite when d???1.4?nm. In particular, the two-dimensional structure with single-layer boron nitride, the so-called white graphene, could be formed as a result of such transformation.

  11. The semi-empirical tight-binding model for carbon allotropes “between diamond and graphite

    SciTech Connect (OSTI)

    Lytovchenko, V.; Kurchak, A.; Strikha, M.

    2014-06-28

    The new carbon allotropes “between diamond and graphite” have come under intensive examination during the last decade due to their numerous technical applications. The modification of energy gap in thin films of these allotropes was studied experimentally using optical methods. The proposed simple model of carbon clusters with variable lengths of chemical bonds allows us to imitate the transfer from diamond and diamond-like to graphite-like structures, as well as the corresponding modification of hybridization sp{sup 3}/sp{sup 2} for diamond-like and sp{sub z} for graphite-like phases. This enables us to estimate various allotropes parameters, like the gap E{sub g}, energies of valence E{sub v}, and conduction E{sub c} band edges, and the value of electronic affinity, i.e., optical work function X, which are all of practical importance. The obtained estimations correspond to the experimental data.

  12. Synthesis and characterization of nanocrystalline graphite from coconut shell with heating process

    SciTech Connect (OSTI)

    Wachid, Frischa M., E-mail: frischamw@yahoo.com, E-mail: adhiyudhaperkasa@yahoo.com, E-mail: afandisar@yahoo.com, E-mail: nurulrosyidah92@gmail.com, E-mail: darminto@physics.its.ac.id; Perkasa, Adhi Y., E-mail: frischamw@yahoo.com, E-mail: adhiyudhaperkasa@yahoo.com, E-mail: afandisar@yahoo.com, E-mail: nurulrosyidah92@gmail.com, E-mail: darminto@physics.its.ac.id; Prasetya, Fandi A., E-mail: frischamw@yahoo.com, E-mail: adhiyudhaperkasa@yahoo.com, E-mail: afandisar@yahoo.com, E-mail: nurulrosyidah92@gmail.com, E-mail: darminto@physics.its.ac.id; Rosyidah, Nurul, E-mail: frischamw@yahoo.com, E-mail: adhiyudhaperkasa@yahoo.com, E-mail: afandisar@yahoo.com, E-mail: nurulrosyidah92@gmail.com, E-mail: darminto@physics.its.ac.id; Darminto, E-mail: frischamw@yahoo.com, E-mail: adhiyudhaperkasa@yahoo.com, E-mail: afandisar@yahoo.com, E-mail: nurulrosyidah92@gmail.com, E-mail: darminto@physics.its.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Sepuluh Nopember, Campus ITS Sukolilo, Surabaya 60111 (Indonesia)

    2014-02-24

    Graphite were synthesized and characterized by heating process of coconut shell with varying temperature (400, 800 and 1000°C) and holding time (3 and 5 hours). After heating process, the samples were characterized by X-ray diffraction (XRD) and analyzed by X'pert HighScore Plus Software, Scanning Electron Microcope-Energy Dispersive X-Ray (SEM-EDX) and Transmission Electron Microscope-Energy Dispersive X-Ray (TEM-EDX). Graphite and londsdaelite phase were analyzed by XRD. According to EDX analysis, the sample was heated in 1000°C got the highest content of carbon. The amorphous carbon and nanocrystalline graphite were observed by SEM-EDX and TEM-EDX.

  13. Phonon mean free path of graphite along the c-axis

    SciTech Connect (OSTI)

    Wei, Zhiyong; Yang, Juekuan; Chen, Weiyu; Bi, Kedong; Chen, Yunfei, E-mail: yunfeichen@seu.edu.cn [Jiangsu Key Laboratory for Design and Manufacture of Micro/Nano Biomedical Instruments and School of Mechanical Engineering, Southeast University, Nanjing 210096 (China); Li, Deyu, E-mail: deyu.li@vanderbilt.edu [Department of Mechanical Engineering, Vanderbilt University, Nashville, Tennessee 37235-1592 (United States)

    2014-02-24

    Phonon transport in the c-axis direction of graphite thin films has been studied using non-equilibrium molecular dynamics (MD) simulation. The simulation results show that the c-axis thermal conductivities for films of thickness ranging from 20 to 500 atomic layers are significantly lower than the bulk value. Based on the MD data, a method is developed to construct the c-axis thermal conductivity as an accumulation function of phonon mean free path (MFP), from which we show that phonons with MFPs from 2 to 2000?nm contribute ?80% of the graphite c-axis thermal conductivity at room temperature, and phonons with MFPs larger than 100?nm contribute over 40% to the c-axis thermal conductivity. These findings indicate that the commonly believed value of just a few nanometers from the simple kinetic theory drastically underestimates the c-axis phonon MFP of graphite.

  14. Coated graphite articles useful in metallurgical processes and method for making same

    DOE Patents [OSTI]

    Holcombe, Cressie E. (Knoxville, TN); Bird, Eugene L. (Knoxville, TN)

    1995-01-01

    Graphite articles including crucibles and molds used in metallurgical processes involving the melting and the handling of molten metals and alloys that are reactive with carbon when in a molten state and at process temperatures up to about 2000.degree. C. are provided with a multiple-layer coating for inhibiting carbon diffusion from the graphite into the molten metal or alloys. The coating is provided by a first coating increment of a carbide-forming metal on selected surfaces of the graphite, a second coating increment of a carbide forming metal and a refractory metal oxide, and a third coating increment of a refractory metal oxide. The second coating increment provides thermal shock absorbing characteristics to prevent delamination of the coating during temperature cycling. A wash coat of unstabilized zirconia or titanium nitride can be applied onto the third coating increment to facilitate release of melts from the coating.

  15. Colloidal graphite/graphene nanostructures using collagen showing enhanced thermal conductivity

    E-Print Network [OSTI]

    Bhattacharya, Soumya; Das, Sarit K; Ganguly, Ranjan; Webster, Thomas; Nayar, Suprabha

    2015-01-01

    Time kinetics of interaction of natural graphite (GR) to colloidal graphene (G) collagen (C) nanocomposites was studied at ambient conditions, and observed that just one day at ambient conditions is enough to form colloidal graphene directly from graphite using the protein collagen. Neither controlled temperature and pressure ambiance nor sonication was needed for the same; thereby rendering the process biomimetic. Detailed spectroscopy, X ray diffraction, electron microscopy as well as fluorescence and luminescence assisted characterization of the colloidal dispersions on day one and day seven reveals graphene and collagen interaction and subsequent rearrangement to form an open structure. Detailed confocal microscopy, in the liquid state, reveals the initial attack at the zigzag edges of GR, the enhancement of auto fluorescence and finally the opening up of graphitic stacks of GR to form near transparent G. Atomic Force Microscopy studies prove the existence of both collagen and graphene and the disruption ...

  16. Graphite electrode DC arc furnace. Innovative technology summary report

    SciTech Connect (OSTI)

    1999-05-01

    The Graphite Electrode DC Arc Furnace (DC Arc) is a high-temperature thermal process, which has been adapted from a commercial technology, for the treatment of mixed waste. A DC Arc Furnace heats waste to a temperature such that the waste is converted into a molten form that cools into a stable glassy and/or crystalline waste form. Hazardous organics are destroyed through combustion or pyrolysis during the process and the majority of the hazardous metals and radioactive components are incorporated in the molten phase. The DC Arc Furnace chamber temperature is approximately 593--704 C and melt temperatures are as high as 1,500 C. The DC Arc system has an air pollution control system (APCS) to remove particulate and volatiles from the offgas. The advantage of the DC Arc is that it is a single, high-temperature thermal process that minimizes the need for multiple treatment systems and for extensive sorting/segregating of large volumes of waste. The DC Arc has the potential to treat a wide range of wastes, minimize the need for sorting, reduce the final waste volumes, produce a leach resistant waste form, and destroy organic contaminants. Although the DC arc plasma furnace exhibits great promise for treating the types of mixed waste that are commonly present at many DOE sites, several data and technology deficiencies were identified by the Mixed Waste Focus Area (MWFA) regarding this thermal waste processing technique. The technology deficiencies that have been addressed by the current studies include: establishing the partitioning behavior of radionuclides, surrogates, and hazardous metals among the product streams (metal, slag, and offgas) as a function of operating parameters, including melt temperature, plenum atmosphere, organic loading, chloride concentration, and particle size; demonstrating the efficacy of waste product removal systems for slag and metal phases; determining component durability through test runs of extended duration, evaluating the effect of feed composition variations on process operating conditions and slag product performance; and collecting mass balance and operating data to support equipment and instrument design.

  17. A wall-crawling robot for reactor vessel inspection in advanced reactors

    SciTech Connect (OSTI)

    Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

    1994-06-01

    A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected.

  18. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

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

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  19. Simplified Electrochemical and Thermal Model of LiFePO4-Graphite Li-Ion Batteries for Fast Charge Applications

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Simplified Electrochemical and Thermal Model of LiFePO4- Graphite Li-Ion Batteries for Fast Charge, a simplified electrochemical and thermal model of LiFePO4-graphite based Li-ion batteries is developed for battery management system (BMS) applications and comprehensive aging investigations. Based on a modified

  20. A Simplified Electrochemical and Thermal Aging Model of LiFePO4-Graphite Li-ion Batteries

    E-Print Network [OSTI]

    1 A Simplified Electrochemical and Thermal Aging Model of LiFePO4-Graphite Li-ion Batteries: Power of a commercial LiFePO4-graphite Li-ion battery. Compared to the isothermal reference, the mechanism of porosity;2 Due to their high power and energy densities, Li-ion technologies are the leading battery systems