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1

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

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

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

2

Environmental Management Brookhaven Graphite Research Reactor  

E-Print Network [OSTI]

-out report · Transition to long-term surveillance and maintenance · Office of Environmental ManagementEnvironmental Management Brookhaven Graphite Research Reactor (BGRR) Project Completion John Sattler Federal Project Director Office of Environmental Management U.S. Department of Energy BNL

Homes, Christopher C.

3

Brookhaven Lab Completes Decommissioning of Graphite Research Reactor:  

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

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

4

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

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

Why Was the BGRR Decommissioned? Why Was the BGRR Decommissioned? BGRR The Brookhaven Graphite Research Reactor (BGRR) at Brookhaven National Laboratory (BNL) was decommissioned to ensure the complex is in a safe and stable condition and to reduce sources of groundwater contamination. The BGRR contained over 8,000 Curies of radioactive contaminants from past operations consisting of primarily nuclear activation products such as hydrogen-3 (tritium) and carbon-14 and fission products cesium-137 and strontium-90. The nature and extent of contamination varied by location depending on historic uses of the systems and components and releases, however, the majority of the contamination (over 99 percent) was bound within the graphite pile and biological shield. Radioactive contamination was identified in the fuel handling system deep

5

Brookhaven Graphite Research Reactor Workshop | Department of Energy  

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

Services » Site & Facility Restoration » Deactivation & Services » Site & Facility Restoration » Deactivation & Decommissioning (D&D) » D&D Workshops » Brookhaven Graphite Research Reactor Workshop Brookhaven Graphite Research Reactor Workshop 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 operating in August 1950. In the next 18 years, an estimated 25,000 scientific experiments were carried out at the BGRR using neutrons produced in the facility's 700-ton graphite core, made up of more than 60,000 individual graphite blocks. The BGRR was placed on standby in 1968 and then permanently shut down as the next-generation reactor, the High Flux Beam Reactor (HFBR), was

6

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

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

- Cleanup Actions - Cleanup Actions Since the Brookhaven Graphite Research Reactor (BGRR) was shut down in 1968, many actions have been taken as part of the complex decommissioning. The actions undertaken throughout the BGRR complex ensure that the structures that remain are in a safe and stable condition and prepared it for long-term surveillance and maintenance. Regulatory Requirements The decommissioning of the BGRR was conducted under the federal Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). In 1992, an Interagency Agreement (PDF) among the DOE, the U.S. Environmental Protection Agency (EPA) and the New York State Department of Environmental Conservation (NYSDEC) became effective. The IAG provided the overall framework for conducting environmental restoration activities at

7

Photo of the Week: The Brookhaven Graphite Research Reactor ...  

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

which is a research collaboration between institutions from the U.S., Brazil, U.K., Germany, Spain and Switzerland. The survey aims to explore the dynamics of the universe's...

8

Graphite Reactor | ornl.gov  

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

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

9

X-10 Graphite Reactor | Department of Energy  

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

X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor When President Roosevelt in December 1942 authorized the Manhattan Project, the Oak Ridge site in eastern Tennessee had already been obtained and plans laid for an air-cooled experimental pile, a pilot chemical separation plant, and support facilities. The X-10 Graphite Reactor, designed and built in ten months, went into operation on November 4, 1943. The X-10 used neutrons emitted in the fission of uranium-235 to convert uranium-238 into a new element, plutonium-239. The reactor consists of a huge block of graphite, measuring 24 feet on each side, surrounded by several feet of high-density concrete as a radiation shield. The block is pierced by 1,248 horizontal diamond-shaped channels in

10

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)

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).

Evan Harpenau

2011-07-15T23:59:59.000Z

11

History of Research Reactors at Brookhaven  

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

History of Research Reactors at Brookhaven History of Research Reactors at Brookhaven Brookhaven National Laboratory has three nuclear reactors on its site that were used for scientific research. The reactors are all shut down, and the Laboratory is addressing environmental issues associated with their operations. photo of BGRR Brookhaven Graphite Research Reactor - Beginning operations in 1950, the graphite reactor was used for research in medicine, biology, chemistry, physics and nuclear engineering. One of the most significant achievements at this facility was the development of technetium-99m, a radiopharmaceutical widely used to image almost any organ in the body. The graphite reactor was shut down in 1969. Parts of it have been decommissioned, with the remainder to be addressed by 2011. More history

12

Reactor Safety Research Programs  

SciTech Connect (OSTI)

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

Edler, S. K.

1981-07-01T23:59:59.000Z

13

BNL | Our History: Reactors as Research Tools  

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

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

14

Research reactors - an overview  

SciTech Connect (OSTI)

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

West, C.D.

1997-03-01T23:59:59.000Z

15

Microsoft Word - 911136_0 SSC-4b Reactor Graphite Test Plan_rel...  

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

Services for the Next Generation Nuclear Plant (NGNP) with Hydrogen Production Test Plan for Reactor Graphite Elements Prepared by General Atomics for the Battelle Energy...

16

Advanced Nuclear Research Reactor  

SciTech Connect (OSTI)

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

Lolich, J.V.

2004-10-06T23:59:59.000Z

17

Nuclear Research Reactors  

Science Journals Connector (OSTI)

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

T. E. ALLIBONE

1963-07-20T23:59:59.000Z

18

Canadian university research reactors  

SciTech Connect (OSTI)

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

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

1989-11-01T23:59:59.000Z

19

Chicago Pile reactors create enduring research legacy - Argonne's  

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

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

20

Inelastic Thermal Neutron Scattering Cross Sections for Reactor-grade Graphite  

Science Journals Connector (OSTI)

Abstract Current calculations of the inelastic thermal neutron scattering cross sections of graphite are based on representing the material using ideal single crystal models. However, the density of reactor-grade graphite is usually in the range of 1.5 g/cm3 to approximately 1.8 g/cm3, while ideal graphite is characterized by a density of nearly 2.25 g/cm3. This difference in density is manifested as a significant fraction of porosity in the structure of reactor-grade graphite. To account for the porosity effect on the cross sections, classical molecular dynamics (MD) techniques were employed to simulate graphite structures with porosity concentrations of 10% and 30%, which are taken to be representative of reactor-grade graphite. The phonon density of states for the porous systems were generated as the power spectrum of the MD velocity autocorrelation functions. The analysis revealed that for porous graphite the phonon density of states exhibit a rise in the lower frequency region that is relevant to neutron thermalization. Using the generated phonon density of states, the inelastic thermal neutron scattering cross sections were calculated using the NJOY code system. While marked discrepancies exist between measurements and calculations based on ideal graphite models, favorable agreement is found between the calculations based on the porous graphite models and measured data.

A.I. Hawari; V.H. Gillete

2014-01-01T23:59:59.000Z

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


21

Diversion assumptions for high-powered research reactors  

SciTech Connect (OSTI)

This study deals with diversion assumptions for high-powered research reactors -- specifically, MTR fuel; pool- or tank-type research reactors with light-water moderator; and water, beryllium, or graphite reflectors, and which have a power level of 25 MW(t) or more. The objective is to provide assistance to the IAEA in documentation of criteria and inspection observables related to undeclared plutonium production in the reactors described above, including: criteria for undeclared plutonium production, necessary design information for implementation of these criteria, verification guidelines including neutron physics and heat transfer, and safeguards measures to facilitate the detection of undeclared plutonium production at large research reactors.

Binford, F.T.

1984-01-01T23:59:59.000Z

22

Treatment of Irradiated Graphite from French Bugey Reactor - 13424  

SciTech Connect (OSTI)

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)

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)

2013-07-01T23:59:59.000Z

23

Treatment of Irradiated Graphite to meet Acceptance Criteria for Waste Disposal: A New IAEA Collaborative Research Program - 12443  

SciTech Connect (OSTI)

World-wide, more than 250,000 tonnes of irradiated graphite have arisen through commercial nuclear-power operations and from military production reactors. Whilst most nations responsible for the generation of this material have in mind repository disposal alongside other radwaste, the lack of progress in this regard has led in some cases to difficulties where, for example, the site of an existing graphite-moderated reactor is required for re-utilisation. In any case, graphite as a radwaste stream has unique chemical and physical properties which may lend itself to more radical and innovative treatment and disposal options, including the recovery of useful isotopes and also recycling within the nuclear industry. Such aspects are important in making the case for future graphite-moderated reactor options (for example, High-Temperature Reactors planned for simultaneous power production and high-grade heat sources for such applications as hydrogen production for road fuel). A number of initiatives have taken place since the mid 1990s aimed at exploring such alternative strategies and, more recently, improving technology offers new options at all stages of the dismantling and disposal process. A new IAEA Collaborative Research Program aims to build upon the work already done and the knowledge achieved, in order to identify the risks and uncertainties associated with alternative options for graphite disposal, along with cost comparisons, thus enabling individual Member States to have the best-available information at their disposal to configure their own programs. (authors)

Wickham, A.J. [Nuclear Technology Consultancy, PO Box 50, Builth Wells, Powys LD2 3XA (United Kingdom); Drace, Z. [Waste Technology Section, Division of Nuclear Fuel Cycle and Waste Technology, International Atomic Energy Agency, Wagramerstrasse 5, PO Box 100, A-1400, Vienna (Austria)

2012-07-01T23:59:59.000Z

24

Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors  

SciTech Connect (OSTI)

A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on the ir graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment.

Grover, S.B. (INEEL); Metcalfe, M.P. (BNFL, United Kingdom)

2002-03-14T23:59:59.000Z

25

Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors  

SciTech Connect (OSTI)

A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on their graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment.

Grover, Stanley Blaine; Metcalfe, M. P.

2002-04-01T23:59:59.000Z

26

Advanced Reactor Research and Development Funding Opportunity...  

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

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

27

2012 Annual Report Research Reactor Infrastructure Program  

SciTech Connect (OSTI)

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

Douglas Morrell

2012-11-01T23:59:59.000Z

28

Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072  

SciTech Connect (OSTI)

About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

Poncet, Bernard [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)] [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)

2013-07-01T23:59:59.000Z

29

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

SciTech Connect (OSTI)

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.

Romano, T.

1997-09-29T23:59:59.000Z

30

Oxidation Resistant Graphite Studies  

SciTech Connect (OSTI)

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 740C 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.

W. Windes; R. Smith

2014-07-01T23:59:59.000Z

31

Diversion assumptions for high-powered research reactors. ISPO C-50 Phase 1  

SciTech Connect (OSTI)

This study deals with diversion assumptions for high-powered research reactors -- specifically, MTR fuel; pool- or tank-type research reactors with light-water moderator; and water, beryllium, or graphite reflectors, and which have a power level of 25 MW(t) or more. The objective is to provide assistance to the IAEA in documentation of criteria and inspection observables related to undeclared plutonium production in the reactors described above, including: criteria for undeclared plutonium production, necessary design information for implementation of these criteria, verification guidelines including neutron physics and heat transfer, and safeguards measures to facilitate the detection of undeclared plutonium production at large research reactors.

Binford, F.T.

1984-01-01T23:59:59.000Z

32

Project of Rotating Carbon High-Power Neutron Target. Research of Graphite Properties for Production of High Intensity Neutron Source  

E-Print Network [OSTI]

Project of Rotating Carbon High-Power Neutron Target. Research of Graphite Properties for Production of High Intensity Neutron Source

Gubin, K V; Bak, P A; Kot, N K; Logatchev, P V

2001-01-01T23:59:59.000Z

33

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

SciTech Connect (OSTI)

The United States Department of Energys 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.

Blaine Grover

2012-10-01T23:59:59.000Z

34

Research Program of a Super Fast Reactor  

SciTech Connect (OSTI)

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki [Nuclear Professional School / Department of Nuclear Engineering and Management, The University of Tokyo, Tokaimura, Naka-gun, Ibaraki, 319-1188 (Japan); Mori, Hideo [Department of Mechanical Engineering, Kyushu University (Japan); Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki [Japan Atomic Energy Agency (Japan); GOTO, Shoji [Tokyo Electric Power Company (Japan)

2006-07-01T23:59:59.000Z

35

Removal of {sup 14}C from Irradiated Graphite for Graphite Recycle and Waste Volume Reduction  

SciTech Connect (OSTI)

The aim of the research presented here was to identify the checmical from of {sup 14}C inirradiated 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. Approimately 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 ({sup 14}C), with a half-life of 5730 years.

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

2014-06-10T23:59:59.000Z

36

Organic free radicals and micropores in solid graphitic carbonaceous matter at the Oklo natural fission reactors, Gabon  

SciTech Connect (OSTI)

The presence, concentration, and distribution of organic free radicals as well as their association with specific surface areas and microporosities help characterize the evolution and behavior of the Oklo carbonaceous matter. Such information is necessary in order to evaluate uranium mineralization, liquid bitumen solidification, and radio nuclide containment at Oklo. In the Oklo ore deposits and natural fission reactors carbonaceous matter is often referred to as solid graphitic bitumen. The carbonaceous parts of the natural reactors may contain as much as 65.9% organic C by weight in heterogeneous distribution within the clay-rich matrix. The solid carbonaceous matter immobilized small uraninite crystals and some fission products enclosed in this uraninite and thereby facilitated radio nuclide containment in the reactors. Hence, the Oklo natural fission reactors are currently the subjects of detailed studies because they may be useful analogues to support performance assessment of radio nuclide containment at anthropogenic radioactive waste repository sites. Seven carbonaceous matter rich samples from the 1968 {+-} 50 Ma old natural fission reactors and the associated Oklo uranium ore deposit were studied by electron spin resonance (ESR) spectroscopy and by measurements of specific surface areas (BET method). Humic acid, fulvic acid, and fully crystalline graphite standards were also examined by ESR spectroscopy for comparison with the Oklo solid graphitic bitumens. With one exception, the ancient Oklo bitumens have higher organic free radical concentrations than the modem humic and fulvic acid samples. The presence of carbon free radicals in the graphite standard could not be determined due to the conductivity of this material. 72 refs., 7 figs., 1 tab.

Rigali, M.J.; Nagy, B. [Univ. of Arizona, Tucson, AZ (United States)] [Univ. of Arizona, Tucson, AZ (United States)

1997-01-01T23:59:59.000Z

37

University Research Reactor Task Force to the Nuclear Energy Research  

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

University Research Reactor Task Force to the Nuclear Energy University Research Reactor Task Force to the Nuclear Energy Research Advisory Committee University Research Reactor Task Force to the Nuclear Energy Research Advisory Committee In mid-February, 2001 The University Research Reactor (URR) Task Force (TF), a sub-group of the Department of Energy (DOE) Nuclear Energy Research Advisory Committee (NERAC), was asked to: * Analyze information collected by DOE, the NERAC "Blue Ribbon Panel," universities, and other sources pertaining to university reactors including their research and training capabilities, costs to operate, and operating data, and * Provide DOE with clear, near-term recommendations as to actions that should be taken by the Federal Government and a long-term strategy to assure the continued operation of vital university reactor facilities in

38

Restoration of the graphite memory of a reactor in the third power-generating unit of the Leningrad nuclear power plant  

Science Journals Connector (OSTI)

The restoration of the graphite masonry of cell 52-16 in the reactor in the third power-generating unit of the Leningrad nuclear power plant is described. The process reduces to moving...

V. I. Lebedev; Yu. V. Garusov; M. A. Pavlov; A. N. Peunov; E. P. Kozlov

1999-11-01T23:59:59.000Z

39

A probabilistic safety analysis of incidents in nuclear research reactors  

Science Journals Connector (OSTI)

......System for Research Reactor (IRSRR). Available...System for Research Reactor (IRSRR). Available...76. 7 Manual on reliability data collection for research reactor PSAs. (1992) IAEA...probabilistic safety analysis of incidents in nuclear......

Valdir Maciel Lopes; Gian Maria Agostinho Angelo Sordi; Mauricio Moralles; Tufic Madi Filho

2012-06-01T23:59:59.000Z

40

Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators  

SciTech Connect (OSTI)

The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

Palabrica, R.J.

1981-01-01T23:59:59.000Z

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


41

Irradiation Creep in Graphite  

SciTech Connect (OSTI)

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.

Ubic, Rick; Butt, Darryl; Windes, William

2014-03-13T23:59:59.000Z

42

Experimental Study of the Effect of Graphite Dispersion on the Heat Transfer Phenomena in a Reactor Cavity Cooling System  

SciTech Connect (OSTI)

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 (16.5 x 16.5 x 30.4 cm) used for this activity represents half of the reactor cavity with an electrically heated vessel. Water flowing through five vertical pipes removes the heat produced in the vessel and releases it into the environment by mixing with cold water in a large tank. The particle image velocimetry technique was used to study the velocity field of the air inside the cavity. A set of 52 thermocouples was installed in the facility to monitor the temperature profiles of the vessel, pipe walls, and air. Ten grams of a fine graphite powder (average particle size 2 m) was injected into the cavity through a spraying nozzle placed at the bottom of the vessel. The temperatures and air velocity field were recorded and compared with the measurements obtained before the graphite dispersion, showing a decrease of the temperature surfaces that was related to an increase in their emissivity. The results contribute to the understanding of RCCS capability in an accident scenario.

Rodolfo Vaghetto; Luigi Capone; Yassin A. Hassan

2011-05-31T23:59:59.000Z

43

Role of Nuclear Grade Graphite in Oxidation in Modular HTGRs  

SciTech Connect (OSTI)

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 1940s 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.

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

2014-11-01T23:59:59.000Z

44

Corrosion Minimization for Research Reactor Fuel  

SciTech Connect (OSTI)

Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

Eric Shaber; Gerard Hofman

2005-06-01T23:59:59.000Z

45

Activation analysis of concrete and graphite in the experimental reactor RUS  

Science Journals Connector (OSTI)

......in the whole reactor, as a function...consistency and the reliability of the nuclear...for activation analysis. We believe...industrial nuclear reactors. INTRODUCTION...consistency and the reliability of the nuclear...for activation analysis. We believe...industrial nuclear reactors. | CEA Saclay......

M. Cometto; D. Ridikas; M. C. Aubert; F. Damoy; D. Ancius

2005-12-20T23:59:59.000Z

46

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]

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...

Vaghetto, Rodolfo

2012-07-16T23:59:59.000Z

47

The Argonaut Reactor - Reactors designed/built by Argonne National  

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

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

48

Sterile Neutrino Search Using China Advanced Research Reactor  

E-Print Network [OSTI]

We study the feasibility of a sterile neutrino search at the China Advanced Research Reactor by measuring $\\bar {\

Gang Guo; Fang Han; Xiangdong Ji; Jianglai Liu; Zhaoxu Xi; Huanqiao Zhang

2013-06-18T23:59:59.000Z

49

Preparation of binderless nanopore-isotropic graphite for inhibiting the liquid fluoride salt and Xe135 penetration for molten salt nuclear reactor  

Science Journals Connector (OSTI)

Abstract Mesocarbon microbeads and the isostatic pressing method were used to prepare binderless nanopore-isotropic graphite (NPIG) as a neutron moderator and reflector, to inhibit liquid fluoride salt and Xe135 penetration during use in a molten salt nuclear reactor. The microstructure, thermophysical, and other properties of the NPIG were studied and compared with isostatic graphite (IG-110, TOYO TANSO CO., Ltd., Japan). A high-pressure reactor and a vacuum device were constructed to evaluate the molten salt and Xe135 penetration in the graphite, respectively. The results indicated that NPIG possessed a graphitization degree of 74% and excellent properties such as a high bending strength of 94.12.5MPa, a high compressive strength of 2303MPa, a low porosity of 8.7%, and an average pore diameter of 69nm. The fluoride salt occupation of IG-110 was 13.5wt% under 1.5atm, whereas the salt gain in NPIG remained steady even up to 10atm with an increase of <0.06wt%, demonstrating that the graphite could inhibit the liquid fluoride salt infiltration effectively. The helium diffusion coefficient for NPIG was 8.76נ10?5cm2/s, much less than 1.21נ10?2cm2/s for IG-110. The NPIG could effectively inhibit liquid fluoride salt and Xe135 penetration.

Jinliang Song; Yanling Zhao; Junpeng Zhang; Xiujie He; Baoliang Zhang; Pengfei Lian; Zhanjun Liu; Dongsheng Zhang; Zhoutong He; Lina Gao; Huihao Xia; Xingtai Zhou; Ping Huai; Quangui Guo; Lang Liu

2014-01-01T23:59:59.000Z

50

Ames Laboratory Research Reactor Facility Ames, Iowa  

Office of Legacy Management (LM)

,, *' ; . Final Radiological Condition of the Ames Laboratory Research Reactor Facility Ames, Iowa _, . AGENCY: Office of Operational Safety, Department of Energy ' ACTION: Notice of Availability of Archival Information Package SUMMARY: The'Office of Operational Safety of the Department O i Energy (DOE) has reviewed documentation relating to the decontamination and decommissioning operations conducted at the Ames Laboratory Research Reactor Facility, Ames, Iowa and has prepared an archival informati0.n package to permanently document the results of the action and the site conditions and use restriction placed on the . site at the tim e of release. This review is based on post-decontamination survey data and other pertinent documentation referenced in and included in the archival package. The material and

51

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

SciTech Connect (OSTI)

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.

Not Available

1987-05-01T23:59:59.000Z

52

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

E-Print Network [OSTI]

in it and how nuclear material flows between facilities. The goals of the safeguards system were established next, using the normal IAEA standards for the non-detection and false alarm probabilities. The 5 MWe Reactor was modeled for both plutonium production...

Rauch, Eric B.

2010-07-14T23:59:59.000Z

53

German Pebble Bed Research Reactor Highly Enriched Uranium (HEU...  

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

Potential Acceptance and Disposition of German Pebble Bed Research Reactor Highly Enriched Uranium (HEU) Fuel Environmental Assessment Maxcine Maxted, DOE-SR Used Nuclear Fuel...

54

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

National Nuclear Security Administration (NNSA)

Annular Core Research Reactor - Critical to Science-Based Weapons Design, Certification | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People...

55

Health physics research reactor reference dosimetry  

SciTech Connect (OSTI)

Reference neutron dosimetry is developed for the Health Physics Research Reactor (HPRR) in the new operational configuration directly above its storage pit. This operational change was physically made early in CY 1985. The new reference dosimetry considered in this document is referred to as the 1986 HPRR reference dosimetry and it replaces any and all HPRR reference documents or papers issued prior to 1986. Reference dosimetry is developed for the unshielded HPRR as well as for the reactor with each of five different shield types and configurations. The reference dosimetry is presented in terms of three different dose and six different dose equivalent reporting conventions. These reporting conventions cover most of those in current use by dosimetrists worldwide. In addition to the reference neutron dosimetry, this document contains other useful dosimetry-related data for the HPRR in its new configuration. These data include dose-distance measurements and calculations, gamma dose measurements, neutron-to-gamma ratios, ''9-to-3 inch'' ratios, threshold detector unit measurements, 56-group neutron energy spectra, sulfur fluence measurements, and details concerning HPRR shields. 26 refs., 11 figs., 31 tabs.

Sims, C.S.; Ragan, G.E.

1987-06-01T23:59:59.000Z

56

Reactor Safety Research Programs Quarterly Report October - December 1981  

SciTech Connect (OSTI)

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

Edler, S. K.

1982-03-01T23:59:59.000Z

57

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

E-Print Network [OSTI]

Gel-Form Graphene Monash researchers led by Dr. Dan Li have developed a novel method for converting natural graphite into highly porous graphene film for advanced applications. Figure 1: illustrates (eg capacitors, batteries and fuel cells) n LCD displays and photovoltaic devices n Composites

Albrecht, David

58

Light Water Reactors Technology Development - Nuclear Reactors  

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

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

59

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

SciTech Connect (OSTI)

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

Homan, F.J.

1990-06-20T23:59:59.000Z

60

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

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

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program The Department of Energy's (DOE's) Light Water Reactor Sustainability (LWRS) Program is a five year effort that 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 operation of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging

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


61

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear  

National Nuclear Security Administration (NNSA)

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Video Gallery > Maria Research Reactor loaded with LEU - ... Maria Research Reactor loaded with LEU - Otwock, Poland Maria Research Reactor loaded with LEU - Otwock, Poland

62

RERTR program reduces use of enriched uranium in research reactors  

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

RERTR program reduces use of enriched uranium in research reactors RERTR program reduces use of enriched uranium in research reactors worldwide Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share RERTR program reduces use of enriched uranium in research reactors worldwide The High Flux Reactor in Petten, the Netherlands READY TO CONVERT - The High Flux Reactor in Petten, the Netherlands, has

63

Advanced Reactor Research and Development Funding Opportunity Announcement  

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

Reactor Research and Development Funding Opportunity Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) sponsors a program of research, development, and demonstration related to advanced non-light water reactor concepts. A goal of the program is to facilitate greater engagement between DOE and industry. During FY12, DOE established a Technical Review Panel (TRP) process to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. That process involved the use of a Request for Information (RFI) to solicit concept information from industry and engage technical experts to evaluate those concepts. Having completed this process, DOE desires to

64

Advanced Reactor Research and Development Funding Opportunity Announcement  

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

Advanced Reactor Research and Development Funding Opportunity Advanced Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) sponsors a program of research, development, and demonstration related to advanced non-light water reactor concepts. A goal of the program is to facilitate greater engagement between DOE and industry. During FY12, DOE established a Technical Review Panel (TRP) process to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. That process involved the use of a Request for Information (RFI) to solicit concept information from industry and engage technical experts to evaluate those concepts. Having completed this process, DOE desires to

65

Sandia National Laboratories: Research: Facilities: Sandia Pulsed Reactor  

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

Sandia Pulsed Reactor Facility - Critical Experiments Sandia Pulsed Reactor Facility - Critical Experiments Sandia scientist John Ford places fuel rods in the Seven Percent Critical Experiment (7uPCX) at the Sandia Pulsed Reactor Facility Critical Experiments (SPRF/CX) test reactor - a reactor stripped down to its simplest form. The Sandia Pulsed Reactor Facility - Critical Experiments (SPRF/CX) provides a flexible, shielded location for performing critical experiments that employ different reactor core configurations and fuel types. The facility is also available for hands-on nuclear criticality safety training. Research and other activities The 7% series, an evaluation of various core characteristics for higher commercial-fuel enrichment, is currently under way at the SPRF/CX. Past critical experiments at the SPRF/CX have included the Burnup Credit

66

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 Oak Ridge selected as site for World War II Manhattan Project First sustained and controlled nuclear 430 nuclear power reactors are operating in the world, and 103 nuclear power plants produce 20

Pennycook, Steve

67

Photoneutron effects on pulse reactor kinetics for the Annular Core Research Reactor (ACRR).  

SciTech Connect (OSTI)

The Annular Core Research Reactor (ACRR) is a swimming-pool type pulsed reactor that maintains an epithermal neutron flux and a nine-inch diameter central dry cavity. One of its uses is neutron and gamma-ray irradiation damage studies on electronic components under transient reactor power conditions. In analyzing the experimental results, careful attention must be paid to the kinetics associated with the reactor to ensure that the transient behavior of the electronic device is understood. Since the ACRR fuel maintains a substantial amount of beryllium, copious quantities of photoneutrons are produced that can significantly alter the expected behavior of the reactor power, especially following a reactor pulse. In order to understand these photoneutron effects on the reactor kinetics, the KIFLE transient reactor-analysis code was modified to include the photoneutron groups associated with the beryllium. The time-dependent behavior of the reactor power was analyzed for small and large pulses, assuming several initial conditions including following several pulses during the day, and following a long steady-state power run. The results indicate that, for these types of initial conditions, the photoneutron contribution to the reactor pulse energy can have a few to tens of percent effect.

Parma, Edward J., Jr.

2009-06-01T23:59:59.000Z

68

A probabilistic safety analysis of incidents in nuclear research reactors  

Science Journals Connector (OSTI)

......ICRP-64. INTRODUCTION Nuclear research reactors are considered important tools in nuclear science. For more than...as well as prevention policy, have stimulated the development...level 3 in the International Nuclear Events Scale (INES) of......

Valdir Maciel Lopes; Gian Maria Agostinho Angelo Sordi; Mauricio Moralles; Tufic Madi Filho

2012-06-01T23:59:59.000Z

69

Managerial Aspects of BNCT at a Nuclear Research Reactor  

Science Journals Connector (OSTI)

All BNCT facilities worldwide, performing clinical trials, are presently located at a nuclear research reactor. They are nevertheless, to all ... 43/EURATOM which stipulates that radiotherapy quality assurance programmes

Wolfgang A. G. Sauerwein; Ray Moss

2012-01-01T23:59:59.000Z

70

Argonne's pyroprocessing and advanced reactor research featured on WGN  

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

Argonne's pyroprocessing and advanced reactor research featured on WGN Argonne's pyroprocessing and advanced reactor research featured on WGN radio Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share Argonne's pyroprocessing and advanced reactor research featured on WGN radio Uranium dendrites These tiny branches, or "dendrites," of pure uranium form when engineers

71

Uranium Oxide Aerosol Transport in Porous Graphite  

SciTech Connect (OSTI)

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 reactors lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

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

2012-01-23T23:59:59.000Z

72

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

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

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

73

Achievements: Nuclear Reactors designed/built by Argonne National  

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

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

74

Evaluation of co-cokes from bituminous coal with vacuum resid or decant oil, and evaluation of anthracites, as precursors to graphite.  

E-Print Network [OSTI]

??Graphite is utilized as a neutron moderator and structural component in some nuclear reactor designs. During the reactor operaction the structure of graphite is damaged (more)

Nyathi, Mhlwazi

2011-01-01T23:59:59.000Z

75

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

E-Print Network [OSTI]

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 ...

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

1968-01-01T23:59:59.000Z

76

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab | Department  

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

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

77

Graphite Oxidation Simulation in HTR Accident Conditions  

SciTech Connect (OSTI)

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 900C 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.

Mohamed El-Genk

2012-10-19T23:59:59.000Z

78

Nuclear Energy Research Brookhaven National  

E-Print Network [OSTI]

Nuclear Energy Research Brookhaven National Laboratory William C. Horak, Chair Nuclear Science and Technology Department #12;BNL Nuclear Energy Research Brookhaven Graphite Research Reactor - 1948 National&T Department #12;Nuclear Energy Today 435 Operable Power Reactors, 12% electrical generation (100 in US, 19

Ohta, Shigemi

79

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

SciTech Connect (OSTI)

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 900C 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, 2Cr-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.

J. K. Wright; R. N. Wright

2008-04-01T23:59:59.000Z

80

Research reactor usage at the Idaho National Engineering Laboratory in support of university research and education  

SciTech Connect (OSTI)

The Idaho National Engineering Laboratory is a US Department of Energy laboratory which has a substantial history of research and development in nuclear reactor technologies. There are a number of available nuclear reactor facilities which have been incorporated into the research and training needs of university nuclear engineering programs. This paper addresses the utilization of the Advanced Reactivity Measurement Facility (ARMF) and the Coupled Fast Reactivity Measurement Facility (CFRMF) for thesis and dissertation research in the PhD program in Nuclear Science and Engineering by the University of Idaho and Idaho State University. Other reactors at the INEL are also being used by various members of the academic community for thesis and dissertation research, as well as for research to advance the state of knowledge in innovative nuclear technologies, with the EBR-II facility playing an essential role in liquid metal breeder reactor research. 3 refs.

Woodall, D.M.; Dolan, T.J.; Stephens, A.G. (Idaho National Engineering Lab., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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We encourage you to perform a real-time search of NLEBeta
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81

RERTR 2009 (Reduced Enrichment for Research and Test Reactors)  

SciTech Connect (OSTI)

The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Test Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.

Totev, T.; Stevens, J.; Kim, Y. S.; Hofman, G.; Matos, J.; Hanan, N.; Garner, P.; Dionne, B.; Olson, A.; Feldman, E.; Dunn, F.; Nuclear Engineering Division; Atomic Research Center; Inst. of Nuclear Physics; LLNL; INL; Korea Atomic Energy Research Inst.; Comisi?n Nacional de Energ?a At?mica; Nuclear Reactor Lab.; Inst. of Atomic Energy-Poland; AECL-Canada; Hungarian Academy of Sciences KFKI Atomic Energy Research Inst.; Japan Atomic Energy Agency; Nuclear Power Inst. of China; Kyoto Univ. Research Reactor Inst.

2010-03-01T23:59:59.000Z

82

Early Exploration - Reactors designed/built by Argonne National Laboratory  

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

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

83

Preliminary analysis of graphite dust releasing behavior in accident for HTR  

SciTech Connect (OSTI)

The behavior of the graphite dust is important to the safety of High Temperature Gas-cooled Reactors. This study investigated the flow of graphite dust in helium mainstream. The analysis of the stresses acting on the graphite dust indicated that gas drag played the absolute leading role. Based on the understanding of the importance of gas drag, an experimental system is set up for the research of dust releasing behavior in accident. Air driven by centrifugal fan is used as the working fluid instead of helium because helium is expensive, easy to leak which make it difficult to seal. The graphite particles, with the size distribution same as in HTR, are added to the experiment loop. The graphite dust releasing behavior at the loss-of-coolant accident will be investigated by a sonic nozzle. (authors)

Peng, W.; Yang, X. Y.; Yu, S. Y.; Wang, J. [Inst. of Nuclear and New Energy Technology, Tsinghua Univ., Beijing100084 (China)

2012-07-01T23:59:59.000Z

84

AGC-2 Graphite Preirradiation Data Package  

SciTech Connect (OSTI)

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.

David Swank; Joseph Lord; David Rohrbaugh; William Windes

2012-10-01T23:59:59.000Z

85

Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos  

SciTech Connect (OSTI)

This reports 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.

Heeger, Karsten M [Yale University

2014-09-13T23:59:59.000Z

86

Status of reduced enrichment program for research reactors in Japan  

SciTech Connect (OSTI)

The status of reduced enrichment program for research reactors in Japan will be reviewed. The reduced enrichment programs for the JRR-3M, JRR-4 and JMTR of Japan Atomic Energy Agency (JAEA, former name is Japan Atomic Energy Research Institute (JAERI)) has been completed by 1999, and the reactors are being satisfactory operated using LEU fuels. The KUR of Kyoto University Research Reactor Institute (KURRI) has been partially completed and is still in progress under the Joint Study Program with Argonne National Laboratory (ANL). The JRR-3M using LEU silicide fuel elements have done a functional test by the Japanese Government in 2000, and the property of the reactor core was satisfied. JAEA has established a 'U-Mo fuel ad hoc committee' and has been studying the U-Mo fuel installation plan by carefully observing the development situation of the U-Mo fuel. In KURRI, the KUR has terminated its operation using HEU fuel on February 2006. The HEU KUR spent fuel elements will be sent to the U.S. by March 2008. Licensing for the full core conversion of KUR to LEU fuel is under progress and the core conversion to LEU is expected to be completed in 2008. (author)

Unesaki, Hironobu [Research Reactor Institute, Kyoto University, Asashiro-nishi 2-1010, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Ohta, Kazunori; Inoue, Takeshi [Japan Atomic Energy Agency, Shirakata Shirane 2-4, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan)

2008-07-15T23:59:59.000Z

87

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

E-Print Network [OSTI]

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 ...

Kempf, Stephanie Anne

2011-01-01T23:59:59.000Z

88

Inhibition of Oxidation in Nuclear Graphite  

SciTech Connect (OSTI)

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 900C. 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.

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

2013-10-01T23:59:59.000Z

89

Graphite Gamma Scan Results  

SciTech Connect (OSTI)

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 Ci at a measurement distance of 47 cm.

Mark W. Drigert

2014-04-01T23:59:59.000Z

90

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab  

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

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

91

SUPPLEMENT ANALYSIS OF FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL  

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

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

92

Manhattan Project: Production Reactor (Pile) Design, Met Lab, 1942  

Office of Scientific and Technical Information (OSTI)

Schematic of the X-10 Graphite Reactor, Oak Ridge PRODUCTION REACTOR (PILE) DESIGN Schematic of the X-10 Graphite Reactor, Oak Ridge PRODUCTION REACTOR (PILE) DESIGN (Met Lab, 1942) Events > The Plutonium Path to the Bomb, 1942-1944 Production Reactor (Pile) Design, 1942 DuPont and Hanford, 1942 CP-1 Goes Critical, December 2, 1942 Seaborg and Plutonium Chemistry, 1942-1944 Final Reactor Design and X-10, 1942-1943 Hanford Becomes Operational, 1943-1944 By 1942, scientists had established that some of the uranium exposed to radioactivity in a reactor (pile) would eventually decay into plutonium, which could then be separated by chemical means from the uranium. Important theoretical research on this was ongoing, but the work was scattered at various universities from coast to coast. In early 1942, Arthur Compton arranged for all pile research to be moved to the Met Lab at the University of Chicago.

93

Chemical Characterization and Removal of Carbon-14 from Irradiated Graphite II - 13023  

SciTech Connect (OSTI)

Approximately 250,000 tonnes of irradiated graphite waste exists worldwide and that quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation IV gas-cooled, graphite moderated reactors. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from some nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research presented last year and updated here is to identify the chemical form of C-14 in irradiated graphite and develop a practical method by which C-14 can be removed. A nuclear-grade graphite, NBG-18, and a high-surface-area graphite foam, POCOFoam{sup R}, were exposed to liquid nitrogen (to increase the quantity of C-14 precursor) and neutron-irradiated (10{sup 13} neutrons/cm{sup 2}/s). Finer grained NBG-25 was not exposed to liquid nitrogen prior to irradiation at a neutron flux on the order of 10{sup 14} /cm{sup 2}/s. Characterization of pre- and post-irradiation graphite was conducted to determine the chemical environment and quantity of C-14 and its precursors via the use of surface sensitive characterization techniques. Scanning Electron Microscopy (SEM) was used to evaluate the morphological features of graphite samples. The concentration, chemical composition, and bonding characteristics of C-14 and its precursors were determined through X-ray Photoelectron Spectroscopy (XPS), Time-of-Flight Secondary Ion Mass Spectrometry (SIMS), and Energy Dispersive X-ray Analysis Spectroscopy (EDX). Results of post-irradiation characterization of these materials indicate a variety of surface functional groups containing carbon, oxygen, nitrogen and hydrogen. During thermal treatment, irradiated graphite samples are heated in the presence of an inert carrier gas (with or without oxidant gas), which carries off gaseous products released during treatment. Graphite gasification occurs via interaction with adsorbed oxygen complexes. Experiments in argon were performed at 900 deg. C and 1400 deg. C to evaluate the selective removal of C-14. Thermal treatment also was performed with the addition of 3 and 5 volume % oxygen at temperatures 700 deg. C and 1400 deg. C. Thermal treatment experiments were evaluated for the effective selective removal of C-14. Lower temperatures and oxygen levels correlated to more efficient C-14 removal. (authors)

Dunzik-Gougar, Mary Lou; Cleaver, James; LaBrier, Daniel; McCrory, Shilo; Smith, Tara E. [Idaho State University: 1776 Science Center Dr., Idaho Falls, ID, 83401 (United States)] [Idaho State University: 1776 Science Center Dr., Idaho Falls, ID, 83401 (United States)

2013-07-01T23:59:59.000Z

94

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

SciTech Connect (OSTI)

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)

Snoj, L. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Sklenka, L.; Rataj, J. [Dept. of Nuclear Reactor, Czech Technical Univ. in Prague, V Holesovickach 2, 180 00 Prague 8 (Czech Republic); Boeck, H. [Vienna Univ. of Technology/Atominstitut, Stadionallee 2, 1020 Vienna (Austria)

2012-07-01T23:59:59.000Z

95

Structural graphitic carbon foams  

SciTech Connect (OSTI)

Graphitic carbon foams are a unique material form with very high structural and thermal properties at a light weight. A process has been developed to produce microcellular, open-celled graphitic foams. The process includes heating a mesophase pitch preform above the pitch melting temperature in a pressurized reactor. At the appropriate time, the pressure is released, the gas nucleates bubbles, and these bubbles grow forming the pitch into the foam structure. The resultant foamed pitch is then stabilized in an oxygen environment. At this point a rigid structure exists with some mechanical integrity. The foam is then carbonized to 800 C followed by a graphitization to 2700 C. The shear action from the growing bubbles aligns the graphitic planes along the foam struts to provide the ideal structure for good mechanical properties. Some of these properties have been characterized for some of the foam materials. It is known that variations of the blowing temperature, blowing pressure and saturation time result in foams of variously sized with mostly open pores; however, the mechanism of bubble nucleation is not known. Therefore foams were blown with various gases to begin to determine the nucleation method. These gases are comprised of a variety of molecular weights as well as a range of various solubility levels. By examining the resultant structures of the foam, differences were noted to develop an explanation of the foaming mechanism.

Kearns, K.M.; Anderson, H.J. [Air Force Lab., Wright-Patterson AFB, OH (United States). Materials and Mfg. Directorate

1998-12-31T23:59:59.000Z

96

Research in nondestructive evaluation techniques for nuclear reactor concrete structures  

SciTech Connect (OSTI)

The purpose of the Materials Aging and Degradation (MAaD) Pathway of the Department of Energy's Light Water Reactor Sustainability (LWRS) Program is to develop the scientific basis for understanding and predicting longterm environmental degradation behavior of material in nuclear power plants and to provide data and methods to assess the performance of systems, structures, and components (SSCs) essential to safe and sustained nuclear power plant operations. The understanding of aging-related phenomena and their impacts on SSCs is expected to be a significant issue for any nuclear power plant planning for long-term operations (i.e. service beyond the initial license renewal period). Management of those phenomena and their impacts during long-term operations can be better enable by improved methods and techniques for detection, monitoring, and prediction of SSC degradation. The MAaD Pathway R and D Roadmap for Concrete, 'Light Water Reactor Sustainability Nondestructive Evaluation for Concrete Research and Development Roadmap', focused initial research efforts on understanding the recent concrete issues at nuclear power plants and identifying the availability of concrete samples for NDE techniques evaluation and testing. [1] An overview of the research performed by ORNL in these two areas is presented here.

Clayton, Dwight; Smith, Cyrus [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

2014-02-18T23:59:59.000Z

97

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

SciTech Connect (OSTI)

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)

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)

2012-07-01T23:59:59.000Z

98

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

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

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

99

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

SciTech Connect (OSTI)

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.

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

1995-12-31T23:59:59.000Z

100

USA/FRG umbrella agreement for cooperation in GCR [Gas Cooled Reactor] development: Fuel, fission products and graphite subprogram. Part 1, Management meeting report: Part 2, Revised subprogram plan, Revision 10  

SciTech Connect (OSTI)

This Subprogram Plan describes cooperative work in the areas of HTR fuel and graphite development and fission product studies that is being carried out under US/FRG/Swiss Implementing Agreement for cooperation in Gas Cooled Reactor development. Only bilateral US/FRG cooperation is included, since it is the only active work in this subprogram area at this time. The cooperation has been in progress since February 1977. A number of Project Work Statements have been developed in each of the major areas of the subprogram, and work on many of them is in progress. The following specific areas are included in the scope of this plan: fuel development; graphite development; fission product release; and fission product behavior outside the fuel elements.

NONE

1986-05-01T23:59:59.000Z

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


101

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

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

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

102

E-Print Network 3.0 - application research reactor Sample Search...  

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

Nuclear Technologies 28 Research Aptitude Problem 1 Scavenging of aerosol particles by ice crystals Summary: strategies that would be required to operate these reactor systems....

103

Graphite-moderated, gas-cooled, and water-moderated, water-cooled reactors as power units in nuclearelectric power stations  

Science Journals Connector (OSTI)

The present article reviews a number of papers submitted at the Second International Conference on the Peaceful Uses of Atomic Energy bearing on water-cooled, water-moderated, graphite-moderated, and gas-coole...

Yu. I. Koryakin

1960-11-01T23:59:59.000Z

104

PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.  

SciTech Connect (OSTI)

Detailed reactor physics and 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 analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core power transient is terminated by a reactor trip at 26 MW. The calculations show that both the peak reactor power and the excursion energy depend on the negative reactivity insertion from reactor trip. In one of the loss-of-flow accidents offsite electrical power is assumed lost to the three operating primary pumps. A slightly delayed reactor scram is initiated as a result of primary flow coast down. The RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur, because of the momentum of the coolant flowing through the fuel channels and the negative scram reactivity insertion. For both the primary pump seizure and inadvertent throttling of a flow control valve, the RELAP5 analyses indicate that the reduction in power following the trip is sufficient to ensure that there is adequate margin to CHF and that the fuel cladding does not fail. The analysis of the loss-of-flow accident in the extremely unlikely case where both shutdown pumps fail, shows that the cooling provided by the D{sub 2}O is sufficient to ensure the cladding does not fail. The power distributions were examined for a set of fuel misloadings in which a fresh fuel element is moved from a peripheral low-reactivity location to a central high-reactivity location. The calculations show that there is adequate margin to CHF and the cladding does not fail. An additional analysis was performed to simulate the operation at low power (500 kW) without forced flow cooling. The result indicates that natural convection cooling is adequate for operation of the NBSR at a power level of 500 kW.

CHENG,L.HANSON,A.DIAMOND,D.XU,J.CAREW,J.RORER,D.

2004-03-31T23:59:59.000Z

105

Nuclear plant-aging research on reactor protection systems  

SciTech Connect (OSTI)

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.

Meyer, L.C.

1988-01-01T23:59:59.000Z

106

Foreign Research Reactor Spent Nuclear Fuel Acceptance Program  

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

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

107

MODULAR PEBBLE BED REACTOR PROJECT UNIVERSITY RESEARCH CONSORTIUM  

E-Print Network [OSTI]

Annual Report Page ii MODULAR PEBBLE BED REACTOR ABSTRACT This project is developing a fundamental. Publication of an archival journal article covering this work is being prepared. · Detailed gas reactor Abstract

108

PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.  

SciTech Connect (OSTI)

Detailed reactor physics and 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 analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional MCNP Monte Carlo neutron and photon transport calculations were performed to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model including the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core power transient is terminated by a reactor trip at 30 MW. The calculations show that both the peak reactor power and the excursion energy depend on the negative reactivity insertion from reactor trip. Two cases were considered for loss of electrical power. In the first case offsite power is lost, resulting in an immediate scram caused by loss of power to the control rod system. In the second case power is lost to only the three operating primary pumps, resulting in a slightly delayed scram when loss-of-flow is detected as the pumps coast down. In both instances, RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur, because of the momentum of the coolant flowing through the fuel channels and the negative scram reactivity insertion. For both the primary pump seizure and inadvertent throttling of a flow control valve, the RELAP5 analyses indicate that the reduction in power following the trip is sufficient to ensure that there is adequate margin to CHF and the fuel cladding does not fail. The analysis of the loss-of-flow accident in the extremely unlikely case where both shutdown pumps fail shows that the cooling provided by the D{sub 2}O is sufficient to ensure the cladding does not fail. The power distributions were examined for a set of fuel misloadings in which a fresh fuel element is moved from a peripheral low-reactivity location to a central high-reactivity location. The calculations show that there is adequate margin to CHF and the cladding does not fail.

CAREW,J.CHENG,L.HANSON,AXU,J.RORER,D.DIAMOND,D.

2003-08-26T23:59:59.000Z

109

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

E-Print Network [OSTI]

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

Montfrooij, Wouter

110

Early Argonne reactor lit the way for worldwide nuclear industry -  

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

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

111

Development and transfer of fuel fabrication and utilization technology for research reactors  

SciTech Connect (OSTI)

Approximately 300 research reactors supplied with US-enriched uranium are currently in operation in about 40 countries, with a variety of types, sizes, experiment capabilities and applications. Despite the usefulness and popularity of research reactors, relatively few innovations in their core design have been made in the last fifteen years. The main reason can be better understood by reviewing briefly the history of research reactor fuel technology and enrichment levels. Stringent requirements on the enrichment of the uranium to be used in research reactors were considered and a program was launched to assist research reactors in continuing their operation with the new requirements and with minimum penalties. The goal of the new program, the Reduced Enrichment Research and Test Reactor (RERTR) Program, is to develop the technical means to utilize LEU instead of HEU in research reactors without significant penalties in experiment performance, operating costs, reactor modifications, and safety characteristics. This paper reviews briefly the RERTR Program activities with special emphasis on the technology transfer aspects of interest to this conference.

Travelli, A.; Domagala, R.F.; Matos, J.E.; Snelgrove, J.L.

1982-01-01T23:59:59.000Z

112

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

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

Environmental Protection Division Environmental Protection Division Home Reactor Projects Celebrating DOE's Cleanup Accomplishments (PDF) Brookhaven Graphite Research Reactor(BGRR) BGRR Overview BGRR Complex Description Decommissioning Decision BGRR Complex Cleanup Actions BGRR Documents BGRR Science & Accomplishments High Flux Beam Reactor (HFBR) HFBR Overview HFBR Complex Description Decommissioning Decision HFBR Complex Cleanup Actions HFBR Documents HFBR Science & Accomplishments Groundwater Protection Group Environmental Protection Division Contact > See also: HFBR Science & Accomplishments High Flux Beam Reactor Under the U.S. Department of Energy (DOE), the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) underwent stabilization and partial decommissioning to prepare the HFBR confinement for long-term safe

113

Medical isotope production: A new research initiative for the Annular Core Research Reactor  

SciTech Connect (OSTI)

An investigation has been performed to evaluate the capabilities of the Annular Core Research Reactor and its supporting Hot Cell Facility for the production of {sup 99}Mo and its separation from the fission product stream. Various target irradiation locations for a variety of core configurations were investigated, including the central cavity, fuel and reflector locations, and special target configurations outside the active fuel region. Monte Carlo techniques, in particular MCNP using ENDF B-V cross sections, were employed for the evaluation. The results indicate that the reactor, as currently configured, and with its supporting Hot Cell Facility, would be capable in meeting the current US demand if called upon. Modest modifications, such as increasing the capacity of the external heat exchangers, would permit significantly higher continuous power operation and even greater {sup 99}Mo production ensuring adequate capacity for future years.

Coats, R.L.; Parma, E.J.

1993-12-31T23:59:59.000Z

114

Engineered Graphite Oxide Materials for Application in Water Purification  

Science Journals Connector (OSTI)

Engineered Graphite Oxide Materials for Application in Water Purification ... The research results could open avenues for developing low-cost water purification materials for the developing economies. ... water purification; graphite oxide; mercury removal; diazonium chemistry; Rhodamine B; sand coating ...

Wei Gao; Mainak Majumder; Lawrence B. Alemany; Tharangattu N. Narayanan; Miguel A. Ibarra; Bhabendra K. Pradhan; Pulickel M. Ajayan

2011-05-13T23:59:59.000Z

115

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

E-Print Network [OSTI]

A RESEARCH ON SEAMLESS PLATFORM CHANGE OF REACTOR PROTECTION SYSTEM FROM PLC TO FPGA JUNBEOM YOO1 1. INTRODUCTION A safety grade PLC is an industrial digital computer used to develop safety-critical systems such as RPS (Reactor Protection System) for nuclear power plants. The software loaded into a PLC

116

Risk management at the Oak Ridge National Laboratory research reactors  

SciTech Connect (OSTI)

In November of 1986, the High Flux Isotope Reactor (HFIR) was shut down by Oak Ridge National Laboratory (ORNL) due to a concern regarding embrittlement of the reactor vessel. A massive review effort was undertaken by ORNL and the Department of Energy (DOE). This review resulted in an extensive list of analyses and design modifications to be completed before restart could take place. The review also focused on the improvement of management practices including implementation of several of the Institute of Nuclear Power Operations (INPO) requirements. One of the early items identified was the need to perform a Probabilistic Risk Assessment (PRA) on the reactor. It was decided by ORNL management that this PRA would not be just an exercise to assess the ``bottom`` line in order to restart, but would be used to improve the overall safety of the reactor, especially since resources (both manpower and dollars) were severely limited. The PRA would become a basic safety tool to be used instead of a more standard deterministic approach to safety used in commercial reactor power plants. This approach was further reinforced, because the reactor was nearly 25 years old at this time, and the design standards and regulations had changed significantly since the original design, and many of the safety issues could not be addressed by compliance to codes and standards.

Flanagan, G.F.; Linn, M.A.; Proctor, L.D.; Cook, D.H.

1994-12-31T23:59:59.000Z

117

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

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

Light Water Reactor Fuel Cladding Research 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 purpose of this research is to furnish U.S. industry (EPRI, Areva, Westinghouse), and regulators (NRC) with much-needed data supporting safe and economical nuclear power generation and used fuel management. LWR fuel cladding work is tightly integrated with ORNL accident tolerant fuel development and used fuel disposition programs thereby providing a powerful capability that couples basic materials science research with the nuclear applications research and development. The ORNL LWR fuel cladding program consists of five complementary areas of research: Accident tolerant fuel and cladding material testing under design

118

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

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

U.S. University Research Reactors to be Converted From Highly U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium Two U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium April 11, 2005 - 11:34am Addthis WASHINGTON, D.C. - As part of the Bush administration's aggressive effort to reduce the amount of weapons-grade nuclear material worldwide, Secretary of Energy Samuel W. Bodman announced today that the Department 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 University. This effort, by DOE's National Nuclear Security Administration (NNSA) and the Office of Nuclear Energy, Science and Technology, are the latest steps

119

Effect of reduced enrichment on the fuel cycle for research reactors  

SciTech Connect (OSTI)

The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel.

Travelli, A.

1982-01-01T23:59:59.000Z

120

Benchmarking of Graphite Reflected Critical Assemblies of UO2  

SciTech Connect (OSTI)

A series of experiments were carried out in 1963 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for use in space reactor research programs. A core containing 93.2% enriched UO2 fuel rods was used in these experiments. The first part of the experimental series consisted of 253 tightly-packed fuel rods (1.27 cm triangular pitch) with graphite reflectors [1], the second part used 253 graphite-reflected fuel rods organized in a 1.506 cm triangular pitch [2], and the final part of the experimental series consisted of 253 beryllium-reflected fuel rods with a 1.506 cm triangular pitch. [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. The first part of this experimental series has 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] and is 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. [6

Margaret A. Marshall; John D. Bess

2011-11-01T23:59:59.000Z

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


121

German concept and status of the disposal of spent fuel elements from German research reactors  

SciTech Connect (OSTI)

Eight research reactors with a power {>=} 100 kW are currently being operated in the Federal Republic of Germany. These comprise three TRIGA-type reactors (power 100 kW to 250 kW), four swimming-pool reactors (power 1 MW to 10 MW) and one DIDO type reactor (power 23 MW). The German research reactors are used for neutron scattering for basic research in the field of solid state research, neutron metrology, for the fabrication of isotopes and for neutron activation analysis for medicine and biology, for investigating the influence of radiation on materials and for nuclear fuel behavior. It will be vital to continue current investigations in the future. Further operation of the German research reactors is therefore indispensable. Safe, regular disposal of the irradiated fuel elements arising now and in future operation is of primary importance. Furthermore, there are several plants with considerable quantities of spent fuel, the safe disposal of which is a matter of urgency. These include above all the VKTA facilities in Rossendorf and also the TRIGA reactors, where disposal will only be necessary upon decommissioning. The present paper report is concerned with the disposal of fuel from the German research reactors. It briefly deals with the situation in the USA since the end of 1988, describes interim solutions for current disposal requirements and then mainly concentrates on the German disposal concept currently being prepared. This concept initially envisages the long-term (25--50 years) dry interim storage of fuel elements in special containers in a central German interim store with subsequent direct final disposal without reprocessing of the irradiated fuel.

Komorowski, K. [Bundesministerium fuer Bildung Wissenschaft, Bonn (Germany); Storch, S.; Thamm, G. [Forschungszentrum Juelich GmbH (Germany)

1995-12-31T23:59:59.000Z

122

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

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

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

123

Modeling Fission Product Sorption in Graphite Structures  

SciTech Connect (OSTI)

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).

Szlufarska, Izabela [University of Wisconsin, Madison, WI (United States); Morgan, Dane [University of Wisconsin, Madison, WI (United States); Allen, Todd [University of Wisconsin, Madison, WI (United States)

2013-04-08T23:59:59.000Z

124

Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report  

SciTech Connect (OSTI)

This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOEs Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

2002-11-01T23:59:59.000Z

125

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321  

SciTech Connect (OSTI)

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States); Moore, E.N. [Moore Nuclear Energy, LLC (United States)] [Moore Nuclear Energy, LLC (United States)

2013-07-01T23:59:59.000Z

126

Materials Research Needs for Near-Term Nuclear Reactors  

Science Journals Connector (OSTI)

Technical Paper / NSF Workshop on the Research Needs of the Next Generation Nuclear Power Technology / Material

John R. Weeks

127

Modular Pebble Bed Reactor High Temperature Gas Reactor  

E-Print Network [OSTI]

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

128

Chapter 20: Graphite  

SciTech Connect (OSTI)

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.

Burchell, Timothy D [ORNL

2012-01-01T23:59:59.000Z

129

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

SciTech Connect (OSTI)

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

Parma, E.J.

1997-11-01T23:59:59.000Z

130

GUM Analysis for TIMS and SIMS Isotopic Ratios in Graphite  

SciTech Connect (OSTI)

This report describes GUM calculations for TIMS and SIMS isotopic ratio measurements of reactor graphite samples. These isotopic ratios are used to estimate reactor burn-up, and currently consist of various ratios of U, Pu, and Boron impurities in the graphite samples. The GUM calculation is a propagation of error methodology that assigns uncertainties (in the form of standard error and confidence bound) to the final estimates.

Heasler, Patrick G.; Gerlach, David C.; Cliff, John B.; Petersen, Steven L.

2007-04-01T23:59:59.000Z

131

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

SciTech Connect (OSTI)

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.

Not Available

1993-07-01T23:59:59.000Z

132

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

Science Journals Connector (OSTI)

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

Ian Hore-Lacy

2007-01-01T23:59:59.000Z

133

Pyrolytic graphite production : automation of material placement  

E-Print Network [OSTI]

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 ...

Olle, Chase R

2014-01-01T23:59:59.000Z

134

Using low-enriched uranium in research reactors: The RERTR program  

SciTech Connect (OSTI)

The goal of the RERTR program is to minimize and eventually eliminate use of highway enriched uranium (HEU) in research and test reactors. The program has been very successful, and has developed low-enriched uranium (LEU) fuel materials and designs which can be used effectively in approximately 90 percent of the research and test reactors which used HEU when the program began. This progress would not have been possible without active international cooperation among fuel developers, commercial vendors, and reactor operators. The new tasks which the RERTR program is undertaking at this time include development of new and better fuels that will allow use of LEU fuels in all research and test reactors; cooperation with Russian laboratories, which will make it possible to minimize and eventually eliminate use of HEU in research reactors throughout the world, irrespective of its origin; and development of an LEU-based process for the production of {sup 99}Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program.

Travelli, A.

1994-05-01T23:59:59.000Z

135

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

E-Print Network [OSTI]

??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 (more)

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

2008-01-01T23:59:59.000Z

136

Major Safety Aspects of Advanced Candu Reactor and Associated Research and Development  

SciTech Connect (OSTI)

The Advanced Candu{sup R} Reactor design is built on the proven technology of existing Candu plants and on AECL's knowledge base acquired over decades of nuclear power plant design, engineering, construction and research. Two prime objectives of ACR-700TM1 are cost reduction and enhanced safety. To achieve them some new features were introduced and others were improved from the previous Candu 6 and Candu 9 designs. The ACR-700 reactor design is based on the modular concept of horizontal fuel channels surrounded by a heavy water moderator, the same as with all Candu reactors. The major novelty in the ACR-700 is the use of slightly enriched fuel and light water as coolant circulating in the fuel channels. This results in a more compact reactor design and a reduction of heavy water inventory, both contributing to a significant decrease in cost compared to Candu reactors, which employ natural uranium as fuel and heavy water as coolant. The reactor core design adopted for ACR-700 also has some features that have a bearing on inherent safety, such as negative power and coolant void reactivity coefficient. Several improvements in engineered safety have been made as well, such as enhanced separation of the safety support systems. Since the ACR-700 design is an evolutionary development of the currently operating Candu plants, limited research is required to extend the validation database for the design and the supporting safety analysis. A program of safety related research and development has been initiated to address the areas where the ACR-700 design is significantly different from the Candu designs. This paper describes the major safety aspects of the ACR-700 with a particular focus on novel features and improvements over the existing Candu reactors. It also outlines the key areas where research and development efforts are undertaken to demonstrate the effectiveness and robustness of the design. (authors)

Bonechi, M.; Wren, D.J.; Hopwood, J.M. [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

2002-07-01T23:59:59.000Z

137

On the operator action analysis to reduce operational risk in research reactors  

Science Journals Connector (OSTI)

Abstract Human errors during operation and the resulting increase in operational risk are major concerns for nuclear reactors, just as they are for all industries. Additionally, human reliability analysis together with probabilistic risk analysis is a key element in reducing operational risk. The purpose of this paper is to analyze human reliability using appropriate methods for the probabilistic representation and calculation of human error to be used alongside probabilistic risk analysis in order to reduce the operational risk of the reactor operation. We present a technique for human error rate prediction and standardized plant analysis risk. Human reliability methods have been utilized to quantify different categories of human errors, which have been applied extensively to nuclear power plants. The Tehran research reactor is selected here as a case study, and after consultation with reactor operators and engineers human errors have been identified and adequate performance shaping factors assigned in order to calculate accurate probabilities of human failure.

Ramin Barati; Saeed Setayeshi

2014-01-01T23:59:59.000Z

138

RERTR Program: goals, progress and plans. [Reduced Enrichment Research and Test Reactor  

SciTech Connect (OSTI)

The status of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the nearly null value of 1982 to the 7.0 g U/cm/sup 3/ which will be reached in early 1989. The technical needs of research reactors for HEU exports are also estimated to undergo a gradual but dramatic decline in the coming years.

Travelli, A.

1984-09-25T23:59:59.000Z

139

MYRRHA a multi-purpose hybrid research reactor for high-tech applications  

SciTech Connect (OSTI)

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental accelerator driven system (ADS) in development at SCK-CEN. MYRRHA is able to work both in subcritical (ADS) as in critical mode. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for generation IV (GEN IV) systems, material developments for fusion reactors, radioisotope production and industrial applications, such as Si-doping. MYRRHA will also demonstrate the ADS full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. MYRRHA is based on the heavy liquid metal technology and so it will contribute to the development of lead fast reactor (LFR) technology and in critical mode, MYRRHA will play the role of European technology pilot plant in the roadmap for LFR. In this paper the historical evolution of MYRRHA and the rationale behind the design choices is presented and the latest configuration of the reactor core and primary system is described. (authors)

Abderrahim, H. A.; Baeten, P. [SCK CEN, Boeretang 200, 2400 Mol (Belgium)

2012-07-01T23:59:59.000Z

140

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

SciTech Connect (OSTI)

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

Not Available

1994-04-01T23:59:59.000Z

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


141

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

SciTech Connect (OSTI)

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.

Vinson, D.

2010-07-11T23:59:59.000Z

142

Research Reactors and Radiation Facilities for Joint Use Program  

E-Print Network [OSTI]

Observatory, a facility of the Graduate School of Science of Kyoto University, was constructed in 1968's current main research target is to unveil the origin of the solar magnetic activities that govern science and technology, material science, radiation life science, and radiation medical science

Takada, Shoji

143

Analysis of ITU TRIGA Mark II research reactor using Monte Carlo method  

Science Journals Connector (OSTI)

Abstract Research reactors include many complicated components with various shapes and sizes. Such complex parts also observed in TRIGA core are modelled by the researchers as simplified physical geometries when a particle transport computer code is used to analyse the reactors. These models are used to gain information on possible modifications in the reactors with no cost except a certain computational time demand. Besides, they can be used to understand the fabrication uncertainties of the core components and the methodologies used in the design process. The main objective of this study is to make a detailed three-dimensional full-core model of ITU (Istanbul Technical University) TRIGA Mark II research reactor for the use of Monte Carlo method and making a comparison of the simulation with the experimental observations. In case of lacking of experimental values reported, Final Safety Analysis Report values are used as reference. Furthermore, it is aimed to observe possible influences of using various neutron cross-section libraries (ENDF/Bs and JEFFs) onto the simulation results. The Monte Carlo simulations are carried out by using MCNP5 radiation transport code. All unsteady conditions are ignored, assuming the reactor operates at cold-zero power under the steady-state condition. For comparison, effective core multiplication factor (keff) and effective delayed neutron fraction (?eff) are computed. Reactivity worth ($) of control rods with rod position is presented. Pin power distribution within the fuel elements, axial power peaking distribution along the fuel length and normalized distribution of fast/thermal neutron flux throughout the core are analysed. The simulation results show that MCNP5 model of the reactor is properly established with sufficient detail in such a way that all simulation results are in an excellent agreement with the experimental data (or FSAR values). Results also show that the model yields more or less the same value even different neutron libraries are used.

Mehmet Trkmen; ner olak

2014-01-01T23:59:59.000Z

144

Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors` spent fuel  

SciTech Connect (OSTI)

One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE`s Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE`s efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE`s activities in taking back spent fuel.

NONE

1994-03-25T23:59:59.000Z

145

Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria  

SciTech Connect (OSTI)

The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

K. J. Allen; T. G. Apostolov; I. S. Dimitrov

2009-03-01T23:59:59.000Z

146

Research helps safeguard nuclear workers worldwide - Argonne's Historical  

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

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

147

Experience with the operation, maintenance and utilisation of the 3 MW TRIGA Mark-II research reactor of Bangladesh  

Science Journals Connector (OSTI)

The 3 MW TRIGA (Training, Research, Isotope, General Atomics) Mark-II research reactor of the Bangladesh Atomic Energy Commission (BAEC) has been operating at Atomic Energy Research Establishment (AERE), Savar, Dhaka, since September 1986. Since its commissioning, the reactor has been used in various fields of research and utilisation, such as Neutron Activation Analysis (NAA), Neutron Radiography (NRG), Neutron Scattering (NS), manpower training and education, and production of radioisotopes for medical applications. The reactor facility encountered a couple of incidents, which were successfully handled by BAEC personnel. In some cases, the help of experts from various local organisations/institutions as well as from the International Atomic Energy Agency (IAEA) was obtained. The upgrading of the Safety Analysis Report (SAR) of the reactor facility was completed in 2005 as per the format of the IAEA Safety Guide, SG-35-G1. The cooling system of the reactor as well as some parts of the instrumentations used in the reactor systems were also upgraded/modified during this period. The paper highlights the experience with the operation, maintenance and utilisation of the research reactor for the last 21 years. It also presents some of the modification and upgrading works carried out to enhance the operational safety of the research reactor.

M.A. Zulquarnain; M.M. Haque; M.A. Salam; M.S. Islam; P.K. Saha; M.A. Sarder; A. Haque; M.A.M. Soner; M.M. Uddin; M.M. Rahman; I. Kamal; M.N. Islam; S.M. Hossain

2009-01-01T23:59:59.000Z

148

A one-group parametric sensitivity analysis for the graphite isotope ratio method and other related techniques using ORIGEN 2.2  

E-Print Network [OSTI]

Several methods have been developed previously for estimating cumulative energy production and plutonium production from graphite-moderated reactors. The Graphite Isotope Ratio Method (GIRM) is one well-known technique. This method is based...

Chesson, Kristin Elaine

2009-06-02T23:59:59.000Z

149

Strategic Plan for Light Water Reactor Research and Development  

SciTech Connect (OSTI)

The purpose of this strategic plan is to establish a framework that will allow the Department of Energy (DOE) and the nuclear power industry to jointly plan the nuclear energy research and development (R&D) agenda important to achieving the Nation's energy goals. This strategic plan has been developed to focus on only those R&D areas that will benefit from a coordinated government/industry effort. Specifically, this plan focuses on safely sustaining and expanding the electricity output from currently operating nuclear power plants and expanding nuclear capacity through the deployment of new plants. By focusing on R&D that addresses the needs of both current and future nuclear plants, DOE and industry will be able to take advantage of the synergism between these two technology areas, thus improving coordination, enhancing efficiency, and further leveraging public and private sector resources. By working together under the framework of this strategic plan, DOE and the nuclear industry reinforce their joint commitment to the future use of nuclear power and the National Energy Policy's goal of expanding its use in the United States. The undersigned believe that a public-private partnership approach is the most efficient and effective way to develop and transfer new technologies to the marketplace to achieve this goal. This Strategic Plan is intended to be a living document that will be updated annually.

None

2004-02-01T23:59:59.000Z

150

Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research  

SciTech Connect (OSTI)

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue Universitys Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called Users Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. Users week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

John Jackson; Todd Allen; Frances Marshall; Jim Cole

2013-03-01T23:59:59.000Z

151

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

SciTech Connect (OSTI)

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.

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

1982-03-01T23:59:59.000Z

152

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

E-Print Network [OSTI]

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.

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

2012-05-15T23:59:59.000Z

153

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)  

SciTech Connect (OSTI)

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

Yamada, K. [Vienna International Centre, P.O. Box 100, 1400 Vienna (Austria); Aksan, S. N. [International Atomic Energy Agency, 1400 Vienna (Austria)

2012-07-01T23:59:59.000Z

154

DOE/EIS-0218-SA-3: Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program (November 2004)  

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

SUPPLEMENT ANALYSIS FOR THE FOREIGN SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM NOVEMBER 2004 DOE/EIS-0218-SA-3 U.S. Department of Energy National Nuclear Security Administration Washington, DC Final Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program Final i TABLE OF CONTENTS Page 1. Introduction.............................................................................................................................................. 1 2. Background .............................................................................................................................................. 1 3. The Proposed Action ...............................................................................................................................

155

The RERTR (Reduced Enrichment Research and Test Reactor) program: A progress report  

SciTech Connect (OSTI)

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1985, the activities, results, and new developments which occurred in 1986 are reviewed. The second miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was expanded and its irradiation continued. Postirradiation examinations of several of these miniplates and of six previously irradiated U/sub 3/Si/sub 2/-Al full-size elements were completed with excellent results. The whole-core ORR demonstration with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/ is well under way and due for completion before the end of 1987. DOE removed an important barrier to conversions by announcing that the new LEU fuels will be accepted for reprocessing. New DOE prices for enrichment and reprocessing services were calculated to have minimal effect on HEU reactors, and to reduce by about 8 to 10% the total fuel cycle costs of LEU reactors. New program activities include preliminary feasibility studies of LEU use in DOE reactors, evaluation of the feasibility to use LEU targets for the production of fission-product /sup 99/Mo, and responsibility for coordinating safety evaluations related to LEU conversions of US university reactors, as required by NRC. Achievement of the final program goals is projected for 1990. This progress could not have been achieved without close international cooperation, whose continuation and intensification are essential to the achievement of the ultimate goals of the RERTR Program.

Travelli, A.

1986-11-01T23:59:59.000Z

156

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting  

SciTech Connect (OSTI)

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

NONE

2008-07-15T23:59:59.000Z

157

Research and Development Roadmaps for Nondestructive Evaluation of Cables, Concrete, Reactor Pressure Vessels, and Piping Fatique  

SciTech Connect (OSTI)

To address these research needs, the MAaD Pathway supported a series of workshops in the summer of 2012 for the purpose of developing R&D roadmaps for enhancing the use of Nondestructive Evaluation (NDE) technologies and methodologies for detecting aging and degradation of materials and predicting the remaining useful life. The workshops were conducted to assess requirements and technical gaps related to applications of NDE for cables, concrete, reactor pressure vessels (RPV), and piping fatigue for extended reactor life. An overview of the outcomes of the workshops is presented here. Details of the workshop outcomes and proposed R&D also are available in the R&D roadmap documents cited in the bibliography and are available on the LWRS Program website (http://www.inl.gov/lwrs).

Clayton, Dwight A [ORNL] [ORNL; Bakhtiari, Sasan [Argonne National Laboratory (ANL)] [Argonne National Laboratory (ANL); Smith, Cyrus M [ORNL] [ORNL; Simmons, Kevin [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Coble, Jamie [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Brenchley, David [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Meyer, Ryan [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL)

2013-01-01T23:59:59.000Z

158

ORNL/TM-2012/380 Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

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

2/380 2/380 Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program September 2012 Prepared by Cyrus Smith Randy Nanstad Robert Odette Dwight Clayton Katie Matlack Pradeep Ramuhalli Glenn Light DOCUMENT AVAILABILITY Reports produced after January 1, 1996, are generally available free via the U.S. Department of Energy (DOE) Information Bridge. Web site http://www.osti.gov/bridge Reports produced before January 1, 1996, may be purchased by members of the public from the following source. National Technical Information Service 5285 Port Royal Road Springfield, VA 22161 Telephone 703-605-6000 (1-800-553-6847) TDD 703-487-4639 Fax 703-605-6900

159

B Reactor | Department of Energy  

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

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

160

JAPAN: Investment in Graphite  

Science Journals Connector (OSTI)

JAPAN: Investment in Graphite ... Union Carbide, attempting entry into Japan's graphite electrode industry, has just had one obstacle removed from its path by the Fair Trade Commission in Tokyo. ... In deciding that the company's proposed joint venture with Nippon Carbon Co. conforms with provision's of Japan's antimonopoly law, FTC has improved the prospects for a new tactic in the often difficult process of foreign direct investment in Japanese production firms. ...

1969-08-25T23:59:59.000Z

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


161

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

SciTech Connect (OSTI)

Disordering mechanisms in graphite have a long history with conflicting viewpoints. Using Raman and x-ray photon spectroscoy, 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.

Eapen, Jacob; Murty, Korukonda; Burchell, Timothy

2014-06-02T23:59:59.000Z

162

International Atomic Energy Agency support of research reactor highly enriched uranium to low enriched uranium fuel conversion projects  

SciTech Connect (OSTI)

The IAEA has been involved for more than twenty years in supporting international nuclear non- proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly assisted efforts to convert research reactors from HEU to LEU fuel. HEU to LEU fuel conversion projects differ significantly depending on several factors including the design of the reactor and fuel, technical needs of the member state, local nuclear infrastructure, and available resources. To support such diverse endeavours, the IAEA tailors each project to address the relevant constraints. This paper presents the different approaches taken by the IAEA to address the diverse challenges involved in research reactor HEU to LEU fuel conversion projects. Examples of conversion related projects in different Member States are fully detailed. (author)

Bradley, E.; Adelfang, P.; Goldman, I.N. [Research Reactors Unit, Division of Nuclear Fuel Cycle and Waste Technology, International Atomic Energy Agency, Wagramer Strasse 5, P.O. Box 100, A-1400 Vienna (Austria)

2008-07-15T23:59:59.000Z

163

Current status of the development of high density LEU fuel for Russian research reactors  

SciTech Connect (OSTI)

One of the main directions of the Russian RERTR program is to develop U-Mo fuel and fuel elements/FA with this fuel. The development is carried out both for existing reactors, and for new advanced designs of reactors. Many organizations in Russia, i.e. 'TVEL', RDIPE, RIAR, IRM, NPCC participate in the work. Two fuels are under development: dispersion and monolithic U-Mo fuel, as well two types of FA to use the dispersion U-Mo fuel: with tubular type fuel elements and with pin type fuel elements. The first stage of works was successfully completed. This stage included out-pile, in-pile and post irradiation examinations of U-Mo dispersion fuel in experimental tubular and pin fuel elements under parameters similar to operation conditions of Russian design pool-type research reactors. The results received both in Russia and abroad enabled to go on to the next stage of development which includes irradiation tests both of full-scale IRT pin-type and tube-type fuel assemblies with U-Mo dispersion fuel and of mini-fuel elements with modified U-Mo dispersion fuel and monolithic fuel. The paper gives a generalized review of the results of U-Mo fuel development accomplished by now. (author)

Vatulin, A.; Dobrikova, I.; Suprun, V.; Trifonov, Y. [Federal State Unitary Enterprise, A.A. Bochvar All-Russian Scientific Research Institute of Inorganic Materials (VNIINM), 123060 Rogov 5a, Moscow (Russian Federation); Kartashev, E.; Lukichev, V. [Federal State Unitary Enterprise RDIPE, 101000 P.O. Box 788, Moscow (Russian Federation)

2008-07-15T23:59:59.000Z

164

Performance characteristics of the annular core research reactor fuel motion detection system  

SciTech Connect (OSTI)

Recent proof tests have shown that the annular core research reactor (ACRR) fuel motion detection system has reached its design goals of providing high temporal and spatial resolution pictures of fuel distributions in the ACRR. The coded aperture imaging system (CAIS) images the fuel by monitoring the fission gamma rays from the fuel that pass through collimators in the reactor core. The gamma-ray beam is modulated by coded apertures before producing a visible light coded image in thin scintillators. Each coded image is then amplified and recorded by an opticalimage-intensifier/fast-framing-camera combination. The proximity to the core and the coded aperture technique provide a high data collection rate and high resolution. Experiments of CAIS at the ACRR conducted under steady-state operation have documented the beneficial effects of changes in the radiation shielding and imaging technique. The spatial resolutions are 1.7 mm perpendicular to the axis of a single liquid-metal fast breeder reactor fuel pin and 9 mm in the axial dimension. Changes in mass of 100 mg in each resolution element can be detected each frame period, which may be from 5 to 100 ms. This diagnostic instrument may help resolve important questions in fuel motion phenomenology.

Kelly, J.G.; Stalker, K.T.

1983-12-01T23:59:59.000Z

165

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

Science Journals Connector (OSTI)

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

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

2014-10-01T23:59:59.000Z

166

Progress of the RERTR (Reduced Enrichment Research and Test Reactor) Program in 1989  

SciTech Connect (OSTI)

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1988, the major events, findings, and activities of 1989 are reviewed. The scope of the RERTR Program activities was curtailed, in 1989, by an unexpected legislative restriction which limited the ability of the Arms Control and Disarmament Agency to adequately fund the program. Nevertheless, the thrust of the major planned program activities was maintained, and meaningful results were obtained in several areas of great significance for future work. 15 refs., 12 figs.

Travelli, A.

1989-01-01T23:59:59.000Z

167

E-Print Network 3.0 - advanced reactor research Sample Search...  

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

to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, near the...

168

E-Print Network 3.0 - advanced research reactor Sample Search...  

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

to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, near the...

169

E-Print Network 3.0 - austrian research reactor Sample Search...  

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

to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, near the...

170

E-Print Network 3.0 - anuclear research reactor Sample Search...  

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

to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, near the...

171

Status of the RERTR (Reduced Enrichment Research and Test Reactor) Program  

SciTech Connect (OSTI)

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1987, the major events, findings and activities of 1988 are reviewed. The US Nuclear Regulatory Commission issued a formal and generic approval of the use of U3Si2-Al dispersion fuel in research and test reactors, with densities up to 4.8 g U/cmT. New significant findings from postirradiation examinations, from ion-beam irradiations, and from analytical modeling, have raised serious doubts about the potential of LEU U3Si-Al dispersion fuel for applications requiring very high uranium densities and high burnups (>6 g U/cmT, >50% burnup). As a result of these findings, the fuel development efforts have been redirected towards three new initiatives: (1) a systematic application of ion-beam irradiations to screen new materials; (2) application of Hot Isostatic Pressing (HIP) procedures to produce U3Si2-Al plates with high uranium densities and thin uniform cladding; and (3) application of HIP procedures to produce plates with U3Si wires imbedded in an aluminum matrix, achieving stability, high uranium density, and thin uniform cladding. The new fuel concepts hold the promise of extraordinary performance potential and require approximately five years to develop.

Travelli, A.

1988-01-01T23:59:59.000Z

172

Improved graphite furnace atomizer  

DOE Patents [OSTI]

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.

Siemer, D.D.

1983-05-18T23:59:59.000Z

173

Diamond-graphite field emitters  

DOE Patents [OSTI]

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

Valone, Steven M. (Santa Fe, NM)

1997-01-01T23:59:59.000Z

174

Graphite-based photovoltaic cells  

DOE Patents [OSTI]

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.

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

2010-12-28T23:59:59.000Z

175

Record of Decision for the Final EIS on Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel  

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

5091 5091 Friday May 17, 1996 Part IV Department of Energy Record of Decision for the Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel; Notice 25092 Federal Register / Vol. 61, No. 97 / Friday, May 17, 1996 / Notices DEPARTMENT OF ENERGY Record of Decision for the Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel AGENCY: Department of Energy. ACTION: Record of decision. SUMMARY: DOE, in consultation with the Department of State, has decided to implement a new foreign research reactor spent fuel acceptance policy as specified in the Preferred Alternative contained in the Final Environmental Impact Statement on a Proposed

176

The RERTR (Reduced Enrichment Research and Test Reactor) Program: Progress and plans  

SciTech Connect (OSTI)

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results, and new developments which occurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/ was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40% average burnup. Good progress was made in the area of LEU usage for the production of fission /sup 99/Mo, and in the coordination of safety evaluations related to LEU conversions of US university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U/sub 3/Si-Al with 19.75% enrichment and U/sub 3/Si/sub 2/-Al with 45% enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR Program.

Travelli, A.

1987-01-01T23:59:59.000Z

177

Raman spectroscopy of graphite  

Science Journals Connector (OSTI)

...G for graphite. The other modes are either observed only on defective samples or are very weak in intensity like the G peak that was...al. 2001); a similar behaviour is also observed in other car- bon materials (Ferrari & Robertson 2001; Maultzsch et al...

2004-01-01T23:59:59.000Z

178

LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR  

SciTech Connect (OSTI)

Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J

1981-02-01T23:59:59.000Z

179

NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669  

SciTech Connect (OSTI)

The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document.'' This SER provides 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.

Not Available

1992-08-01T23:59:59.000Z

180

A computer model for the transient analysis of compact research reactors with plate type fuel  

SciTech Connect (OSTI)

A coupled neutronics and core thermal-hydraulic performance model has been developed for the analysis of plate type U-Al fueled high-flux research reactor transients. The model includes point neutron kinetics, one-dimensional, non-homogeneous, equilibrium two-phase flow and beat transfer with provision for subcooled boiling, and spatially averaged one-dimensional beat conduction. The feedback from core regions other than the fuel elements is included by employing a lumped parameter approach. Partial differential equations are discretized in space and the combined equation set representing the model is converted to an initial value problem. A variable-order, variable-time-step time advancement scheme is used to solve these ordinary differential equations. The model is verified through comparisons with two other computer code results and partially validated against SPERT-II tests. It is also used to analyze a series of HFIR reactivity transients.

Sofu, T. [Argonne National Lab., IL (United States); Dodds, H.L. [Tennessee Univ., Knoxville, TN (United States). Dept. of Nuclear Engineering

1994-03-01T23:59:59.000Z

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


181

Final Site Specific Decommissioning Inspection Report #2 for the University of Washington Research and Test Reactor, Seattle, Washington  

SciTech Connect (OSTI)

During the period of August through November 2006, ORISE performed a comprehensive IV at the University of Washington Research and Test Reactor Facility. The objective of the ORISE IV was to validate the licensees final status survey processes and data, and to assure the requirements of the DP and FSSP were met.

S.J. Roberts

2007-03-20T23:59:59.000Z

182

Estimation of graphite density and mechanical strength variation of VHTR during air-ingress accident  

SciTech Connect (OSTI)

An air-ingress accident in a Very High Temperature Gas-Cooled Reactor (VHTR) is anticipated to cause severe changes to graphite density and mechanical strength by an oxidation process that has many side effects. However, quantitative estimations have not yet been performed. This study focuses on predicting the changes in graphite density and mechanical strength via thermal hydraulic system analysis code. In order to analyze the change in graphite density, a simple graphite burn-off model was developed. The model is based on the similarities between a parallel electrical circuit and graphite oxidation. It was used to determine overall changes in the graphites geometry and density. The model was validated by comparing its results to experimental data that was obtained for several temperatures. In the experiment, cylindrically shaped graphite specimens were oxidized in an electrical furnace and the variations of its mass were measured against time. The experiments range covered temperatures between 6000C and 9000 C. Experimental data validated the models accuracy. Finally, the developed model along with other comprehensive graphite oxidation models was integrated into the VHTR system analysis code, GAMMA. GT-MHR 600 MWt reactor was selected as a reference reactor. Based on the calculation, the main oxidation process was observed 5.5 days after the accident when followed by natural convection. The core maximum temperature reached 16000 C, but never exceeded the maximum temperature criteria, 18000 C. However, the oxidation process did significantly decrease the density of bottom reflector, making it vulnerable to mechanical stress. The stress on the bottom reflector is greatly increased because it sustains the reactor core. The calculation proceeded until 11 days after the accident, resulting in an observed 4.5% decrease in density and a 25% reduction of mechanical strength.

Eung Soo Kim

2008-04-01T23:59:59.000Z

183

Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 5: Graphite PIRTs  

SciTech Connect (OSTI)

Here we report the outcome of the application of the Nuclear Regulatory Commission (NRC) Phenomena Identification and Ranking Table (PIRT) process to the issue of nuclear-grade graphite for the moderator and structural components of a next generation nuclear plant (NGNP), considering both routine (normal operation) and postulated accident conditions for the NGNP. The NGNP is assumed to be a modular high-temperature gas-cooled reactor (HTGR), either a gas-turbine modular helium reactor (GTMHR) version [a prismatic-core modular reactor (PMR)] or a pebble-bed modular reactor (PBMR) version [a pebble bed reactor (PBR)] design, with either a direct- or indirect-cycle gas turbine (Brayton cycle) system for electric power production, and an indirect-cycle component for hydrogen production. NGNP design options with a high-pressure steam generator (Rankine cycle) in the primary loop are not considered in this PIRT. This graphite PIRT was conducted in parallel with four other NRC PIRT activities, taking advantage of the relationships and overlaps in subject matter. The graphite PIRT panel identified numerous phenomena, five of which were ranked high importance-low knowledge. A further nine were ranked with high importance and medium knowledge rank. Two phenomena were ranked with medium importance and low knowledge, and a further 14 were ranked medium importance and medium knowledge rank. The last 12 phenomena were ranked with low importance and high knowledge rank (or similar combinations suggesting they have low priority). The ranking/scoring rationale for the reported graphite phenomena is discussed. Much has been learned about the behavior of graphite in reactor environments in the 60-plus years since the first graphite rectors went into service. The extensive list of references in the Bibliography is plainly testament to this fact. Our current knowledge base is well developed. Although data are lacking for the specific grades being considered for Generation IV (Gen IV) concepts, such as the NGNP, it is fully expected that the behavior of these graphites will conform to the recognized trends for near isotropic nuclear graphite. Thus, much of the data needed is confirmatory in nature. Theories that can explain graphite behavior have been postulated and, in many cases, shown to represent experimental data well. However, these theories need to be tested against data for the new graphites and extended to higher neutron doses and temperatures pertinent to the new Gen IV reactor concepts. It is anticipated that current and planned future graphite irradiation experiments will provide the data needed to validate many of the currently accepted models, as well as providing the needed data for design confirmation.

Burchell, Timothy D [ORNL; Bratton, Rob [Idaho National Laboratory (INL); Marsden, Barry [University of Manchester, UK; Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission; Penfield, Scott [Technology Insights; Mitchell, Mark [PBMR (Pty) Ltd.; Windes, Will [Idaho National Laboratory (INL)

2008-03-01T23:59:59.000Z

184

Hydro-mechanical analysis of low enriched uranium fuel plates for University of Missouri Research Reactor .  

E-Print Network [OSTI]

??As part of the Global Threat Reduction Initiative (GTRI) Reactor Conversion program, work is underway to analyze and validate a new fuel assembly for the (more)

Kennedy, John C.

2012-01-01T23:59:59.000Z

185

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

E-Print Network [OSTI]

??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 (more)

Connaway, Heather M. (Heather Moira)

2012-01-01T23:59:59.000Z

187

Radiation Exposures Associated with Shipments of Foreign Research Reactor Spent Nuclear Fuel  

SciTech Connect (OSTI)

Experience has shown that the analyses of marine transport of spent fuel in the Environmental Impact Statement (EIS) were conservative. It is anticipated that for most shipments. The external dose rate for the loaded transportation cask will be more in line with recent shipments. At the radiation levels associated with these shipments, we would not expect any personnel to exceed radiation exposure limits for the public. Package dose rates usually well below the regulatory limits and personnel work practices following ALARA principles are keeping human exposures to minimal levels. However, the potential for Mure shipments with external dose rates closer to the exclusive-use regulatory limit suggests that DOE should continue to provide a means to assure that individual crew members do not receive doses in excess of the public dose limits. As a minimum, the program will monitor cask dose rates and continue to implement administrative procedures that will maintain records of the dose rates associated with each shipment, the vessel used, and the crew list for the vessel. DOE will continue to include a clause in the contract for shipment of the foreign research reactor spent nuclear fuel requiring that the Mitigation Action Plan be followed.

MASSEY,CHARLES D.; MESSICK,C.E.; MUSTIN,T.

1999-11-01T23:59:59.000Z

188

Status of the RERTR program: overview, progress and plans. [Reduced Enrighment Research and Test Reactor  

SciTech Connect (OSTI)

The status of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a summary of the accomplishments which the RERTR Program had achieved by the end of 1984 with its many international partners, emphasis is placed on the progress achieved during 1985 and on current plans and schedules. A new miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was fabricated and is well into irradiation. The whole-core ORR demonstration is scheduled to begin in November 1985, with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/. Altogether, 921 full-size test and prototype elements have been ordered for fabrication with reduced enrichment and the new technologies. Qualification of U/sub 3/Si-Al fuel with approx.7 g U/cm/sup 3/ is still projected for 1989. This progress could not have been achieved without the close international cooperation which has existed since the beginning, and whose continuation and intensification will be essential to the achievement of the long-term RERTR goals.

Travelli, A.

1985-01-01T23:59:59.000Z

189

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

SciTech Connect (OSTI)

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.

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

1994-10-01T23:59:59.000Z

190

Effect of microstructure on air oxidation resistance of nuclear graphite  

SciTech Connect (OSTI)

Oxidation resistance in air of three grades of nuclear graphite with different structures was compared using a standard thermogravimetric method. Differences in the oxidation behavior have been identified with respect to both (i) the rate of oxidation in identical conditions and the derived apparent activation energy and pre-exponential factor and (ii) the penetration depth of the oxidant and the development of the oxidized layer. These differences were ascribed to structural differences between the three graphite grades, in particular the grain size and shape of the graphite filler, and the associated textural properties, such as total BET surface area and porosity distribution in the un-oxidized material. It was also found that the amount of strongly bonded surface oxygen complexes measured by thermodesorption significantly exceeds the amount afforded by the low BET surface area, and therefore low temperature oxygen chemisorption is not a reliable method for determining the amount of surface sites (re)active during air oxidation. The relationship between nuclear graphite microstructure and its oxidation resistance demonstrated in this work underlines the importance of performing comprehensive oxidation characterization studies of the new grades of nuclear graphite considered as candidates for very high temperature gas-cooled reactors.

Contescu, Cristian I [ORNL; Guldan, Tyler R [ORNL; Wang, Peng [ORNL; Burchell, Timothy D [ORNL

2012-01-01T23:59:59.000Z

191

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

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

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

192

Novel, Magnetically Fluidized-Bed Reactor Development for the Looping Process: Coal to Hydrogen Production Research and Development  

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

Novel, Magnetically Fluidized-Bed Novel, Magnetically Fluidized-Bed Reactor Development for the Looping Process: Coal to Hydrogen Production Research and Development Background The U.S. Department of Energy (DOE) National Energy Technology Laboratory (NETL) is committed to improving methods for co-producing power and chemicals, fuels, and hydrogen (H2). Gasification is a process by which fuels such as coal can be used to produce synthesis gas (syngas), a mixture of H2, carbon monoxide (CO), and carbon

193

Research and Development of High Temperature Light Water Cooled Reactor Operating at Supercritical-Pressure in Japan  

SciTech Connect (OSTI)

This paper summarizes the status and future plans of research and development of the high temperature light water cooled reactor operating at supercritical-pressure in Japan. It includes; the concept development; material for the fuel cladding; water chemistry under supercritical pressure; thermal hydraulics of supercritical fluid; and the conceptual design of core and plant system. Elements of concept development of the once-through coolant cycle reactor are described, which consists of fuel, core, reactor and plant system, stability and safety. Material studies include corrosion tests with supercritical water loops and simulated irradiation tests using a high-energy transmission electron microscope. Possibilities of oxide dispersion strengthening steels for the cladding material are studied. The water chemistry research includes radiolysis and kinetics of supercritical pressure water, influence of radiolysis and radiation damage on corrosion and behavior on the interface between water and material. The thermal hydraulic research includes heat transfer tests of single tube, single rod and three-rod bundles with a supercritical Freon loop and numerical simulations. The conceptual designs include core design with a three-dimensional core simulator and sub-channel analysis, and balance of plant. (authors)

Yoshiaki Oka [Nuclear Engineering Research Laboratory, The University of Tokyo, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 112-0006 (Japan); Katsumi Yamada [Isogo Nuclear Engineering Center, Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan)

2004-07-01T23:59:59.000Z

194

Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor  

E-Print Network [OSTI]

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 ...

Bean, Malcolm K.

2011-08-01T23:59:59.000Z

195

Analysis of a research reactor under anticipated transients without scram events using the RELAP5/MOD3.2 computer program  

E-Print Network [OSTI]

Simulations for two series of anticipated transients phics. without scram (ATWS) events have been carried out for a small, hypothetical, research reactor based on the High Flux Australian Reador HIFAR using the RELAPS/MOD3.Z computer program...

Hari, Sridhar

1998-01-01T23:59:59.000Z

196

Expanding and optimizing fuel management and data analysis capabilities of MCODE-FM in support of MIT research reactor (MITR-II) LEU conversion  

E-Print Network [OSTI]

Studies are underway in support of the MIT research reactor (MITR-II) conversion from high enriched Uranium (HEU) to low enriched Uranium (LEU), as required by recent non-proliferation policy. With the same core configuration ...

Horelik, Nicholas E. (Nicholas Edward)

2012-01-01T23:59:59.000Z

197

Graphitized needle cokes and natural graphites for lithium intercalation  

SciTech Connect (OSTI)

This paper examined effects of heat treatment and milling (before or after heat treatment) on the (electrochemical) intercalating ability of needle petroleum coke; natural graphite particles are included for comparison. 1 tab, 4 figs, 7 refs.

Tran, T.D.; Spellman, L.M.; Pekala, R.W. [Lawrence Livermore National Lab., CA (United States); Goldberger, W.M. [Superior Graphite Co., Chicago, IL (United States); Kinoshita, K. [Lawrence Berkeley National Lab., CA (United States)

1996-05-10T23:59:59.000Z

198

Chemical kinetics parameters of nuclear graphite gasification  

Science Journals Connector (OSTI)

This paper provides chemical kinetics parameters for the gasification of nuclear graphite grades of IG-110, IG-430, NBG-18 and NBG-25 and presents empirical correlations for their surface areas of free active sites as a function of mass. The kinetics parameters for the four elementary chemical reactions of gasification of these grades of nuclear graphite include the values and Gaussian distributions of the specific activation energies and the values of the pre-exponential rate coefficients for the adsorption of oxygen and desorption of CO and CO2 gases. The values of these parameters and the surface area of free active sites for IG-110 and NB-25, with fine and medium petroleum coke filler particles, are nearly the same, but slightly different from those of NBG-18 and IG-430, with medium and fine coal tar pitch coke filler particles. Recommended parameters are applicable to future safety analysis of high and very high temperature gas cooled reactors in the unlikely event of a massive air ingress accident.

Mohamed S. El-Genk; Jean-Michel P. Tournier

2013-01-01T23:59:59.000Z

199

Status of axial heterogeneous liquid-metal fast breeder reactor core design studies and research and development  

SciTech Connect (OSTI)

The current status of axial heterogeneous core (AHC) design development in Japan, which consists of an AHC core design in a pool-type demonstration fast breeder reactor (DFBR) and research and development activities supporting AHC core design, is presented. The DFBR core design objectives developed by The Japan Atomic Power Company include (a) favorable core seismic response, (b) core compactness, (c) high availability, and (d) lower fuel cycle cost. The AHC concept was selected as a reference pool-type DFBR core because it met these objectives more suitably than the homogeneous core (HOC). The AHC core layouts were optimized emphasizing the reduction of the burnup reactivity swing, peak fast fluence, and power peaking. The key performance parameters resulting from the AHC, such as flat axial power/flux distribution, lower peak fast fluence, lower burnup reactivity swing, etc., were evaluated in comparison with the HOC. The critical experiments at the Japan Atomic Energy Research Institute's Fast Critical Assembly facility demonstrate the key AHC performance characteristics. The large AHC engineering benchmark experiments using the zero-power plutonium reactor and the AHC fuel pin irradiation test program using the JOYO reactor are also presented.

Nakagawa, H.; Inagaki, T.; Yoshimi, H.; Shirakata, K.; Watari, Y.; Suzuki, M.; Inoue, K.

1988-11-01T23:59:59.000Z

200

Using Graphite to view network data  

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

Graphite Graphite to Visualize Network Data Jon Dugan Summer ESCC 2010, Columbus, OH Lawrence Berkeley National Laboratory U.S. Department of Energy | Office of Science ESnet Statistics Overview ESxSNMP (Data Collection) ESxSNMP (Data Collection) Graphite (Visualization) Graphite (Visualization) Analytics (Custom Reports) Analytics (Custom Reports) Net Almanac (Metadata) Net Almanac (Metadata) Lawrence Berkeley National Laboratory U.S. Department of Energy | Office of Science What is Graphite? "Enterprise scalable realtime graphing" * Developed by Orbitz for visualizing internal performance data * Open source: https://launchpad.net/graphite * Has own RRD like database called Carbon * RRD Compatible ESxSNMP Integration * via REST interface * Easy integration, Graphite is well written

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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201

EXPERIMENT OPERATIONS PLAN FOR A LOSS-OF-COOLANT ACCIDENT SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR  

SciTech Connect (OSTI)

Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship between the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. This document contains both experiment proposal and assembly proposal information. The intent of this document is to supply information required by the Chalk River Nuclear Laboratories (CRNL), and to identify the planned procedures and data that will be used both to establish readiness to proceed from one test phase to the next and to operate the experiment. Operating control settings and limits are provided for both experimenter systems and CRNL systems. A hazards review summarizes safety issues that have been addressed during the development of the experiment plan.

Russcher, G. E.; Cannon, L. W.; Goodman, R. L.; Hesson, G. M.; King, L. L.; McDuffie, P. N.; Marshall, R. K.; Nealley, C.; Pilger, J. P.; Mohr, C. L.

1981-04-01T23:59:59.000Z

202

Hydrogen Storage in Graphite Nanofibers  

Science Journals Connector (OSTI)

Hydrogen Storage in Graphite Nanofibers ... Subsequent lowering of the pressure to nearly atmospheric conditions results in the release of a major fraction of the stored hydrogen at room temperature. ...

Alan Chambers; Colin Park; R. Terry K. Baker; Nelly M. Rodriguez

1998-05-12T23:59:59.000Z

203

Baseline Concept Description of a Small Modular High Temperature Reactor  

SciTech Connect (OSTI)

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 DOEs 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.

Hans Gougar

2014-05-01T23:59:59.000Z

204

DOE Drops Plan to Restart Reactor  

Science Journals Connector (OSTI)

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

Robert F. Service

2000-12-01T23:59:59.000Z

205

Examples of the use of PSA in the design process and to support modifications at two research reactors  

SciTech Connect (OSTI)

Many, if not most, of the world`s commercial nuclear power plants have been the subject of plant-specific probabilistic safety assessments (PSA). A growing number of other nuclear facilities as well as other types of industrial installations have been the focus of plant-specific PSAs. Such studies have provided valuable information concerning the nature of the risk of the individual facility and have been used to identify opportunities to manage that risk. This paper explores the risk management activities associated with two research reactors in the United States as a demonstration of the versatility of the use of PSA to support risk-related decision making.

Johnson, D.H.; Bley, D.C.; Lin, J.C. [PLG, Inc., Newport Beach, CA (United States); Ramsey, C.T.; Linn, M.A. [Oak Ridge National Lab., TN (United States)

1994-03-01T23:59:59.000Z

206

Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters  

SciTech Connect (OSTI)

Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies.

Konzek, G.J.

1983-07-01T23:59:59.000Z

207

Experimental evaluation of gamma fluence-rate predictions from Argon-41 releases to the atmosphere over a nuclear research reactor site  

Science Journals Connector (OSTI)

......radiation experimental dataset has been subsequently...reactor at the Belgium Nuclear Research Center...measurements. The dataset obtained may also...3, 4) used for nuclear emergency preparedness...radiation experimental dataset has been subsequently...RIMPUFF. | Belgian Nuclear Research Center......

Carlos Rojas-Palma; Helle Karina Aage; Poul Astrup; Kim Bargholz; Martin Drews; Hans E. Jrgensen; Uffe Korsbech; Bent Lauritzen; Torben Mikkelsen; Sren Thykier-Nielsen; Raf Van Ammel

2004-01-01T23:59:59.000Z

208

Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements  

SciTech Connect (OSTI)

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

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

2011-09-01T23:59:59.000Z

209

First Direct Evidence of Dirac Fermions in Graphite  

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

Direct Evidence of Dirac Fermions in Graphite Print Direct Evidence of Dirac Fermions 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 driven by the discovery that electron mobility in graphene is ten times higher than in commercial-grade silicon, raising the possibility of high-efficiency, low-power, carbon-based electronics. Scientists attribute graphene's surprising current capacity (as well as a number of even stranger phenomena) to the presence of charge carriers that behave as if they are massless, "relativistic" quasiparticles called Dirac fermions. Harnessing these quasiparticles in real-world carbon-based devices, however, requires a deeper knowledge of their behavior under less-than-ideal circumstances, such as around defects, at edges, or in three dimensions-in other words, in graphite. At the ALS, a team of researchers using angle-resolved photoemission spectroscopy (ARPES) have now produced the first direct evidence of massless Dirac fermions in graphite coexisting with quasiparticles of finite effective mass and defect-induced localized states.

210

First Direct Evidence of Dirac Fermions in Graphite  

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

Direct Evidence of Dirac Fermions in Graphite Print Direct Evidence of Dirac Fermions 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 driven by the discovery that electron mobility in graphene is ten times higher than in commercial-grade silicon, raising the possibility of high-efficiency, low-power, carbon-based electronics. Scientists attribute graphene's surprising current capacity (as well as a number of even stranger phenomena) to the presence of charge carriers that behave as if they are massless, "relativistic" quasiparticles called Dirac fermions. Harnessing these quasiparticles in real-world carbon-based devices, however, requires a deeper knowledge of their behavior under less-than-ideal circumstances, such as around defects, at edges, or in three dimensions-in other words, in graphite. At the ALS, a team of researchers using angle-resolved photoemission spectroscopy (ARPES) have now produced the first direct evidence of massless Dirac fermions in graphite coexisting with quasiparticles of finite effective mass and defect-induced localized states.

211

First Direct Evidence of Dirac Fermions in Graphite  

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

Direct Evidence of Dirac Direct Evidence of Dirac Fermions in Graphite First Direct Evidence of Dirac Fermions in Graphite Print Wednesday, 27 June 2007 00:00 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 driven by the discovery that electron mobility in graphene is ten times higher than in commercial-grade silicon, raising the possibility of high-efficiency, low-power, carbon-based electronics. Scientists attribute graphene's surprising current capacity (as well as a number of even stranger phenomena) to the presence of charge carriers that behave as if they are massless, "relativistic" quasiparticles called Dirac fermions. Harnessing these quasiparticles in real-world carbon-based devices, however, requires a deeper knowledge of their behavior under less-than-ideal circumstances, such as around defects, at edges, or in three dimensions-in other words, in graphite. At the ALS, a team of researchers using angle-resolved photoemission spectroscopy (ARPES) have now produced the first direct evidence of massless Dirac fermions in graphite coexisting with quasiparticles of finite effective mass and defect-induced localized states.

212

Report of the ANS Project Feasibility Workshop for a High Flux Isotope Reactor-Center for Neutron Research Facility  

SciTech Connect (OSTI)

The Advanced Neutron Source (ANS) Conceptual Design Report (CDR) and its subsequent updates provided definitive design, cost, and schedule estimates for the entire ANS Project. A recent update to this estimate of the total project cost for this facility was $2.9 billion, as specified in the FY 1996 Congressional data sheet, reflecting a line-item start in FY 1995. In December 1994, ANS management decided to prepare a significantly lower-cost option for a research facility based on ANS which could be considered during FY 1997 budget deliberations if DOE or Congressional planners wished. A cost reduction for ANS of about $1 billion was desired for this new option. It was decided that such a cost reduction could be achieved only by a significant reduction in the ANS research scope and by maximum, cost-effective use of existing High Flux Isotope Reactor (HFIR) and ORNL facilities to minimize the need for new buildings. However, two central missions of the ANS -- neutron scattering research and isotope production-were to be retained. The title selected for this new option was High Flux Isotope Reactor-Center for Neutron Research (HFIR-CNR) because of the project`s maximum use of existing HFIR facilities and retention of selected, central ANS missions. Assuming this shared-facility requirement would necessitate construction work near HFIR, it was specified that HFIR-CNR construction should not disrupt normal operation of HFIR. Additional objectives of the study were that it be highly credible and that any material that might be needed for US Department of Energy (DOE) and Congressional deliberations be produced quickly using minimum project resources. This requirement made it necessary to rely heavily on the ANS design, cost, and schedule baselines. A workshop methodology was selected because assessment of each cost and/or scope-reduction idea required nearly continuous communication among project personnel to ensure that all ramifications of propsed changes.

Peretz, F.J.; Booth, R.S. [comp.

1995-07-01T23:59:59.000Z

213

Graphite dust resuspension in an HTR-10 steam generator  

Science Journals Connector (OSTI)

Abstract Graphite dust has an important effect on the safety of high-temperature gas-cooled reactors (HTR). The flow field in the steam generator was studied by the computational fluid dynamics (CFD) method, with the results indicating that the friction velocity in the windward and the leeward of the heat transfer tubes is relatively low and is higher at the sides. Further analysis of the resuspension of graphite dust indicates that the resuspension fraction reaches nearly zero for particles with a diameter less than 1?m, whereas it will increases as the helium velocity in the steam generator increases for particle size larger than 1?m. Moreover, the resuspension fraction increases as the particle size increases. The results also indicate that resuspension of the particles with sizes larger than 1?m exhibited obvious differences in different parts of the steam generator.

Wei Peng; Tianqi Zhang; Yanan Zhen; Suyuan Yu

2014-01-01T23:59:59.000Z

214

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

SciTech Connect (OSTI)

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.

Mark C. Carroll; David T. Rohrbaugh

2013-08-01T23:59:59.000Z

215

Is Graphite a Diamonds Best Friend? New Information on Material  

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

November 18th, 2003 November 18th, 2003 Is Graphite a Diamond's Best Friend? New Information on Material Transformation Science has yet to achieve the alchemist's dream of turning lead into gold. But a group of re-searchers using the GeoSoilEn-viroCARS (GSECARS) and High-Pressure Collaborative Access Team (HP-CAT) facilities at the Department of Energy's Advanced Photon Source (APS) at Argonne National Laboratory, may have found a way to turn ordinary soft graphite (source of the "lead" found in pencils) into a new, super-hard material that "looks" just like diamond. Using the high-brilliance x-ray beams from the APS, the group discovered that, under extreme pressure, graphite (among the softest of materials and the source of the lead found in pencils) becomes as hard as diamond, the

216

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

SciTech Connect (OSTI)

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)

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

2012-07-01T23:59:59.000Z

217

Calculational-experimental research models for a fast reactor with a heterogeneous core  

SciTech Connect (OSTI)

The physical characteristics of heterogeneous metallic oxide cores were experimentally studied by physical tests of the critical assemblies BFS-46 and BFS-46AZ, which simulate a reactor of the BN-1600 type, into the core of which a fuel assembly with metallic uranium is inserted. A calculational model for the critical assemblies being investigated, showing the zones and their dimensions, is presented. The critical assembly BFS-46AZ is a modification of the basic critical assembly BFS-46 which adds plutonium to the IBZ to simulate its accumulation during reactor operation. The BFS-46 and BFS-46AZ assemblies have identical dimensions for the IBZ and LEZ, and have different HEZ dimensions, necessary to ensure the criticality of each assembly. Plutonium with a /sup 240/Pu content equal to 3.8% is used in the LEZ. The critically parameters are calculated using one-dimensional and two-dimensional models in a 26-group diffusion approximation based on the BNAP-78 system of group constants.

Belov, S.P.; Bobrov, S.B.; Kazanskii, Yu.A.; Kuzin, E.N.; Matveev, V.I.; Novozhilov, A.I.; Chernyi, V.A.

1987-11-01T23:59:59.000Z

218

Irradiation research capabilities at HFIR (High Flux Isotope Reactor) and ANS (Advanced Neutron Source)  

SciTech Connect (OSTI)

A variety of materials irradiation facilities exist in the High Flux Isotope Reactor (HFIR) and are planned for the Advanced Neutron Source (ANS) reactor. In 1986 the HFIR Irradiation Facilities Improvement (HIFI) project began modifications to the HFIR which now permit the operation of two instrumented capsules in the target region and eight capsules of 46-mm OD in the RB region. Thus, it is now possible to perform instrumented irradiation experiments in the highest continuous flux of thermal neutrons available in the western world. The new RB facilities are now large enough to permit neutron spectral tailoring of experiments and the modified method of access to these facilities permit rotation of experiments thereby reducing fluence gradients in specimens. A summary of characteristics of irradiation facilities in HFIR is presented. The ANS is being designed to provide the highest thermal neutron flux for beam facilities in the world. Additional design goals include providing materials irradiation and transplutonium isotope production facilities as good, or better than, HFIR. The reference conceptual core design consists of two annular fuel elements positioned one above the other instead of concentrically as in the HFIR. A variety of materials irradiation facilities with unprecedented fluxes are being incorporated into the design of the ANS. These will include fast neutron irradiation facilities in the central hole of the upper fuel element, epithermal facilities surrounding the lower fuel element, and thermal facilities in the reflector tank. A summary of characteristics of irradiation facilities presently planned for the ANS is presented. 2 tabs.

Thoms, K.R.

1990-01-01T23:59:59.000Z

219

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel  

SciTech Connect (OSTI)

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed.

Deen, J.R.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Papastergiou, C. [National Center for Scientific Research, Athens (Greece)

1992-12-31T23:59:59.000Z

220

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel  

SciTech Connect (OSTI)

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed.

Deen, J.R.; Snelgrove, J.L. (Argonne National Lab., IL (United States)); Papastergiou, C. (National Center for Scientific Research, Athens (Greece))

1992-01-01T23:59:59.000Z

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


221

Groundwater Monitoring and Control Before Decommissioning of the Research Reactor VVR-S from Magurele-Bucharest  

SciTech Connect (OSTI)

The research reactor type VVR-S (tank type, water is cooler, moderator and reflector, thermal power- 2 MW, thermal energy- 9. 52 GW d) was put into service in July 1957 and, in December 1997 was shout down. In 2002, Romanian Government decided to put the research reactor in the permanent shut-down in order to start the decommissioning. This nuclear facility was used in nuclear research and radioisotope production for 40 years, without events, incidents or accidents. Within the same site, in the immediate vicinity of the research reactor, there are many other nuclear facilities: Radioactive Waste Treatment Plant, Tandem Van der Graaf heavy ions accelerator, Cyclotron, Industrial Irradiator, Radioisotope Production Center. The objectives of this work were dedicated on the water underground analyses described in the following context: - presentation of the approaches in planning the number of drillings, vertical soil profiles (characteristics, analyses, direction of the flow of underground water, uncertainties in measurements); - presentation of the instrumentation used in analyses of water, soil and vegetation samples - analyses and final conclusions on results of the measurements; - comparison of the results of measurements on underground water from drillings with the measurements results on samples from the town and the system of drinking water - supplied from the second level of underground water. According to the analysis, in general, no values higher than the Minimum Detectable Activity were detected in water samples (MDA) for Pb{sup 212}, Bi{sup 214}, Pb{sup 214}, Ac{sup 228}, but situated under values foreseen in drinking water. Distribution of Uranium As results of the Uranium determination, values higher than 0,004 mg/l (4 ppb) were detected, values that represent the average contents in the underground water. The higher values, 2-3 times higher than background, were detected in the water from the drillings F15, F12, F5, F13, drillings located between RWTP (Radioactive Waste Treatment Plant) - the 300 m{sup 3} tanks and the Spent Filters Storage (SFS). At south of this area, on the leaking direction of the underground water layer, in the drillings F1, F2, F3, F18 and at east, in F6, F7, the natural Uranium values are within the background for the underground-water. Distribution of Radon For the Radon determination with RAD 7 equipment, water samples were taken from the same piezo-metrical drilling, 2 or 4 times during of six months period, and then, the average contents were calculated, which varied between 0,35 - 2,1 Bq/l. The values higher than 1,1 -1,2 Bq/l were detected in the water taken from the drillings located in the northern part (F10, F11) and in the eastern part (F6, F8) of the Institute fences (around of the radioactive waste storage facilities). The concentrations of 0,3 - 0,5 Bq/l are in the underground-water layer 'intercepted' by the piezo-metrical drillings (F1, F2, F3) located near the Nuclear Reactor. Concentration of heavy metals: 0.04-0.08 mg/l Pb in F5, F14, F7, F8 exceeding MCA-Maximum Admissible Concentration (0.01 mg/l) for Pb, and for Zn in F5, F7, F8, F14 are 0.2-0.5 mg/l situated under MCA , and 0.18 mg/l in F18, in accordance with tendency of decreasing of concentration of contaminants. After 50 years of deploying nuclear activities on the site the underground water quality is in very good condition. Taking into consideration the direction of the underground water flow, it results that, only in the area of underground pipe, around of the research reactor and radioactive waste treatment plant, the quality of water is influenced, and remediation actions are not necessary. Based on measurements executed in F18, the water quality is the same with any other part of the region. During the decommissioning of the Research Reactor, the samples from 18 drillings will be analysed monthly, and the contents of the heavy metals, Pb and Zn, will be monitored carefully, together with all the factors: air, soil, vegetation, subsoil, water surface and underground water. A great attention will be paid t

Dragusin, Mitica [National Institute of Physics and Nuclear Engineering-Horia Hulubei - IFIN-HH, Bucharest-Magurele, Romania, POBox MG-6, 077125, Ilfov (Romania)

2008-01-15T23:59:59.000Z

222

Advanced Test Reactor Tour  

SciTech Connect (OSTI)

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

Miley, Don

2011-01-01T23:59:59.000Z

223

Advanced Test Reactor Tour  

ScienceCinema (OSTI)

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

Miley, Don

2013-05-28T23:59:59.000Z

224

Evidence of Graphitic AB Stacking Order of Graphite Oxides Hae-Kyung Jeong,*, Yun Pyo Lee, Rob J. W. E. Lahaye, Min-Ho Park,  

E-Print Network [OSTI]

Machinery Research Center, Jeonju 561-844, Battery Research Group, Korea Electrotechnology Research this simplified Brodie method is further discussed. I. Introduction Graphene sheets have extraordinary electronic of polymer-coated graphene-based sheets have been prepared via exfoliation of graphite oxide (GO) in water

225

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)

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.

Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

1994-04-01T23:59:59.000Z

226

Benchmark calculations for a heavy water research reactor using the WIMS-D4M code and a ENDF/B-V based library  

SciTech Connect (OSTI)

The results of unit-cell and global diffusion and transport calculations performed for the Georgia Tech heavy water research reactor using the WIMS-D4m code and a new ENDF/B-V based library are presented in this paper. Key cross sections, eigenvalues, neutron fluxes and peak power densities obtained from global diffusion calculations are compared.

Mo, S.C.

1993-12-31T23:59:59.000Z

227

Functional reliability evaluation of an MTR-pool type research reactor core using the loadcapacity interference model  

Science Journals Connector (OSTI)

Abstract This paper presents the functional reliability evaluation of Tehran Research Reactor (TRR) core in normal operation. The concept of functional reliability, borrowed from reliability physics, uses the well-known resistancestress or loadcapacity interference model that is used in the structural reliability framework. To use the loadcapacity interference model, uncertainties of significant parameters in system performance are propagated into system dynamics modeled with RELAP5/Mod 3.2 using Latin Hypercube Sampling (LHS) method and exceedance probability (EP) model is used as quantification method. The proposed method in this paper solves a common problem in reliability analysis, i.e., lack of sufficient failure data in specific operating conditions. Although defining failure criteria in normal operation are difficult, this paper focuses on the application of multiple states criteria to determine the status of a system. The status of the reactor core in normal operation is considered multiple states regarding to a performance representative parameter that is temperature in this work. Outlet temperatures of fuel hot and average channels were selected to be performance indicators in normal operation. Consulting with TRR engineers and operators as well as safety analysis report, two failure states were considered exceeding 65.1C and 58.9C for the hot channel and 50.4C and 45.6C for the average channel as upper and lower limits respectively. The calculated reliability was 9.1e?01 with 95% of confidence interval, which is in good agreement with experimental results. Using sensitivity analysis in input parameters, it was concluded that the value of the heat transfer coefficient parameter in fuel has the most significant effect on the results.

Ramin Barati; Saeed Setayeshi

2013-01-01T23:59:59.000Z

228

Nuclear Reactor (atomic reactor)  

Science Journals Connector (OSTI)

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

2008-01-01T23:59:59.000Z

229

Nuclear Fission Reactor Safety Research in FP7 and future perspectives  

E-Print Network [OSTI]

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 EUs 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 ...

Garbil, Roger

2014-01-01T23:59:59.000Z

230

Environmentally benign graphite intercalation compound composition for exfoliated graphite, flexible graphite, and nano-scaled graphene platelets  

SciTech Connect (OSTI)

A carboxylic-intercalated graphite compound composition for the production of exfoliated graphite, flexible graphite, or nano-scaled graphene platelets. The composition comprises a layered graphite with interlayer spaces or interstices and a carboxylic acid residing in at least one of the interstices, wherein the composition is prepared by a chemical oxidation reaction which uses a combination of a carboxylic acid and hydrogen peroxide as an intercalate source. Alternatively, the composition may be prepared by an electrochemical reaction, which uses a carboxylic acid as both an electrolyte and an intercalate source. Exfoliation of the invented composition does not release undesirable chemical contaminants into air or drainage.

Zhamu, Aruna; Jang, Bor Z.

2014-06-17T23:59:59.000Z

231

Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Tests 1 and 2  

SciTech Connect (OSTI)

A loss of Coolant Accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects of LOCA conditions on pressurized water reactor test fuel bundles. This experiment operation plan for the second and third experiments of the program will provide peak fuel cladding temperatures of up to 1172K (1650{degree}F) and 1061K (1450{degree}) respectively. for a long enough time to cause test fuel cladding deformation and rupture in both. Reflood coolant delay times and the reflooding rates for the experiments were selected from thermal-hydraulic data measured in the National Research Universal (NRU) reactor facilities and test train assembly during the first experiment.

Russcher, G. E.; Wilson, C. L.; Marshall, R, K.; King, L. L.; Parchen, L. J.; Pilger, J. P.; Hesson, G. M.; Mohr, C. L.

1981-09-01T23:59:59.000Z

232

Sixteenth water reactor safety information meeting: Proceedings: Volume 5, NUREG-1150, accident managment, recent advances in severe accident research, TMI-2, BWR Mark l shell failure  

SciTech Connect (OSTI)

This five-volume report contains 141 papers out of the 175 that were presented at the Sixteenth Water Reactor Safety Information Meeting held at the National Institute of Standards and Technology, Gaithersburg, Maryland, during the week of October 24--27, 1988. 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 twenty different papers presented by researchers from Germany, Italy, Japan, Sweden, Switzerland, Taiwan and the 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. This document, Volume 5, discusses NUREG-1150, Accident Management, Recent Advances in Severe Accident Research, BWR Mark I Shell Failure, and the Three Mile Island-2 Reactor.

Weiss, A.J. (comp.)

1989-03-01T23:59:59.000Z

233

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

E-Print Network [OSTI]

??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 (more)

Romano, Paul K. (Paul Kollath)

2009-01-01T23:59:59.000Z

234

Graphitized-carbon fiber/carbon char fuel  

DOE Patents [OSTI]

A method for recovery of intact graphitic fibers from fiber/polymer composites is described. The method comprises first pyrolyzing the graphite fiber/polymer composite mixture and then separating the graphite fibers by molten salt electrochemical oxidation.

Cooper, John F. (Oakland, CA)

2007-08-28T23:59:59.000Z

235

Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing  

Science Journals Connector (OSTI)

Abstract Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ~ 40 K , followed by a converter stage, solid deuterium at 5K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

E. Korobkina; G. Medlin; B. Wehring; A.I. Hawari; P.R. Huffman; A.R. Young; B. Beaumont; G. Palmquist

2014-01-01T23:59:59.000Z

236

Radiological Survey of Contaminated Installations of Research Reactor before Dismantling in High Dose Conditions with Complex for Remote Measurements of Radioactivity - 12069  

SciTech Connect (OSTI)

Decontamination and decommissioning of the research reactors MR (Testing Reactor) and RFT (Reactor of Physics and Technology) has recently been initiated in the National Research Center (NRC) 'Kurchatov institute', Moscow. These research reactors have a long history and many installations - nine loop facilities for experiments with different kinds of fuel. When decommissioning nuclear facilities it is necessary to measure the distribution of radioactive contamination in the rooms and at the equipment at high levels of background radiation. At 'Kurchatov Institute' some special remote control measuring systems were developed and they are applied during dismantling of the reactors MR and RFT. For a survey of high-level objects a radiometric system mounted on the robotic Brokk vehicle is used. This system has two (4? and collimated) dose meters and a high resolution video camera. Maximum measured dose rate for this system is ?8.5 Sv/h. To determine the composition of contaminants, a portable spectrometric system is used. It is a remotely controlled, collimated detector for scanning the distribution of radioactive contamination. To obtain a detailed distribution of contamination a remote-controlled gamma camera is applied. For work at highly contaminated premises with non-uniform background radiation, another camera is equipped with rotating coded mask (coded aperture imaging). As a result, a new system of instruments for remote radioactivity measurements with wide range of sensitivity and angular resolution was developed. The experience and results of measurements in different areas of the reactor and at its loop installations, with emphasis on the radioactive survey of highly-contaminated samples, are presented. These activities are conducted under the Federal Program for Nuclear and Radiation Safety of Russia. Adaptation of complex remote measurements of radioactivity and survey of contaminated installations of research reactor before dismantling in high dose conditions has proven successful. The radioactivity measuring devices for operation at high, non-uniform dose background were tested in the field and a new data of measurement of contamination distribution in the premises and installations were obtained. (authors)

Danilovich, Alexey; Ivanov, Oleg; Lemus, Alexey; Smirnov, Sergey; Stepanov, Vyacheslav; Volkovich, Anatoly [National Research Centre 'Kurchatov Institute', Moscow (Russian Federation)

2012-07-01T23:59:59.000Z

237

Polyelectrolyte-Induced Reduction of Exfoliated Graphite Oxide...  

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

Reduction of Exfoliated Graphite Oxide: A Facile Route to Synthesis of Soluble Graphene Nanosheets. Polyelectrolyte-Induced Reduction of Exfoliated Graphite Oxide: A Facile...

238

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

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

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

239

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

E-Print Network [OSTI]

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 ...

Forbes, Ian Alexander

1970-01-01T23:59:59.000Z

240

Utilization of Ceramic Uranium Fuels in ARIES-RS Fusion Reactor  

Science Journals Connector (OSTI)

This study presents the neutronic performance of the ARIES-RS fusion reactor design using different natural ceramic uranium fuels,...2, UN or U3Si2..., dispersed in graphite matrix. These fissionable fuels insert...

Mustafa beyli

2004-03-01T23:59:59.000Z

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


241

Microstructural Characterization of Next Generation Nuclear Graphites  

SciTech Connect (OSTI)

This article reports the microstructural characteristics of various petroleum and pitch based nuclear graphites (IG-110, NBG-18, and PCEA) that are of interest to the next generation nuclear plant program. Bright-field transmission electron microscopy imaging was used to identify and understand the different features constituting the microstructure of nuclear graphite such as the filler particles, microcracks, binder phase, rosette-shaped quinoline insoluble (QI) particles, chaotic structures, and turbostratic graphite phase. The dimensions of microcracks were found to vary from a few nanometers to tens of microns. Furthermore, the microcracks were found to be filled with amorphous carbon of unknown origin. The pitch coke based graphite (NBG-18) was found to contain higher concentration of binder phase constituting QI particles as well as chaotic structures. The turbostratic graphite, present in all of the grades, was identified through their elliptical diffraction patterns. The difference in the microstructure has been analyzed in view of their processing conditions.

Karthik Chinnathambi; Joshua Kane; Darryl P. Butt; William E. Windes; Rick Ubic

2012-04-01T23:59:59.000Z

242

Isotope correlations for determining the isotopic composition of plutonium produced in research and power reactors using the experimental data obtained by alpha and mass spectrometry  

Science Journals Connector (OSTI)

Correlations have been developed for obtaining the isotopic composition of Pu produced in Indian research (CIRUS, DHRUVA) and power (PHWR) reactors. The experimental data obtained on 238Pu/(239Pu + 240Pu) alpha activity ratio using alpha spectrometry and on 240Pu/239Pu, 241Pu/239Pu, 242Pu/239Pu atom ratios by thermal ionisation mass spectrometry were used for developing isotope correlations.

S.K. Aggarwal; D. Alamelu

2005-01-01T23:59:59.000Z

243

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

E-Print Network [OSTI]

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 ...

Romano, Paul K. (Paul Kollath)

2009-01-01T23:59:59.000Z

244

Graphite Oxidation Thermodynamics/Reactions  

SciTech Connect (OSTI)

The vulnerability of graphite-matrix spent nuclear fuel to oxidation by the ambient atmosphere if the fuel canister is breached was evaluated. Thermochemical and kinetic data over the anticipated range of storage temperatures (200 to 400 C) were used to calculate the times required for a total carbon mass loss of 1 mgcm-2 from a fuel specimen. At 200 C, the time required to produce even this small loss is large, 900,000 yr. However, at 400 C the time required is only 1.9 yr. The rate of oxidation at 200 C is negligible, and the rate even at 400 C is so small as to be of no practical consequence. Therefore, oxidation of the spent nuclear fuel upon a loss of canister integrity is not anticipated to be a concern based upon the results of this study.

Propp, W.A.

1998-09-01T23:59:59.000Z

245

RERTR program activities related to the development and application of new LEU fuels. [Reduced Enrichment Research and Test Reactor; low-enriched uranium  

SciTech Connect (OSTI)

The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the current 1.7 g U/cm/sup 3/ to the 7.0 g U/cm/sup 3/ which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years.

Travelli, A.

1983-01-01T23:59:59.000Z

246

Strength scaling of brittle graphitic foam  

Science Journals Connector (OSTI)

...many different type of open- and closed-cell solids such as aerogels (Pekala et al. 1991), graphitic foams (Brezny & Green...mechanical properties and scaling law relationships for silica aerogels and their organic counterparts. Mater. Res. Soc. Symp...

2002-01-01T23:59:59.000Z

247

Immobilization of Rocky Flats Graphite Fines Residue  

SciTech Connect (OSTI)

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.

Rudisill, T.S.

1999-04-06T23:59:59.000Z

248

Electric Resistivity of Interstitial Compounds of Graphite  

Science Journals Connector (OSTI)

Assuming that the formation of interstitial compounds is accompanied by creation of excess holes in the otherwise full band of graphite it is shown that a linear energy?momentum relation at the Brillouin zone corners is incapable of explaining the decrease of the electric resistance with oxidation. It appears that for a more general model the decrease in relative resistance should be independent of temperature for large oxidations if suitable corrections for the initial conditions are made. Data for polycrystallinegraphite corrected for the existence of an energy gap and of excess holes in the untreated material give curves which converge for higher oxidation with the curve for natural graphite. Discussion of the low temperature properties of graphite leads to the conclusion that large graphite crystals possess slightly overlapping zones.

S. Mrozowski

1953-01-01T23:59:59.000Z

249

JOURNAL DE PHYSIQUE Colloque C6, supplment au n 12, Tome 37, dcembre 1976, page C6-53 MOSSBAUER STUDIES OF MOLECULES ADSORBED ON GRAPHITE (*)  

E-Print Network [OSTI]

STUDIES OF MOLECULES ADSORBED ON GRAPHITE (*) S. BUKSHPAN and T. SONNINO Soreq Nuclear Research Center obtained on Nuclear Gamma Resonance studies of molecules adsorbed on the basal plane of graphite. In all [1]. The obtained results supply information on the dynamical properties of the adsorbed molecules

Paris-Sud XI, Université de

250

Generation -IV Reactor Concepts  

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

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

251

Millimeter-Wave Thermal Analysis Development and Application to GEN IV Reactor Materials  

SciTech Connect (OSTI)

New millimeter-wave thermal analysis instrumentation has been developed and studied for characterization of materials required for diverse fuel and structural needs in high temperature reactor environments such as the Next Generation Nuclear Plant (NGNP). A two-receiver 137 GHz system with orthogonal polarizations for anisotropic resolution of material properties has been implemented at MIT. The system was tested with graphite and silicon carbide specimens at temperatures up to 1300 C inside an electric furnace. The analytic and hardware basis for active millimeter-wave radiometry of reactor materials at high temperature has been established. Real-time, non contact measurement sensitivity to anisotropic surface emissivity and submillimeter surface displacement was demonstrated. The 137 GHz emissivity of reactor grade graphite (NBG17) from SGL Group was found to be low, ~ 5 %, in the 500 1200 C range and increases by a factor of 2 to 4 with small linear grooves simulating fracturing. The low graphite emissivity would make millimeter-wave active radiometry a sensitive diagnostic of graphite changes due to environmentally induced stress fracturing, swelling, or corrosion. The silicon carbide tested from Ortek, Inc. was found to have a much higher emissivity at 137 GHz of ~90% Thin coatings of silicon carbide on reactor grade graphite supplied by SGL Group were found to be mostly transparent to millimeter-waves, increasing the 137 GHz emissivity of the coated reactor grade graphite to about ~14% at 1250 C.

Paul Wosko; Sundram, S. K.

2012-10-16T23:59:59.000Z

252

E-Print Network 3.0 - advanced fission reactors Sample Search...  

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

fission reactors, which release energy by splitting atoms... ) International Thermonuclear Experimental Reactor (ITER), which will be ... Source: Fusiongnition Research...

253

Analytical support for the ORR (Oak Ridge Research Reactor) whole-core LEU U/sub 3/Si/sub 2/-Al fuel demonstration  

SciTech Connect (OSTI)

Analytical methods used to analyze neutronic data from the whole-core LEU fuel demonstration in the Oak Ridge Research Reactor are briefly discussed. Calculated eigenvalues corresponding to measured critical control rod positions are presented for each core used in the gradual transition from an all HEU to an all LEU configuration. Some calculated and measured results, including ..beta../sub eff//l/sub p/, are compared for HEU and LEU fresh fuel criticals. Finally, the perturbing influences of the six voided beam tubes on certain core parameters are examined. For reasons yet to be determined, differential shim rod worths are not well-calculated in partially burned cores.

Bretscher, M.M.

1986-01-01T23:59:59.000Z

254

Comparison of gasification kinetics parameters of different types of nuclear graphite  

SciTech Connect (OSTI)

A chemical-reaction kinetics model of nuclear graphite gasification has recently been developed and successfully validated with gasification rate measurements for nuclear graphite grades of IG-110, IG-430, NBG-18 and NBG-25. The model employs 4 elementary chemical reactions with applicable parameters, including the values and Gaussian-like distributions of the specific activation energies, the pre-exponential coefficients for adsorption of oxygen and desorption of CO and CO{sub 2} gases, and the surface area of free active sites. These parameters are determined from the reported measurements of the total gasification and transient weight loss using a multi-parameter optimization algorithm. The determined chemical kinetics parameters for IG-100 and NB-25 are nearly the same, but slightly different from those for NBG-18 and IG-430. The initial specific area of free active sites is inversely proportional to the square root of the mass or volume of the graphite specimens used in experiments. The recommended chemical kinetics parameters in this paper for these grades of nuclear graphite should be applicable to future safety analysis of high-temperature gas cooled reactors in the unlikely event of a massive air ingress accident. (authors)

El-Genk, M. S. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); Tournier, J. M. P. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States)

2012-07-01T23:59:59.000Z

255

Calculated criticality for sup 235 U/graphite systems using the VIM Monte Carlo code  

SciTech Connect (OSTI)

Calculations for highly enriched uranium and graphite systems gained renewed interest recently for the new production modular high-temperature gas-cooled reactor (MHTGR). Experiments to validate the physics calculations for these systems are being prepared for the Transient Reactor Test Facility (TREAT) reactor at Argonne National Laboratory (ANL-West) and in the Compact Nuclear Power Source facility at Los Alamos National Laboratory. The continuous-energy Monte Carlo code VIM, or equivalently the MCNP code, can utilize fully detailed models of the MHTGR and serve as benchmarks for the approximate multigroup methods necessary in full reactor calculations. Validation of these codes and their associated nuclear data did not exist for highly enriched {sup 235}U/graphite systems. Experimental data, used in development of more approximate methods, dates back to the 1960s. The authors have selected two independent sets of experiments for calculation with the VIM code. The carbon-to-uranium (C/U) ratios encompass the range of 2,000, representative of the new production MHTGR, to the ratio of 10,000 in the fuel of TREAT. Calculations used the ENDF/B-V data.

Collins, P.J.; Grasseschi, G.L.; Olsen, D.N. (Argonne National Lab.-West, Idaho Falls (United States)); Finck, P.J. (Argonne National Lab., IL (United States))

1992-01-01T23:59:59.000Z

256

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

DOE Patents [OSTI]

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.

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

2010-11-02T23:59:59.000Z

257

Advanced Reactor Technologies | Department of Energy  

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

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

258

Aerosol and graphitic carbon content of snow  

SciTech Connect (OSTI)

Snow samples from southern New Mexico, west Texas, Antarctica, and Greenland were analyzed for aerosol and graphitic carbon. Graphitic carbon contents were found to be between 2.2 and 25 ..mu..g L/sup -1/ of snow meltwater; water-insoluble aerosol content varied between 0.62 and 8.5 mg L/sup -1/. For comparison, two samples of Camp Century, Greenland, ice core, having approximate ages of 4,000 and 6,000 years, were also analyzed. Ice core graphitic carbon contents were found to be 2.5 and 1.1 ..mu..g L/sup -1/. copyrightAmerican Geophysical Union 1987

Chy-acute-accentlek, P.; Srivastava, V.; Cahenzli, L.; Pinnick, R.G.; Dod, R.L.; Novakov, T.; Cook, T.L.; Hinds, B.D.

1987-08-20T23:59:59.000Z

259

Systems and methods for dismantling a nuclear reactor  

DOE Patents [OSTI]

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.

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

2014-10-28T23:59:59.000Z

260

Characterization of exfoliated graphite for heavy oil sorption  

Science Journals Connector (OSTI)

In this paper are reported some experimental data related to the influence of preparation regimes and characteristics of exfoliated graphite based sorbents produced by thermal expansion of H2SO4-graphite intercal...

Gabriela Hristea; P. Budrugeac

2008-03-01T23:59:59.000Z

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


261

Some optical properties of graphite from IR to millimetric wavelengths  

Science Journals Connector (OSTI)

......powder obtained by striking an electric arc from graphitic electrodes in an...the material from the exciting electric field and more energy is scattered...graphitic material from electrical discharges in various gases, TH, BE, XY......

Robert J. Papoular; Renaud Papoular

2014-10-01T23:59:59.000Z

262

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)

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.

Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

2012-10-01T23:59:59.000Z

263

Method of making segmented pyrolytic graphite sputtering targets  

DOE Patents [OSTI]

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.

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

1994-02-08T23:59:59.000Z

264

NOVEL GRAPHITE SALTS OF HIGH OXIDIZING POTENTIAL  

E-Print Network [OSTI]

hr volatiles none AsFS AsFS > CF4 AsFS AsFS AsFS C13.l AsF 6C16.l AsF 6 C24.3AsF6l/2 F2 > CF4> CF4 > CF4 Table VI-7. Graphite hexafluoroarsenate salts +

McCarron III, Eugene Michael

2010-01-01T23:59:59.000Z

265

Phonon dispersion of graphite J. Maultzsch  

E-Print Network [OSTI]

the acoustic branches below 400 cm-1 were measured by inelastic neutron scattering [1]. The optical phonons. The phonon dispersion of graphite was determined by inelastic X-ray scattering along the -K, K-M, and -K. This coupling dominates the scattering mechanism in both electronic transport and Raman scattering. Many

Nabben, Reinhard

266

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

SciTech Connect (OSTI)

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

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

2013-11-01T23:59:59.000Z

267

Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Test 2  

SciTech Connect (OSTI)

A loss-of-coolant accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects on pressurized water reactor (PWR) test fuel bundles. This Experiment Operation Plan (EOP) Addendum 2, together with the referenced EOP, describes the desired operating conditions and additional hazards review associated with the four-part MT-2 experiment. The primary portions of the experiment, MT-2.2 and MT-2.3, will evaluate the following: 1) the mechanical deformation of pressurized fuel rods subjected to a slow LOCA, using reflood water for temperature control, that is designed to produce cladding temperatures in the range from 1033 to 1089K (1400 to 1500F) for an extended time, and 2) the effects of the deformed and possibly failed cladding on the thermal-hydraulic performance of the test assembly during simulated LOCA heating and reflooding. The secondary portions of the experiment, MT-2.1 and MT-2.4, are intended to provide thermal-hydraulic calibration information during two-stage reflood conditions for 1) relatively low cladding temperatures, <839K (1050F), on nondeformed rods, and 2) moderately high cladding temperatures, <1089K (1500F), on deformed rods.

Russcher, G. E.; Barner, J. O.; Hesson, G. M.; Wilson, C. L.; Parchen, L. J.; Cunningham, M. E.; Marshall, R. K.; Mohr, C. L.

1981-09-01T23:59:59.000Z

268

Diffusion velocity correlation for nuclear graphite gasification at high temperature and low Reynolds numbers  

SciTech Connect (OSTI)

The safety analysis of High-Temperature and Very High Temperature gas-cooled Reactors requires reliable estimates of nuclear graphite gasification as a function of temperature, among other parameters, in the unlikely event of an air ingress accident. Although the rates of the prevailing chemical reactions increase exponentially with temperature, graphite gasification at high temperatures is limited by the oxygen diffusion through the boundary layer. The effective diffusion velocity depends on the total flow rate and pressure of the bulk air-gas mixture. This paper develops a semi-empirical Sherwood number correlation for calculating the oxygen diffusion velocity. The correlation is based on a compiled database of the results of convective heat transfer experiments with wires and cylinders of different diameters in air, water and paraffin oil at 0.006 {<=} Re {<=} 1,604 and 0.068 {<=} Sc {<=} 35.2, and of mass transfer experiments at 4.8 {<=} Re {<=} 77 and 1,300 {<=} Sc {<=} 2,000. The developed correlation is within {+-} 8% of the compiled database of 567 data points and consistent with reported gasification rate measurements at higher temperatures in experiments using different size specimens of nuclear graphite grades of NBG-18 and NB-25, IG-11, IG-110 and IG-430 in atmospheric air at 0.08 {<=} Re {<=} 30. Unlike the Graetz solution that gives a constant Sh of 3.66 at Re {<=} 1.0, the present correlation shows Sh decreases monotonically to much lower values with decreasing Re. (authors)

El-Genk, M. S. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); Tournier, J. M. P. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States)

2012-07-01T23:59:59.000Z

269

Radiological consequences of ship collisions that might occur in U.S. Ports during the shipment of foreign research reactor spent nuclear fuel to the United States in break-bulk freighters  

SciTech Connect (OSTI)

Accident source terms, source term probabilities, consequences, and risks are developed for ship collisions that might occur in U.S. ports during the shipment of spent fuel from foreign research reactors to the United States in break-bulk freighters.

Sprung, J.L.; Bespalko, S.J.; Massey, C.D.; Yoshimura, R. [Sandia National Laboratory, Albuquerque, NM (United States); Johnson, J.D. [GRAM Inc., Albuquerque, NM (United States); Reardon, P.C. [PCRT Technologies, Albuquerque, NM (United States); Ebert, M.W.; Gallagher D.W. [Science Applications International Corp., Reston, VA (United States)

1996-08-01T23:59:59.000Z

270

Evaluation of potential for xenon oscillation in the N reactor  

SciTech Connect (OSTI)

As a result of the Chernobyl accident, the susceptibility of the N reactor, a large graphite-moderated reactor, to xenon oscillations was studied using the xenon simulation code GFN-XE. Core loadings fueled with low-enriched uranium (LEU) metal fuel and with high-enriched uranium (HEU) alloy fuel were evaluated. Susceptibility to radial, axial, and azimuthal oscillations was computed for a series of variables. An improvement in stability was observed for the alloy fuel core. The results of the GFN-XE analyses indicate that large graphite-moderated reactors can be designed that are not susceptible to xenon oscillations and that different core loadings such as a HEU-Zr fuel load can have an even larger margin of stability.

Finfrock, S.H.; Toffer, H.

1990-06-01T23:59:59.000Z

271

Probabilistic risk assessment of N Reactor using NUREG-1150 methods  

SciTech Connect (OSTI)

A Level III probabilistic risk assessment (PRA) has been performed for N Reactor, a US Department of Energy (DOE) Category A production reactor. The main contractor is Westinghouse Hanford Company (Westinghouse Hanford). The PRA methodology developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL) in support of the NUREG-1150 (Reference 1) effort were used for this analysis. N Reactor is a graphite-moderated pressurized water reactor designed by General Electric. The dual-purpose 4000 MWt nuclear plant is located within the Hanford Site in the south-central part of the State of Washington. In addition to producing special materials for the DOE, N Reactor generates 860 MWe for the Washington Public Power Supply System. The reactor has been operated successfully and safely since 1963, and was put into standby status in 1988 due to the changing need in special nuclear material. 3 refs., 4 tabs.

Wang, O.S.; Baxter, J.T.; Coles, G.A.; Powers, T.B.; Zentner, M.D.

1989-11-01T23:59:59.000Z

272

Facile Sonochemical Synthesis of Graphite Intercalation Compounds  

Science Journals Connector (OSTI)

Department of Chemistry and Biochemistry, Texas Tech University, Lubbock, Texas 79409, and Sandia National Laboratories, Albuquerque, New Mexico 87185 ... Similar intercalation of layered inorganic solids has been previously observed using high-intensity ultrasound as the energy source. ... Sonication for a longer period of time produces a blue-black material, which may either be due to partial deintercalation or to the destruction of the intercalated material to reform graphite and potassium. ...

Jess E. Jones; Michael C. Cheshire; Dominick J. Casadonte, Jr.; Carol C. Phifer

2004-05-08T23:59:59.000Z

273

Atomic resolution images of graphite in air  

SciTech Connect (OSTI)

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.

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

1988-12-01T23:59:59.000Z

274

Nondestructive Evaluation of Nuclear-Grade Graphite  

SciTech Connect (OSTI)

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.

Dennis C. Kunerth; Timothy R. McJunkin

2011-07-01T23:59:59.000Z

275

NUCLEAR REACTORS.  

E-Print Network [OSTI]

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

Belachew, Dessalegn

2010-01-01T23:59:59.000Z

276

Use of phenomena identification and ranking (PIRT) process in research related to design certification of the AP600 advanced passive light water reactor (LWR)  

SciTech Connect (OSTI)

The AP600 LWR is a new advanced passive design that has been submitted to the USNRC for design certification. Within the certification process the USNRC will perform selected system thermal hydraulic response audit studies to help confirm parts of the vendor`s safety analysis submittal. Because of certain innovative design features of the safety systems, new experimental data and related advances in the system thermal hydraulic analysis computer code are being developed by the USNRC. The PIRT process is being used to focus the experimental and analytical work to obtain a sufficient and cost effective research effort. The objective of this paper is to describe the application and most significant results of the PIRT process, including several innovative features needed in the application to accommodate the short design certification schedule. The short design certification schedule has required that many aspects of the USNRC experimental and analytical research be performed in parallel, rather than in series as was normal for currently operating LWRS. This has required development and use of management techniques that focus and integrate the various diverse parts of the research. The original PIRTs were based on inexact knowledge of an evolving reactor design, and concentrated on the new passive features of the design. Subsequently, the PIRTs have evolved in two more stages as the design became more firm and experimental and analytical data became available. A fourth and final stage is planned and in progress to complete the PIRT development. The PIRTs existing at the end of each development stage have been used to guide the experimental program, scaling analyses and code development supporting the audit studies.

Wilson, G.E.; Fletcher, C.D. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Eltawila, F. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

1996-07-01T23:59:59.000Z

277

Immobilization of Rocky Flats Graphite Fines Residues  

SciTech Connect (OSTI)

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.

Rudisill, T. S.

1998-11-06T23:59:59.000Z

278

U.S./Belarus/Ukraine joint research on the biomedical effects of the Chernobyl Reactor Accident. Final report  

SciTech Connect (OSTI)

The National Cancer Institute has negotiated with the governments of Belarus and Ukraine (Ministers/Ministries of Health, institutions and scientists) to develop scientific research protocols to study the effects of radioactive iodine released by the Chernobyl accident upon thyroid anatomy and function in defined cohorts of persons under the age of 19 years at the time of the accident. These studies include prospective long term medical follow-up of the cohort and the reconstruction of the radiation dose to each cohort subject's thyroid. The protocol for the study in Belarus was signed by the US and Belorussian governments in May 1994 and the protocol for the study in Ukraine was signed by the US and Ukraine in May 1995. A second scientific research protocol also was negotiated with Ukraine to study the feasibility of a long term study to follow the development of leukemia and lymphoma among Ukrainian cleanup workers; this protocol was signed by the US and Ukraine in October 1996.

Bruce Wachholz

2000-06-20T23:59:59.000Z

279

Using reactor operating experience to improve the design of a new Broad Application Test Reactor  

SciTech Connect (OSTI)

Increasing regulatory demands and effects of plant aging are limiting the operation of existing test reactors. Additionally, these reactors have limited capacities and capabilities for supporting future testing missions. A multidisciplinary team of experts developed sets of preliminary safety requirements, facility user needs, and reactor design concepts for a new Broad Application Test Reactor (BATR). Anticipated missions for the new reactor include fuels and materials irradiation testing, isotope production, space testing, medical research, fusion testing, intense positron research, and transmutation doping. The early BATR design decisions have benefited from operating experiences with existing reactors. This paper discusses these experiences and highlights their significance for the design of a new BATR.

Fletcher, C.D.; Ryskamp, J.M.; Drexler, R.L.; Leyse, C.F.

1993-07-01T23:59:59.000Z

280

Low Cost SiOx-Graphite and Olivine Materials  

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

Sotowa (Showa-Denko) Objective Synthesize and evaluate doped manganese phosphate as low cost cathode material Replace graphite anode with an alternative material that meets the...

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


281

Low Cost SiOx-Graphite and Olivine Materials  

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

Replace graphite anode with an alternative material that meets the requirement for low cost and high energy. Continue development of binders for the cathode and alternative anode...

282

Forming gas treatment of lithium ion battery anode graphite powders  

DOE Patents [OSTI]

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.

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

2014-09-16T23:59:59.000Z

283

Binding and Diffusion of Lithium in Graphite: Quantum Monte Carlo...  

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

Binding and Diffusion of Lithium in Graphite: Quantum Monte Carlo Benchmarks and Validation of van der Waals Density Functional Methods P. Ganesh,* , Jeongnim Kim, Changwon...

284

Reactor and Nuclear Systems Division (RNSD)  

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

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

285

Fission Product Impact Reduction via Protracted In-core Retention in Very High Temperature Reactor (VHTR) Transmutation Scenarios  

E-Print Network [OSTI]

of HTGR by improvements in thermal efficiency and deployment for high-temperature applications such as hydrogen production, sea-water desalination and industrial process heat supply [17]. The VHTR is a graphite-moderated helium-cooled reactor...-based transmutation concept takes advantage of the higher number of steps it takes for a neutron to slow-down to thermal energies in graphite than the steps required in conventional LWR. The reduced slowing-down rate in graphite media favors the attainment...

Alajo, Ayodeji Babatunde

2011-08-08T23:59:59.000Z

286

Low-cost and durable catalyst support for fuel cells: graphite...  

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

cost and durable catalyst support for fuel cells: graphite submicronparticles. Low-cost and durable catalyst support for fuel cells: graphite submicronparticles. Abstract: Low-cost...

287

Evaluation of the Transient Thermal Performance of a Graphite Foam/Phase Change Material Composite.  

E-Print Network [OSTI]

?? The thermal transient response of graphite foam infiltrated with paraffin wax as a thermal protection composite was investigated. Graphite foam is a rigid open-celled (more)

Trammell, Michael Paul

2013-01-01T23:59:59.000Z

288

naval reactors  

National Nuclear Security Administration (NNSA)

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

289

Thermonuclear Reflect AB-Reactor  

E-Print Network [OSTI]

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

Alexander Bolonkin

2008-03-26T23:59:59.000Z

290

Oxidation of graphite surface: the role of water  

E-Print Network [OSTI]

Based on density functional calculations, we demonstrate a significant difference in oxidation patterns between graphene and graphite and the formation of defects after oxidation. Step-by-step modeling demonstrates that oxidation of 80% of the graphite surface is favorable. Oxidation above half of the graphite surface significantly decreases the energy costs of vacancy formation with CO2 production. The presence of water is crucial in the transformation of epoxy groups to hydroxyl, the intercalation with further bundle and exfoliation. In water-rich conditions, water intercalates graphite at the initial stages of oxidation and oxidation, which is similar to the oxidation process of free-standing graphene; in contrast, in water-free conditions, large molecules intercalate graphite only after oxidation occurs on more than half of the surface.

Boukhvalov, D W

2014-01-01T23:59:59.000Z

291

Graphite and its Hidden Superconductivity | Stanford Synchrotron Radiation  

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

Graphite and its Hidden Superconductivity Graphite and its Hidden Superconductivity Wednesday, November 20, 2013 - 2:00pm SLAC, Conference Room 137-322 Pablo Esquinazi, University of Leipzig We review different experimental results that indicate the existence of granular superconductivity at high temperatures at graphite interfaces. In particular we will discuss the following experimental results: The temperature and magnetic field dependence of the electrical resistance of bulk and thin graphite samples and its relation with the existence of two-dimensional (2D) interfaces. The anomalous hysteresis in the magnetoresistance observed in graphite thin samples as well as its enhancement restricting the current path within the sample. The Josephson behavior of the current-voltage characteristics with

292

Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels -- Final Report under the International Nuclear Energy Research Initiative (I-NERI)  

SciTech Connect (OSTI)

The objective of this INERI project was to develop improved fuel behavior models for gas reactor coated-particle fuels and to explore improved coated-particle fuel designs that could be used reliably at very high burnups and potentially in gas-cooled fast reactors. Project participants included the Idaho National Engineering Laboratory (INEEL), Centre tude Atomique (CEA), and the Massachusetts Institute of Technology (MIT). To accomplish the project objectives, work was organized into five tasks.

David Petti; Philippe Martin; Mayeul Phlip; Ronald Ballinger; Petti does not have NT account

2004-12-01T23:59:59.000Z

293

DOE - Office of Legacy Management -- Ames Laboratory Research...  

Office of Legacy Management (LM)

Ames Laboratory Research Reactor Facility - IA 03 FUSRAP Considered Sites Site: Ames Laboratory Research Reactor Facility (IA.03) Designated Name: Alternate Name: Location:...

294

Light Water Reactor Sustainability  

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

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

295

Refueling Liquid-Salt-Cooled Very High-Temperature Reactors  

SciTech Connect (OSTI)

The liquid-salt-cooled very high-temperature reactor (LS-VHTR), also called the Advanced High-Temperature Reactor (AHTR), is a new reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. Depending upon goals, the peak coolant operating temperatures are between 700 and 1000 deg. C, with reactor outputs between 2400 and 4000 MW(t). Several fluoride salt coolants that are being evaluated have melting points between 350 and 500 deg. C, values that imply minimum refueling temperatures between 400 and 550 deg. C. At operating conditions, the liquid salts are transparent and have physical properties similar to those of water. A series of refueling studies have been initiated to (1) confirm the viability of refueling, (2) define methods for safe rapid refueling, and (3) aid the selection of the preferred AHTR design. Three reactor cores with different fuel element designs (prismatic, pebble bed, and pin-type fuel assembly) are being evaluated. Each is a liquid-salt-cooled variant of a graphite-moderated high-temperature reactor. The refueling studies examined applicable refueling experience from high-temperature reactors (similar fuel element designs) and sodium-cooled fast reactors (similar plant design with liquid coolant, high temperatures, and low pressures). The findings indicate that refueling is viable, and several approaches have been identified. The study results are described in this paper. (authors)

Forsberg, Charles W. [Oak Ridge National Laboratory, P.O. Box 2008 Oak Ridge, TN 37831 (United States); Peterson, Per F. [Nuclear Engineering Department, University of California at Berkeley, 6124a Etcheverry Hall, Berkeley, CA 94720 (United States); Cahalan, James E. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Enneking, Jeffrey A. [Areva NP (United States); Phil MacDonald [Consultant, Cedar Hill, TX (United States)

2006-07-01T23:59:59.000Z

296

University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor  

SciTech Connect (OSTI)

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

Eric C. Woolstenhulme; Dana M. Hewit

2008-09-01T23:59:59.000Z

297

(Fuel, fission product, and graphite technology)  

SciTech Connect (OSTI)

Travel to the Forschungszentrum (KFA) -- Juelich described in this report was for the purpose of participating in the annual meeting of subprogram managers for the US/DOE Umbrella Agreement for Fuel, Fission Product, and Graphite Technology. At this meeting the highlights of the cooperative exchange were reviewed for the time period June 1989 through June 1990. The program continues to contribute technology in an effective way for both countries. Revision 15 of the Subprogram Plan will be issued as a result of the meeting. There was interest expressed by KFA management in the level of support received from the NPR program and in potential participation in the COMEDIE loop experiment being conducted at the CEA.

Stansfield, O.M.

1990-07-25T23:59:59.000Z

298

Comparison of Cycling Performance of Lithium Ion Cell Anode Graphites  

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

Comparison of Cycling Performance of Lithium Ion Cell Anode Graphites Comparison of Cycling Performance of Lithium Ion Cell Anode Graphites Title Comparison of Cycling Performance of Lithium Ion Cell Anode Graphites Publication Type Journal Article Year of Publication 2012 Authors Ridgway, Paul L., Honghe Zheng, A. F. Bello, Xiangyun Song, Shidi Xun, Jin Chong, and Vincent S. Battaglia Journal Journal of The Electrochemical Society Volume 159 Issue 5 Pagination A520 Date Published 2012 ISSN 00134651 Abstract Battery grade graphite products from major suppliers to the battery industry were evaluated in 2325 coin cells with lithium counter electrodes. First and ongoing cycle efficiency, total and reversible capacity, cycle life and discharge rate performance were measured to compare these anode materials. We then ranked the graphites using a formula which incorporates these performance measures to estimate the cost of the overall system, relative to the cost of a system using MCMB. This analysis indicates that replacing MCMB with CCP-G8 (Conoco Phillips) would add little to no cost, whereas each of the other graphites would lead to a more costly system. Therefore we chose CCP-G8 as the new baseline graphite for the BATT program.

299

Catalytic reactor  

DOE Patents [OSTI]

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

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

2009-03-10T23:59:59.000Z

300

Light Water Reactor Sustainability Technical Documents | Department of  

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

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

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


301

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

SciTech Connect (OSTI)

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)

Messenger, S. J. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., 54-1717, Cambridge, MA 02139 (United States); Forsberg, C. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., 24-207, Cambridge, MA 02139 (United States); Massie, M. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., NW12-230, Cambridge, MA 02139 (United States)

2012-07-01T23:59:59.000Z

302

ANALYSIS OF SEPCTRUM CHOICES FOR SMALL MODULAR REACTORS-PERFORMANCE AND DEVELOPMENT  

E-Print Network [OSTI]

. The research mainly focused on producing a small modular reactor (Pebble Bed Modular Reactor) design to analyze the fuel depletion and plutonium and minor actinide accumulation with varying power densities. The reactors running at low power densities were found...

Kafle, Nischal

2011-04-26T23:59:59.000Z

303

This article appeared in a journal published by Elsevier. The attached copy is furnished to the author for internal non-commercial research  

E-Print Network [OSTI]

analysis and reliability analysis are included. Various creep models and failure models are used, Swansea SA2 8PP, UK h i g h l i g h t s Nuclear graphite mechanical analysis is accomplished. Both stress material in the high temperature gas-cooled reactors (HTRs). As mechanical properties of graphite change

Martin, Ralph R.

304

A future for nuclear energy: pebble bed reactors  

Science Journals Connector (OSTI)

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

Andrew C. Kadak

2005-01-01T23:59:59.000Z

305

Experimental thermal conductivity and contact conductance of graphite composites  

E-Print Network [OSTI]

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 investigation conductivity...

Jackson, Marian Christine

2012-06-07T23:59:59.000Z

306

An Investigation of the effect of graphite degradation on irreversible  

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

An Investigation of the effect of graphite degradation on irreversible An Investigation of the effect of graphite degradation on irreversible capacity in lithium-ion cells. Title An Investigation of the effect of graphite degradation on irreversible capacity in lithium-ion cells. Publication Type Journal Article Year of Publication 2008 Authors Hardwick, Laurence J., Marek Marcinek, Leanne Beer, John B. Kerr, and Robert Kostecki Journal Electrochemical Society Volume 155 Start Page A442 Issue 6 Pagination A442-A447 Keywords chromatography, electrochemical electrodes, electrochemical impedance spectroscopy, Fourier transform spectra, graphite, infrared spectra, lithium, mass spectra, Raman spectra, scanning electron microscopy, secondary cells, sputtering, surface structure Abstract The effect of surface structural damage on graphitic anodes, commonly observed in tested Li-ion cells, was investigated. Similar surface structural disorder was artificially induced in Mag-10 synthetic graphite anodes using argon-ion sputtering. Raman microscopy, scanning electron microscopy, and Brunauer-Emmett-Teller measurements confirmed that Ar-ion sputtered Mag-10 electrodes display a similar degree of surface degradation as the anodes from tested Li-ion cells. Artificially modified Mag-10 anodes showed double the irreversible charge capacity during the first formation cycle compared to fresh unaltered anodes. Impedance spectroscopy and Fourier transform infrared spectroscopy on surface-modified graphite anodes indicated the formation of a thicker and slightly more resistive solid electrolyte interphase (SEI) layer. Gas chromatography/mass spectroscopy analysis of solvent extracts from the electrodes detected the presence of new compounds with Mw on the order of 1600gmol-1 for the surface-modified electrode with no evidence of elevated Mw species for the unmodified electrode. The structural disorder induced in the graphite during long-term cycling may be responsible for the slow and continuous SEI layer reformation, and consequently, the loss of reversible capacity due to the shift of lithium inventory in cycled Li-ion cells.

307

Compression induced delamination in a unidirectional graphite/epoxy composite  

E-Print Network [OSTI]

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...

Earley, John W.

2012-06-07T23:59:59.000Z

308

An investigation of damage accumulation in graphite/epoxy laminates  

E-Print Network [OSTI]

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...

Norvell, Robert Gerald

2012-06-07T23:59:59.000Z

309

Slow, stable delamination in graphite/epoxy composites  

E-Print Network [OSTI]

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...

Razi, Hamid

2012-06-07T23:59:59.000Z

310

Effect of Vinylene Carbonate on Graphite Anode Cycling Efficiency  

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

Effect of Vinylene Carbonate on Graphite Anode Cycling Efficiency Effect of Vinylene Carbonate on Graphite Anode Cycling Efficiency Title Effect of Vinylene Carbonate on Graphite Anode Cycling Efficiency Publication Type Journal Article Year of Publication 2009 Authors Ridgway, Paul L., Honghe Zheng, Xiangyun Song, Gao Liu, Philip N. Ross, and Vincent S. Battaglia Journal Electrochemical Society Volume 19 Start Page 51 Issue 25 Pagination 51-57 Abstract 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.

311

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

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

312

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

SciTech Connect (OSTI)

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.

Hassan, Yassin

2013-10-22T23:59:59.000Z

313

Research News November 2014  

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

They also toured several of NETL's laboratories, including the Fuel Cells Lab, Chemical Looping Reactor, and the Supercomputer and Visualization Center. There, NETL's researchers...

314

Reactor Project Presses Ahead Despite Protests  

Science Journals Connector (OSTI)

...existing research reactors-in Berlin, Braunschweig, Jiilich, Geesthacht, and Munich were built in the 1950s and '60s and, even...the United States and 15 reactors abroad (including one in Geesthacht, Germany) have so far been converted to low-enriched uranium...

Robert Koenig

1995-08-04T23:59:59.000Z

315

Photocatalytic reactor  

DOE Patents [OSTI]

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

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

1999-01-19T23:59:59.000Z

316

Efficient Production of 5-Hydroxymethylfurfural (HMF) from d-Fructose and Inulin with Graphite Derivatives as the Catalysts  

Science Journals Connector (OSTI)

The selective conversion of d-fructose and inulin to produce 5-hydroxymethyl furfural (HMF) is achieved in the presence of a catalytic amount of graphite derivatives such as graphite oxide (GO), reduced graphite ...

Guangxia Nie; Xinli Tong; Yangyang Zhang; Song Xue

2014-10-01T23:59:59.000Z

317

Hybrid adsorptive membrane reactor  

DOE Patents [OSTI]

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

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

2011-03-01T23:59:59.000Z

318

Advance Reactor Concepts Technical Review Panel Public Report  

Broader source: Energy.gov [DOE]

The Office of Nuclear Energy supports research and development for advanced reactor technologies. This report documents the results of the 2014 Technical Review Panel (TRP) review of seven advanced reactor concepts. The intent of the process was to identify R&D needs for advanced reactor concepts in order to inform Department of Energy (DOE) Office of Nuclear Energy R&D investment decisions.

319

Overview of Sandia National Laboratories pulse nuclear reactors  

SciTech Connect (OSTI)

Sandia National Laboratories has designed, constructed and operated bare metal Godiva-type and pool-type pulse reactors since 1961. The reactor facilities were designed to support a wide spectrum of research, development, and testing activities associated with weapon and reactor systems.

Schmidt, T.R. [Sandia National Labs., Albuquerque, NM (United States); Reuscher, J.A. [Texas A& M Univ., College Station, TX (United States)

1994-10-01T23:59:59.000Z

320

Effect of Graphitic Content on Carbon Supported Catalyst Performance  

SciTech Connect (OSTI)

The effect of graphitic content on carbon supported platinum catalysts was investigated in order to investigate its influence on catalyst performance. Four catalysts of varying surface areas and graphitic content were analyzed using XPS, HREELS, and tested using RDE experiments. The catalysts were also heat treated at 150oC and 100%RH as means to uniformly age them. The heat treated samples were analyzed using the same methods to determine what changes had occurred due to this aging process. When compared to the BOL catalysts, heat treated catalysts displayed increased graphitic carbon and platinum metalic content, however they also showed depressed catalytic activity. The primary cause is still under investigation, though it is believed to be related to loss of amorphous carbon content.

Patel, Anant; Artyushkova, Kateryna; Atanassov, Plamen; Harvey, David; Dutta, Monica; Colbow, Vesna

2011-07-01T23:59:59.000Z

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


321

Effect of Graphitic Content on Carbon Supported Catalyst Performance  

SciTech Connect (OSTI)

The effect of graphitic content on carbon supported platinum catalysts was investigated in order to investigate its influence on catalyst performance. Four catalysts of varying surface areas and graphitic content were analyzed using XPS, HREELS, and tested using RDE experiments. The catalysts were also heat treated at 150 C and 100%RH as means to uniformly age them. The heat treated samples were analyzed using the same methods to determine what changes had occurred due to this aging process. When compared to the BOL catalysts, heat treated catalysts displayed increased graphitic carbon and platinum metallic content, however they also showed depressed catalytic activity. The primary cause is still under investigation, though it is believed to be related to loss of amorphous carbon content.

A. Patel; K. Artyushkova; P. Atanassov; David Harvey; M. Dutta; V. Colbow; S. Wessel

2011-07-01T23:59:59.000Z

322

Management of Naval Reactors' Cyber Security Program, OIG-0884  

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

Naval Reactors' Naval Reactors' Cyber Security Program DOE/IG-0884 April 2013 U.S. Department of Energy Office of Inspector General Office of Audits and Inspections Department of Energy Washington, DC 20585 April 12, 2013 MEMORANDUM FOR THE SECRETARY FROM: Gregory H. Friedman Inspector General SUBJECT: INFORMATION: Audit Report on "Management of Naval Reactors' Cyber Security Program" INTRODUCTION AND OBJECTIVE The Naval Reactors Program (Naval Reactors), an organization within the National Nuclear Security Administration, provides the military with safe and reliable nuclear propulsion plants to power warships and submarines. Naval Reactors maintains responsibility for activities supporting the United States Naval fleet nuclear propulsion systems, including research and

323

GEN-IV Reactors  

Science Journals Connector (OSTI)

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

Taek K. Kim

2013-01-01T23:59:59.000Z

324

The Netherlands Reactor Centre  

Science Journals Connector (OSTI)

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

S. WEINTROUB

1964-04-04T23:59:59.000Z

325

Concept development of rotating bed chemical looping combustion reactor:.  

E-Print Network [OSTI]

??In this research a new rotary chemical looping combustion (CLC) reactor was developed which is suitable for larger scales and solves some of the issues (more)

Hermans, C.W.M.

2013-01-01T23:59:59.000Z

326

Low Cost SiOx-Graphite and Olivine Materials | Department of...  

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

Cost SiOx-Graphite and Olivine Materials Low Cost SiOx-Graphite and Olivine Materials 2010 DOE Vehicle Technologies and Hydrogen Programs Annual Merit Review and Peer Evaluation...

327

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

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

Low-Cost 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...

328

Enhanced performance of graphite anode materials by AlF3 coating...  

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

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

329

Delamination fracture toughness of a unidirectional graphite/epoxy composite  

E-Print Network [OSTI]

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...

Hulsey, Roy Charles

2012-06-07T23:59:59.000Z

330

Electronic properties of graphite in tilted magnetic fields  

SciTech Connect (OSTI)

The minimal nearest-neighbor tight-binding model with the Peierls substitution is employed to describe the electronic structure of Bernal-stacked graphite subject to tilted magnetic fields. We show that while the presence of the in-plane component of the magnetic field has a negligible effect on the Landau level structure at the K point of the graphite Brillouin zone, at the H point it leads to the experimentally observable splitting of Landau levels which grows approximately linearly with the in-plane field intensity.

Goncharuk, Nataliya A.; Smr?ka, Ludvk [Institute of Physics, Academy of Science of the Czech Republic, v. v. i., Cukrovarnick 10, 162 53 Praha 6 (Czech Republic)

2014-05-15T23:59:59.000Z

331

Preparation and thermal properties of expanded graphite/paraffin/organic montmorillonite composite phase change material  

Science Journals Connector (OSTI)

Expanded graphite (EG)/paraffin/organic montmorillonite (OMMT) composite phase change material (PCM) was prepared by using melt...

Hongtao Kao; Min Li; Xuewen Lv; Jinmiao Tan

2012-01-01T23:59:59.000Z

332

Research departments Materials Research Department  

E-Print Network [OSTI]

research reactor and X- radiation from the synchrotron facilities in Hamburg and Grenoble. In this con-parameter experiments in RERAF. Systems Analysis Department The objective of the research is to de- velop and apply are systems reliability, organisation, toxi- cology, informatics, simulation methods, work studies, economics

333

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

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

334

The dependence on excitation energy of the D-mode in graphite and carbon nanotubes  

E-Print Network [OSTI]

The dependence on excitation energy of the D-mode in graphite and carbon nanotubes C. Thomsen, S of carbon nanotubes as well. The corresponding calculated D-mode shift is shown explicitely in the nanotube is seven times narrower than in graphite as agrees with experiment. Graphite was first

Nabben, Reinhard

335

Role of organic matter in the Proterozoic Oklo natural fission reactors, Gabon, Africa  

SciTech Connect (OSTI)

Of the sixteen known Oklo and the Bangombe natural fission reactors (hydrothermally altered elastic sedimentary rocks that contain abundant uraninite and authigenic clay minerals), reactors 1 to 6 at Oklo contain only traces of organic matter, but the others are rich in organic substances. Reactors 7 to 9 are the subjects of this study. These organic-rich reactors may serve as time-tested analogues for anthropogenic nuclear-waste containment strategies. Organic matter helped to concentrate quantities of uranium sufficient to initiate the nuclear chain reactions. Liquid bitumen was generated from organic matter by hydrothermal reactions during nuclear criticality. The bitumen soon became a solid, consisting of polycyclic aromatic hydrocarbons and an intimate mixture of cryptocrystalline graphite, which enclosed and immobilized uraninite and the fission-generated isotopes entrapped in uraninite. This mechanism prevented major loss of uranium and fission products from the natural nuclear reactors for 1.2 b.y. 24 refs., 4 figs.

Nagy, B.; Rigali, M.J. [Univ. of Arizona, Tucson (United States)] [Univ. of Arizona, Tucson (United States); Gauthier-Lafaye, F. [Centre de Geochemie de la Surface, Strasbourg (France)] [Centre de Geochemie de la Surface, Strasbourg (France); Holliger, P. [Centre d`Etudes Nucleaires de Cadarache (France)] [Centre d`Etudes Nucleaires de Cadarache (France); Mossman, D.J. [Mount Allison Univ., Sackville, New Brunswick (Canada)] [Mount Allison Univ., Sackville, New Brunswick (Canada); Leventhal, J.S. [Geological Survey, Denver, CO (United States)] [Geological Survey, Denver, CO (United States)

1993-07-01T23:59:59.000Z

336

SRS Small Modular Reactors  

SciTech Connect (OSTI)

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

None

2012-04-27T23:59:59.000Z

337

SRS Small Modular Reactors  

ScienceCinema (OSTI)

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

None

2014-05-21T23:59:59.000Z

338

Corrosion quantification test for flanges with graphite-based gaskets  

SciTech Connect (OSTI)

The substitution of asbestos with nonasbestos fiber-reinforced materials in some industrial plants has caused corrosion problems in flanges and valves. A novel corrosion apparatus, the Corrosion Qualification Test, quantified corrosion and gives preliminary results of tests on flexible graphite-based gasket products.

Mameri, N.; Piron, D.L.; Bouzid, A.; Derenne, M.; Marchand, L.; Birembaut, Y.

2000-04-01T23:59:59.000Z

339

Tuning electrical properties of graphite oxide by plasma  

Science Journals Connector (OSTI)

...Tuning electrical properties of graphite oxide by plasma Hongfei Zhu Deyang Ji Lang Jiang Huanli Dong Wenping...consecutively by treating samples with ammonia and hydrogen plasma. When altering ammonia plasma time from 10 to 4.5min, large area (greater...

2013-01-01T23:59:59.000Z

340

Lithium intercalated graphite : experimental Compton profile for stage one  

E-Print Network [OSTI]

L-301 Lithium intercalated graphite : experimental Compton profile for stage one G. Loupias, J différence des profils Compton est compatible avec un transfert total de l'électron de conduction du lithium électronique due à l'insertion. Abstract. 2014 Electron momentum distribution of the first stage lithium

Paris-Sud XI, Université de

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


341

Nuclear reactor  

DOE Patents [OSTI]

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

Thomson, Wallace B. (Severna Park, MD)

2004-03-16T23:59:59.000Z

342

Activation analysis of concrete and graphite in the experimental reactor RUS  

Science Journals Connector (OSTI)

......nature of the radiological waste generated during decommissioning...flow going to different waste repositories may substantially...of decommissioning and storage. The present work has...developed for determining the long-term induced activity in the...mass flows going to the waste repositories and hence......

M. Cometto; D. Ridikas; M. C. Aubert; F. Damoy; D. Ancius

2005-12-20T23:59:59.000Z

343

The High Flux Isotope Reactor at Oak Ridge National Laboratory  

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

The High Flux Isotope Reactor at ORNL The High Flux Isotope Reactor at ORNL Aerial of the High Flux Isotope Reactor Site The High Flux Isotope Reactor site is located on the south side of the ORNL campus and is about a three-minute drive from her sister neutron facility, the Spallation Neutron Source. Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for research in the United States, and it provides one of the highest steady-state neutron fluxes of any research reactor in the world. The thermal and cold neutrons produced by HFIR are used to study physics, chemistry, materials science, engineering, and biology. The intense neutron flux, constant power density, and constant-length fuel cycles are used by more than 500 researchers each year for neutron scattering research into

344

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

SciTech Connect (OSTI)

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.

Rich Johnson; Yu-Hsin Tung; Hiroyuki Sato

2011-07-01T23:59:59.000Z

345

Oxygen reduction on a graphite paste and a catalyst loaded graphite paste electrode  

SciTech Connect (OSTI)

Oxygen reduction was studied in basic solution at a graphite paste electrode (GPE). The GPE was used as the disk of a rotating ring disk electrode (RRDE) and experiments were done using the voltage scan technique. The enhancements afforded by catalysts applied to the GPE were also studied. Oxygen reduction on a GPE was shown to be a two-electron process resulting in the formation of peroxide. The Tafel slope (plotted as potential versus log(i/sub l/ x i/(i/sub l/ - i))) was 180 mV. The presence of gold, silver, or platinum on the GPE shifted the oxygen reduction wave approximately 800 mV in the anodic direction. Comparison of the data on a metal catalyzed GPE to the solid metal electrode showed that the former electrode produced a greater fraction of peroxide as product than did the latter. Silver and gold catalyzed GPEs gave Tagel slopes of about 120 mV. The intermediate catalysis of iron and cobalt porphyrin was also examined. While the cobalt porphyrin catalyzed oxygen reduction at a more anodic potential than the iron porphyrin, the latter appeared more active in reacting the peroxide formed as the product of the disk reaction.

DiMarco, D.M.

1980-03-01T23:59:59.000Z

346

Articulated limiter blade for a tokamak fusion reactor  

DOE Patents [OSTI]

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.

Doll, D.W.

1982-10-21T23:59:59.000Z

347

Large passive pressure tube light water reactor with voided calandria  

SciTech Connect (OSTI)

A reactor concept has been developed that can survive loss-of-coolant accidents (LOCAs) without scram and without replenishing primary coolant inventory while maintaining safe temperature limits on the fuel and pressure tube. The proposed concept is a pressure tube reactor of similar design to Canada deuterium uranium reactors but differing in three key aspects. First, a solid silicon carbide-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low-pressure gas instead of heavy water moderator, and this normally voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation while during a LOCA or loss of heat sink accident, it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. The fuel elements can operate under post-critical-heat-flux conditions even at full power without exceeding fuel design limits. The heterogeneous arrangement of the fuel and moderator ensures a negative void coefficient under all circumstances. Although light water is used as both coolant and moderator, the reactor exhibits a high degree of neutron thermalization and a large prompt neutron lifetime, similar to D{sub 2}O-moderated cores. Moreover, the extremely large neutron migration length results in a strongly coupled core with a flat thermal flux profile and inherent stability against xenon spatial oscillations.

Hejzlar, P.; Todreas, N.E.; Driscoll, M.J. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering

1996-02-01T23:59:59.000Z

348

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)

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 Wisconsins 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature up to 800C.

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

2006-09-01T23:59:59.000Z

349

Development of a system model for advanced small modular reactors.  

SciTech Connect (OSTI)

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.

Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

2014-01-01T23:59:59.000Z

350

Features of a subcritical nuclear reactor  

Science Journals Connector (OSTI)

Abstract A subcritical nuclear reactor is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. Using the MCNP5 code, a three-dimensional model of the subcritical reactor was developed to estimate the effective multiplication factor, the neutron spectra, and the total and thermal neutron fluences along the radial and axial axis. The MCNP5 results of the effective multiplication factor were compared with those obtained from the six-factor formula. The effective dose and the Ambient dose equivalent, at three sites outside the reactor, were estimated; the Ambient dose equivalent was also measured and compared with the calculated values.

Hector Rene Vega-Carrillo; Isvi Ruben Esparza-Garcia; Alvaro Sanchez

2015-01-01T23:59:59.000Z

351

Small Modular Nuclear Reactors | Department of Energy  

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

Reactor Technologies » Small Modular Reactor Technologies » Small Modular Nuclear Reactors Small Modular Nuclear Reactors Cutaway of 2-Unit Generation mPower SMR Installation. | © 2012 Generation mPower LLC. All Rights Reserved. Reprinted with permission. Cutaway of 2-Unit Generation mPower SMR Installation. | © 2012 Generation mPower LLC. All Rights Reserved. Reprinted with permission. The development of clean, affordable nuclear power options is a key element of the Department of Energy's Office of Nuclear Energy (DOE-NE) Nuclear Energy Research and Development Roadmap. As a part of this strategy, a high priority of the Department has been to help accelerate the timelines for the commercialization and deployment of small modular reactor (SMR) technologies through the SMR Licensing Technical Support program. Begun

352

Controlled Fusion and Reactors of the Tokamak Type  

Science Journals Connector (OSTI)

Research on fusion reactor problems has increased dramatically as the plasma physics of magnetic confinement continues to make substantial progress. As part of this research several studies (16) have been comple...

Robert W. Conn

1977-01-01T23:59:59.000Z

353

UCLA program in reactor studies: The ARIES tokamak reactor study  

SciTech Connect (OSTI)

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

Not Available

1991-01-01T23:59:59.000Z

354

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

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

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

355

Advanced Test Reactor National Scientific User Facility  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

356

Atomic Heat of Graphite between 1 and 20K  

Science Journals Connector (OSTI)

The atomic heat of pure artificial polycrystalline graphite has been measured between 1 and 4K and 10 and 20K. Precautions were taken to ensure that no error was introduced by gas adsorbed on the graphite. Below 2K the atomic heat can be represented as the sum of a term proportional to T3 arising from lowfrequency lattice waves and a term proportional to T due to electrons: C=0.0325T3+0.031T millijoules/mole degree. Between 2.25K and 4.5K, C=0.115T2+0.031T-0.237 millijoules/mole degree; between 10 20K, C=0.208T2-6.8 millijoules/mole degree.

P. H. Keesom and N. Pearlman

1955-08-15T23:59:59.000Z

357

Rotational Transition of Incommensurate Kr Monolayers on Graphite  

Science Journals Connector (OSTI)

Synchrotron x-ray diffraction studies of the orientational epitaxy of krypton on single-crystal graphite are reported. The system displays a continuous transition from an aligned to a rotated orientation with increasing mean misfit in a manner predicted by a hexagonal-domain-wall model of the weakly incommensurate phase. A hysteresis at the critical misfit is observed. The results are discussed in light of current views of the structure of the incommensurate phase.

K. L. D'Amico; D. E. Moncton; E. D. Specht; R. J. Birgeneau; S. E. Nagler; P. M. Horn

1984-12-03T23:59:59.000Z

358

X-ray diffraction data for graphite to 20 GPa  

Science Journals Connector (OSTI)

X-ray diffraction data have been obtained on polycrystalline graphite at pressures up to 20 GPa. A phase transition is observed at ?11 GPa, as evidenced by softening in the interlayer spacing and the observation of new diffraction lines. Below this pressure the variation of the lattice parameters a and c are compared with elastic stiffnesses obtained from ultrasonic measurements. A new value for C13 is proposed. The variation c(P) is compared to the recently proposed universal isotherm equation.

You Xiang Zhao and Ian L. Spain

1989-07-15T23:59:59.000Z

359

Ethylene on graphite: Heats of adsorption and phase diagram  

Science Journals Connector (OSTI)

Isosteric heats of adsorption (qst) of ethylene on Grafoil MAT have been measured calorimetrically in the region 98ethylene-graphite interaction potential, a value is obtained for the heat of adsorption at zero coverage, qst(0)=20.40.2 kJ?l-1, at T=120 K. At this temperature, the effects of surface heterogeneity and of clustering of the adsorbed molecules are clearly delineated and thus permit an extrapolation to be made with reasonable certainty.

A. Inaba and J. A. Morrison

1986-09-01T23:59:59.000Z

360

Structure of the solid D2 bilayer on graphite  

Science Journals Connector (OSTI)

Low-energy electron-diffraction (LEED) measurements for a solid D2 bilayer physisorbed on graphite are used to investigate the bilayer structure and its azimuthal orientation relative to the substrate. The LEED spot positions are consistent with the mutually commensurate oblique unit cell inferred from neutron-diffraction measurement. Given the azimuthal orientation of the bilayer LEED spots, the possibility of a bilayer composed of two incommensurate layers with mutually modulated triangular lattices is ruled out.

Wei Liu and S. C. Fain; Jr.

1993-06-15T23:59:59.000Z

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


361

MICROMECHANICS IN CONTINOUS GRAPHITE FIBER/EPOXY COMPOSITES DURING CREEP  

SciTech Connect (OSTI)

Micro Raman spectroscopy and classic composite shear-lag models were used to analyze the evolution with time of fiber and matrix strain/stress around fiber breaks in planar model graphite fiber-epoxy matrix composites. Impressive agreements were found between the model predictions and the experimental results. The local matrix creep leads to an increase in the load transfer length around the break under a constant load. This increases the chance of fiber breakage in the neighboring intact fibers.

C. ZHOU; ET AL

2001-02-01T23:59:59.000Z

362

Catalytic Graphitization of Carbon Aerogels by Transition Metals  

Science Journals Connector (OSTI)

Catalytic Graphitization of Carbon Aerogels by Transition Metals ... Carbon aerogels and Cr-, Fe-, Co-, and Ni-containing carbon aerogels were obtained by pyrolysis, at temperatures between 500 and 1800 C, of the corresponding aerogels prepared by the sol?gel method from polymerization of resorcinol with formaldehyde. ... Results obtained show that carbon aerogels are, essentially, macroporous materials that maintain large pore volumes even after pyrolysis at 1800 C. ...

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

2000-03-24T23:59:59.000Z

363

Attrition reactor system  

DOE Patents [OSTI]

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

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

1993-09-28T23:59:59.000Z

364

Elementary Reactor Physics  

Science Journals Connector (OSTI)

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

J. F. HILL

1962-02-10T23:59:59.000Z

365

Colliding Beam Fusion Reactors  

Science Journals Connector (OSTI)

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

Norman Rostoker; Artan Qerushi; Michl Binderbauer

2003-06-01T23:59:59.000Z

366

Graphite having improved thermal stress resistance and method of preparation  

DOE Patents [OSTI]

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.

Kennedy, Charles R. (Oak Ridge, TN)

1980-01-01T23:59:59.000Z

367

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)

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.

PONCET, Bernard R.

2003-02-27T23:59:59.000Z

368

Basic Engineering Research for D&D of R. Reactor Storage Pond Sludge: Electrokinetics, Carbon Dioxide Extraction, and Supercritical Water Oxidation  

SciTech Connect (OSTI)

Collaborating researchers at the University of South Carolina (USC), Clemson University (CU), and the Savannah River Site (SRS) are investigating the fundamentals of a combined extraction and destruction process for the decontamination and decommissioning (D&D) of PCB-contaminated materials as found at DOE sites. Currently, the volume of PCBs and PCB contaminated wastes at DOE sites nationwide is approximately 19,000 m3. While there are a number of existing and proposed processes for the recovery and/or destruction of these persistent 4 pollutants, none has emerged as the preferred choice. Therefore, this research focuses on combining novel processes to solve the problem. The research objectives are to investigate benign dense-fluid extraction with either carbon dioxide (USC) or hot water (CU), followed by destruction of the extracted PCBs via either electrochemical (USC) or hydrothermal (CU) oxidation. Based on the results of these investigations, a combined extraction and destruction process that incorporates the most successful elements of the various processes will be recommended for application to contaminated DOE sites.

Hamilton, Edward A.; Bruce, David A.; Oji, Lawrence; White, Ralph E.; Matthews, Michael A.; Thies, Mark C.

1999-06-01T23:59:59.000Z

369

Proceedings of the conference on electrochemistry of carbon allotropes: Graphite, fullerenes and diamond  

SciTech Connect (OSTI)

This conference provided an opportunity for electrochemists, physicists, materials scientists and engineers to meet and exchange information on different carbon allotropes. The presentations and discussion among the participants provided a forum to develop recommendations on research and development which are relevant to the electrochemistry of carbon allotropes. The following topics which are relevant to the electrochemistry of carbon allotropes were addressed: Graphitized and disordered carbons, as Li-ion intercalation anodes for high-energy-density, high-power-density Li-based secondary batteries; Carbons as substrate materials for catalysis and electrocatalysis; Boron-doped diamond film electrodes; and Electrochemical characterization and electrosynthesis of fullerenes and fullerene-type materials. Abstracts of the presentations are presented.

Kinoshita, K. [ed.] [Lawrence Berkeley National Lab., CA (United States); Scherson, D. [ed.] [Case Western Reserve Univ., Cleveland, OH (United States)

1998-02-01T23:59:59.000Z

370

Prospects for spheromak fusion reactors  

Science Journals Connector (OSTI)

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

T. K. Fowler; D. D. Hua

1995-06-01T23:59:59.000Z

371

Light Water Reactor Sustainability Nondestructive Evaluation for Concrete  

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

Nondestructive Evaluation for Nondestructive Evaluation for Concrete Research and Development Roadmap Light Water Reactor Sustainability Nondestructive Evaluation for Concrete Research and Development Roadmap Materials issues are a key concern for the existing nuclear reactor fleet as material degradation can lead to increased maintenance, increased downtown, and increased risk. Extending reactor life to 60 years and beyond will likely increase susceptibility and severity of known forms of degradation. Additionally, new mechanisms of materials degradation are also possible. The purpose of the US Department of Energy Office of Nuclear Energy's Light Water Reactor Sustainability (LWRS) Program is to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend

372

Improved vortex reactor system  

DOE Patents [OSTI]

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

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

1995-01-01T23:59:59.000Z

373

BASIC ENGINEERING RESEARCH FOR D&D OF R REACTOR STORAGE POND SLUDGE: ELECTROKINETICS, CARBON DIOXIDE EXTRACTION, AND SUPERCRITICAL WATER OXIDATION  

SciTech Connect (OSTI)

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&D operations at DOE sites across the country. Currently, the volume of these wastes is approximately 23,500 m3, and the majority of these wastes (i.e., almost 19,000 m3) consist of PCBs and PCB-contaminated materials. Further, additional PCB-contaminated waste will be generated during D&D operations in the future. The standard process for destruction of this waste is incineration, which generates secondary waste that must be disposed, and the TSCA incinerator at Oak Ridge has an uncertain future. Beyond incineration, no proposed process for the recovery and/or destruction of these persistent pollutants has emerged as the preferred choice for DOE cleanup. The main objective of the project was to investigate and develop a deeper understanding of the thermodynamic and kinetic reactions involved in the extraction and destruction of polychlorinated biphenyls (PCBs) from low-level mixed waste solid matrices in order to provide data that would permit the design of a combined-cycle extraction/destruction process. The specific research objectives were to investigate benign dense-fluid extraction with either carbon dioxide (USC) or hot water (CU), followed by destruction of the extracted PCBs via either electrochemical (USC) or hydrothermal (CU) oxidation. Two key advantages of the process are that it isolates and concentrates the PCBs from the solid matrices (thereby reducing waste volume greatly and removing the remaining low-level mixed waste from TSCA control), and little, if any, secondary solvent or solid wastes are generated. This project was a collaborative effort involving the University of South Carolina (USC), Clemson University (CU), and Westinghouse Savannah River Company (WSRC) (including the Savannah River Technology Center, Facilities Decommissioning Division and Regulatory Compliance). T he project was directed and coordinated by the South Carolina Universities Research and Education Foundation (SCUREF), a consortium of the four public research universities in South Carolina. The original plan was to investigate two PCB extraction processes (supercritical carbon dioxide and hot, pressurized water) and two PCB destruction processes (electrochemical oxidation and hydrothermal oxidation). However, at approximately the mid-point of the three year project, it was decided to focus on the more promising extraction process (supercritical carbon dioxide) and the more promising destruction process (supercritical water oxidation). This decision was taken because the investigation of two processes simultaneously by each university was stretching resources too thin, and because the electrochemical oxidation process needed more concentrated research before it would be ready for application to PCB destruction. The solid matrix chosen for experimental work was Toxi-dry, a commonly used adsorbent made from plant material that is used in cleanup of spills and/or liquid solvents. The Toxi-dry was supplied by the research team member from the Facilities Decommissioning Division, WSRC. This adsorbent is a major component of job control waste.

Matthews, Michael A.; Bruce,David; Davis,Thomas; Thies, Mark; Weidner, John; White, Ralph

2001-12-31T23:59:59.000Z

374

Reactor vessel support system  

DOE Patents [OSTI]

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

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

1982-01-01T23:59:59.000Z

375

Electron spin resonance study of proton-irradiation-induced defects in graphite  

SciTech Connect (OSTI)

Electron spin resonance measurements of proton-irradiated graphite have revealed detailed nature of proton-irradiation-induced defects. Our results indicate that proton-irradiation creates confined defect regions of a metallic island surrounded by an insulating magnetic region which ''isolates'' the metallic island inside from the metallic graphite background outside. We have thus come up with a picture of phase separation in proton-irradiated graphite comprising three regions of distinct electrical and magnetic properties.

Won Lee, Kyu; Kweon, H.; Kweon, J. J.; Lee, Cheol Eui [Department of Physics and Institute for Nano Science, Korea University, Seoul 136-713 (Korea, Republic of)

2010-02-15T23:59:59.000Z

376

Laser-induced dehydration of graphite oxide coatings on polymer substrates  

SciTech Connect (OSTI)

Nanosized graphite has been oxidized by the Hummers method to give high quality graphite oxide. This reaction is characterized by a very fast kinetic behavior and a high yield. The produced graphite oxide has been conveniently used to pattern graphene by using a standard photolithographic method, and the resulting systems have been characterized by optical microscopy (OM), scanning electron microscopy (SEM) and by Fourier transform infrared spectroscopy (FT-IR) and Visible-Near Infrared spectroscopy (Vis-NIR)

Longo, Angela, E-mail: angela.longo@cnr.it; Palomba, Mariano; Carotenuto, Gianfranco; Nicolais, Luigi [Institute for Composite and Biomedical Materials, National Research Council, Viale Kennedy, 54, Mostra d'Oltremare Padiglione 20, 80125 Napoli (Italy); Orabona, Emanuele; Maddalena, Pasqualino [Department of Physics, University of Naples, Federico II, via cintia, 80126, Naples, Italy and SPIN Institute, National Research Council, UOS Naples, via cintia, 80126, Naples (Italy); Ambrosio, Antonio [SPIN Institute, National Research Council, UOS Naples, via cintia, 80126, Naples (Italy)

2014-05-15T23:59:59.000Z

377

From Green Aerogels to Porous Graphite by Emulsion Gelation of Acrylonitrile  

Science Journals Connector (OSTI)

From Green Aerogels to Porous Graphite by Emulsion Gelation of Acrylonitrile ... Ambient pressure drying of wet-gels from copolymeri-zation of acrylonitrile and bifunctional cross-linkers either in toluene or in H2O-based emulsions yields structurally similar PAN aerogels, which are converted to monolithic porous carbons and graphite. ... Porous carbons, including carbon (C-) aerogels, are technologically important materials, while polyacrylonitrile (PAN) is the main industrial source of graphite fiber. ...

Anand G. Sadekar; Shruti S. Mahadik; Abhishek N. Bang; Zachary J. Larimore; Clarissa A. Wisner; Massimo F. Bertino; A. Kaan Kalkan; Joseph T. Mang; Chariklia Sotiriou-Leventis; Nicholas Leventis

2011-12-14T23:59:59.000Z

378

Reactor water cleanup system  

DOE Patents [OSTI]

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

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

1994-01-01T23:59:59.000Z

379

Reactor water cleanup system  

DOE Patents [OSTI]

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

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

1994-12-20T23:59:59.000Z

380

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

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

WORK Identify suitable graphite materials for anodes that meet the requirement for low cost and long cycle life. Fabricate half cells (Ligraphite) and Li-ion (graphiteolivine)...

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


381

Properties of copper/graphite/carbon nanotubes composite reinforced by carbon nanotubes  

Science Journals Connector (OSTI)

Electroless Cu plating was used for flake G powder and CNTs, CuGCNTs (copper/graphite/carbon nanotubes) composites were manufactured by means of powder...

Xin-Ying Liu; Xiong-Zhi Xiang; Fei Niu; Xiao-Jun Bai

2013-06-01T23:59:59.000Z

382

Thermal Characterization of Graphitic Carbon Foams for Use in Thermal Storage Applications.  

E-Print Network [OSTI]

?? Highly conductive graphitic foams are currently being studied for use as thermal conductivity enhancers (TCEs) in thermal energy storage (TES) systems. TES systems store (more)

Drummond, Kevin P.

2012-01-01T23:59:59.000Z

383

E-Print Network 3.0 - ams graphite target Sample Search Results  

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

F. J Appl Phys Lett 1976;29:118-20. Improving colloidal graphite for electromagnetic interference... January 2003; received in revised form 30 January 2003; accepted 30...

384

Graphitic Electrical Contacts to Metallic Single Walled Carbon Nanotubes Using Pt Electrodes  

E-Print Network [OSTI]

NANO LETTERS Graphitic Electrical Contacts to Metallicof the interfacial electrical resistance between nano- tubesprovide excellent electrical contacts to many types of

Collins, Philip G

2009-01-01T23:59:59.000Z

385

Three-dimensional imaging and precision metrology for liquid-salt-cooled reactors  

SciTech Connect (OSTI)

The liquid-salt-cooled very high temperature reactor, also called the Advanced High-Temperature Reactor (AHTR), is a new large high-temperature reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. The AHTR will require refueling, in-service inspection, and maintenance (RIM) with supporting instrumentation systems. The fluoride salts that are being evaluated as potential reactor coolants have melting points between 350 and 500 deg. C, values that imply minimum RIM temperatures between 400 and 550 deg. C. These salts are transparent over a wider range of the light spectrum than is water. The high temperatures, the optical characteristics of the coolant, and advances in metrology may enable the use of lasers to create three-dimensional images of the reactor interior to assist refueling, monitor vibrations in components, map fluid flow, and enable inspections of internal reactor components. A description of the reactor and an initial evaluation of the use of optical techniques for AHTR instrumentation are provided. (authors)

Forsberg, C. W. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6165 (United States); Varma, V. K.; Burgess, T. W. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6304 (United States)

2006-07-01T23:59:59.000Z

386

Heterogeneous Recycling in Fast Reactors  

SciTech Connect (OSTI)

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.

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

2012-07-30T23:59:59.000Z

387

Micro -Thermonuclear AB-Reactors for Aerospace  

E-Print Network [OSTI]

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).

Alexander Bolonkin

2007-01-08T23:59:59.000Z

388

Modeling for Anaerobic Fixed-Bed Biofilm Reactors  

SciTech Connect (OSTI)

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.

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

1989-06-01T23:59:59.000Z

389

Basic Engineering Research for D and D of R Reactor Storage Pond Sludge: Electrokinetics, Carbon Dioxide Extraction, and Supercritical Water Oxidation  

SciTech Connect (OSTI)

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.

Michael A. Matthews; David A. Bruce,; Thomas A. Davis; Mark C. Thies; John W. Weidner; Ralph E. White

2002-04-01T23:59:59.000Z

390

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

SciTech Connect (OSTI)

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).

B. Boer; A. M. Ougouag

2010-09-01T23:59:59.000Z

391

SHARP: Reactor Performance and Safety Simulation Suite  

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

SHARP SHARP Argonne National Laboratory's Reactor Performance and Safety Simulation Suite SHARP could save millions in nuclear reactor design and development... The Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite of codes enables virtual design and engineering of nuclear plant behavior that would be impractical from a traditional experimental approach. ...by leveraging the computational power of one of the world's most powerful supercomputers. Exploiting the power of Argonne Leadership Computing Facility's near-petascale computers, researchers have developed a set of simulation tools that provide a highly detailed description of the reactor core and the nuclear plant behavior. This enables the efficient and precise design of tomorrow's safe and clean nuclear energy sources.

392

Development of a graphite probe calorimeter for absolute clinical dosimetry  

SciTech Connect (OSTI)

The aim of this work is to present the numerical design optimization, construction, and experimental proof of concept of a graphite probe calorimeter (GPC) conceived for dose measurement in the clinical environment (U.S. provisional patent 61/652,540). A finite element method (FEM) based numerical heat transfer study was conducted using a commercial software package to explore the feasibility of the GPC and to optimize the shape, dimensions, and materials used in its design. A functioning prototype was constructed inhouse and used to perform dose to water measurements under a 6 MV photon beam at 400 and 1000 MU/min, in a thermally insulated water phantom. Heat loss correction factors were determined using FEM analysis while the radiation field perturbation and the graphite to water absorbed dose conversion factors were calculated using Monte Carlo simulations. The difference in the average measured dose to water for the 400 and 1000 MU/min runs using the TG-51 protocol and the GPC was 0.2% and 1.2%, respectively. Heat loss correction factors ranged from 1.001 to 1.002, while the product of the perturbation and dose conversion factors was calculated to be 1.130. The combined relative uncertainty was estimated to be 1.4%, with the largest contributors being the specific heat capacity of the graphite (type B, 0.8%) and the reproducibility, defined as the standard deviation of the mean measured dose (type A, 0.6%). By establishing the feasibility of using the GPC as a practical clinical absolute photon dosimeter, this work lays the foundation for further device enhancements, including the development of an isothermal mode of operation and an overall miniaturization, making it potentially suitable for use in small and composite radiation fields. It is anticipated that, through the incorporation of isothermal stabilization provided by temperature controllers, a subpercent overall uncertainty will be achieved.

Renaud, James; Seuntjens, Jan; Sarfehnia, Arman [Medical Physics Unit, McGill University, Montreal, Quebec H3G 1A4 (Canada); Marchington, David [Ionizing Radiation Standards, National Research Council of Canada, Ottawa, Ontario K1A 0R6 (Canada)

2013-02-15T23:59:59.000Z

393

Structural study of monolayer cobalt phthalocyanine adsorbed on graphite  

E-Print Network [OSTI]

We present microscopic investigations on the two-dimensional arrangement of cobalt phthalocyanine molecules on a graphite (HOPG) substrate in the low coverage regime. The initial growth and ordering of molecular layers is revealed in high resolution scanning tunneling microscopy (STM). On low coverages single molecules orient mostly along one of the substrate lattice directions, while they form chains at slightly higher coverage. Structures with two different unit cells can be found from the first monolayer on. A theoretical model based on potential energy calculations is presented, which relates the two phases to the driving ordering forces.

Scheffler, M; Baumann, D; Schlegel, R; Hnke, T; Toader, M; Bchner, B; Hietschold, M; Hess, C

2014-01-01T23:59:59.000Z

394

Interaction of graphite with a hot, dense deuterium plasma  

SciTech Connect (OSTI)

The erosion of ATJ-S graphite caused by a hot, dense deuterium plasma has been investigated experimentally. The plasma was produced in an electromagnetic shock tube. Plasma characteristics were typically: ion temperature approx. = 800 eV (approx. 1 x 10/sup 7/ /sup 0/K), number density approx. = 10/sup 16//cm/sup 3/, and transverse magnetic field approx. = 1 tesla. The energetic ion flux, phi, to the sample surfaces was approx. 10/sup 23/ ions/cm/sup 2/-sec for a single pulse duration of approx. 0.1 usec. Sample surfaces were metallographically prepared and examined with a scanning electron microscope before and after exposure.

Desko, J.C. Jr.

1980-01-01T23:59:59.000Z

395

Method of fabricating silicon carbide coatings on graphite surfaces  

DOE Patents [OSTI]

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.

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

1994-07-26T23:59:59.000Z

396

Method of fabricating silicon carbide coatings on graphite surfaces  

DOE Patents [OSTI]

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 1600.degree.C. which transforms the coating to silicon carbide.

Varacalle, Jr., Dominic J. (Idaho Falls, ID); Herman, Herbert (Port Jefferson, NY); Burchell, Timothy D. (Oak Ridge, TN)

1994-01-01T23:59:59.000Z

397

Manhattan Project: Final Reactor Design and X-10, 1942-1943  

Office of Scientific and Technical Information (OSTI)

Schematic of the X-10 Graphite Reactor, Oak Ridge FINAL REACTOR DESIGN AND X-10 Schematic of the X-10 Graphite Reactor, Oak Ridge FINAL REACTOR DESIGN AND X-10 (Met Lab and Oak Ridge [Clinton], 1942-1943) Events > The Plutonium Path to the Bomb, 1942-1944 Production Reactor (Pile) Design, 1942 DuPont and Hanford, 1942 CP-1 Goes Critical, December 2, 1942 Seaborg and Plutonium Chemistry, 1942-1944 Final Reactor Design and X-10, 1942-1943 Hanford Becomes Operational, 1943-1944 Before any plutonium could be chemically separated from uranium for a bomb, however, that uranium would first have to be irradiated in a production pile. CP-1 had been a success as a scientific experiment, but the pile was built on such a small scale that recovering any significant amounts of plutonium from it was impractical. In the fall of 1942, scientists of the Met Lab had decided to build a second Fermi pile at Argonne as soon as his experiments on the first were completed and to proceed with the "Mae West" design for a helium-cooled production pile as well. When DuPont engineers assessed the Met Lab's plans in the late fall, they agreed that helium should be given first priority. They placed heavy water second and urged an all-out effort to produce more of this highly effective moderator. Bismuth and water were ranked third and fourth in DuPont's analysis. Priorities began to change when Enrico Fermi's CP-1 calculations demonstrated a higher value for the neutron reproduction factor k (for a theoretical reactor of infinite size) than anyone had anticipated. Met Lab scientists concluded that a water-cooled pile was now feasible. Crawford Greenewalt, head of the DuPont effort, continued, however, to support helium cooling.

398

Analysis of Granular Flow in a Pebble-Bed Nuclear Reactor  

E-Print Network [OSTI]

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.

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

2006-02-16T23:59:59.000Z

399

Analysis of granular flow in a pebble-bed nuclear reactor  

Science Journals Connector (OSTI)

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 440000 frictional, viscoelastic 6-cm-diam spheres draining in a cylindrical vessel of diameter 3.5m and height 10m with bottom funnels angled at 30 or 60. 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.

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

2006-08-24T23:59:59.000Z

400

Power Reactor Progress  

Science Journals Connector (OSTI)

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

1957-02-25T23:59:59.000Z

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


401

Improved vortex reactor system  

DOE Patents [OSTI]

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

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

1995-05-09T23:59:59.000Z

402

AEC Pushes Fusion Reactors  

Science Journals Connector (OSTI)

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

1955-10-10T23:59:59.000Z

403

Tokamak reactor first wall  

DOE Patents [OSTI]

This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

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

1984-11-20T23:59:59.000Z

404

Theoretical Description of the STM Images of Alkanes and Substituted Alkanes Adsorbed on Graphite  

E-Print Network [OSTI]

Theoretical Description of the STM Images of Alkanes and Substituted Alkanes Adsorbed on Graphite the scanning tunneling microscopy (STM) images of molecules adsorbed on graphite. The model is applicable diffuse virtual orbitals of the adsorbed molecules, despite being much farther in energy from the Fermi

Goddard III, William A.

405

A first order two-dimensional melting transition : methane adsorbed on (0001) graphite (*)  

E-Print Network [OSTI]

L-543 A first order two-dimensional melting transition : methane adsorbed on (0001) graphite (*) A. Abstract. 2014 The variation of long range order of a registered solid submonolayer of methane adsorbed for various solid mono- layers adsorbed on graphite indicates that solid- fluid phase transitions are more

Paris-Sud XI, Université de

406

Emission characteristics and dynamics of C2 from laser produced graphite plasma  

E-Print Network [OSTI]

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

Harilal, S. S.

407

Adsorption of the first layer of argon on graphite (*) Laboratoire des Composs non St0153chiomtriques,  

E-Print Network [OSTI]

]. The cooling bath is solid nitrogen obtained by pumping. The adsorption cell, enclosed in an aluminium blockL-9 Adsorption of the first layer of argon on graphite (*) F. Millot Laboratoire des Composés non déterminé des isothermes d'adsorption d'argon sur le graphite entre 55 et 62 K. Nous proposons une

Boyer, Edmond

408

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

DOE Patents [OSTI]

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.

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

2008-06-10T23:59:59.000Z

409

Late-time particle emission from laser-produced graphite plasma S. S. Harilal,a)  

E-Print Network [OSTI]

-produced carbon plasma. Furthermore, in the design of Tokamaks for nuclear fusion, graphite has been proposedLate-time particle emission from laser-produced graphite plasma S. S. Harilal,a) A. Hassanein online 6 September 2011) We report a late-time "fireworks-like" particle emission from laser

Harilal, S. S.

410

Generation of graphitic soot by an urban fire storm  

SciTech Connect (OSTI)

The authors have obtained samples of aerosols deposited during the Hiroshima fire storm that was initiated by the atomic bomb detonated on August 6, 1945. These particles, which we extracted from streaks of black rain found on a plaster wall, are being studied. Initial studies show that the artifact appears to contain aerosol particles that may be representative of the aerosols that may lead to a nuclear winter. Aerosol generation in urban fire storms have been considered by studying these particles. The presence of graphite as a component of these particles is suggested by electron photomicrographs and has been confirmed using Raman spectroscopy, surface ionization mass spectroscopy, and electron scattering for chemical analysis. Several hypotheses are being considered to explain the presence of this form of carbon. Among these are generation in sooty clouds, in raindrops, in the interior of the first storm, and on the wall surface itself. The distribution of particle sizes suggests that the residence time of particles in the atmosphere would be long if they were not removed by rainout. An experimental and theoretical examination of the conditions necessary to produce graphitic soot is in progress.

Fields, D.E.; Cole, L.L.

1987-01-01T23:59:59.000Z

411

2012_AdvReactor_Factsheet.indd  

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

nuclear.gov nuclear.gov February 15, 2011 A ADVANCED REACTOR CONCEPTS DVANCED REACTOR CONCEPTS The U.S. Department of Energy's Offi ce of Nuclear Energy T he Advanced Reactor Concepts (ARC) program, an expanded version of the Generation IV research, development and demonstration (RD&D) program, sponsors research, development and deployment activities leading to further safety, technical, economical, and environmental advancements of innovative nuclear energy technologies. The Office of Nuclear Energy (NE) will pursue these advancements through RD&D activities at the Department of Energy (DOE) national laboratories and U.S. universities, as well as through collaboration with nuclear industry and international partners. These activities will focus on advancing

412

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

SciTech Connect (OSTI)

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.

Sun, Xiao-Yu [Department of Engineering Mechanics, School of Civil Engineering, Wuhan University, Wuhan 430072 (China); Key Laboratory of Hubei Province for Water Jet Theory and New Technology, Wuhan University, Wuhan 430072 (China); Wu, RunNi; Xia, Re [Key Laboratory of Hubei Province for Water Jet Theory and New Technology, Wuhan University, Wuhan 430072 (China); Chu, Xi-Hua; Xu, Yuan-Jie, E-mail: yj-xu@whu.edu.cn [Department of Engineering Mechanics, School of Civil Engineering, Wuhan University, Wuhan 430072 (China)

2014-05-05T23:59:59.000Z

413

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

Office of Scientific and Technical Information (OSTI)

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

414

The ARIES tokamak reactor study  

SciTech Connect (OSTI)

The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. 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. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

Not Available

1989-10-01T23:59:59.000Z

415

Portfolio for fast reactor collaboration  

SciTech Connect (OSTI)

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

Rippon, S.

1981-12-01T23:59:59.000Z

416

Impact of high energy ball milling on the nanostructure of magnetitegraphite and magnetitegraphitemolybdenum disulphide blends  

SciTech Connect (OSTI)

Different, partly complementary and partly redundant characterization methods were applied to study the transition of magnetite, graphite and MoS{sub 2} powders to mechanically alloyed nanostructures. The applied methods were: Transmission electron microscopy (TEM), Mssbauer spectroscopy (MS), Raman spectroscopy (RS), X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS). The main objective was to prepare a model material providing the essential features of a typical tribofilm forming during automotive braking, and to assess the impact of different constituents on sliding behaviour and friction level. Irrespective of the initial grain size, the raw materials were transferred to a nanocrystalline structure and mixed on a nanoscopic scale during high energy ball milling. Whereas magnetite remained almost unchanged, graphite and molybdenum disulphide were transformed to a nanocrystalline and highly disordered structure. The observed increase of the coefficient of friction was attributed to a loss of lubricity of the latter ingredient due to this transformation and subsequent oxidation. - Highlights: Characterization of microstructural changes induced by high energy ball milling Assessment of the potential of different characterization methods Impact of mechanical alloying on tribological performance revealed by tests Preparation of an artificial third body resembling the one formed during braking.

sterle, W., E-mail: Werner.oesterle@bam.de [BAM Federal Institute for Materials Research and Testing, 12200 Berlin (Germany); Orts-Gil, G.; Gross, T.; Deutsch, C. [BAM Federal Institute for Materials Research and Testing, 12200 Berlin (Germany); Hinrichs, R. [Instituto de Geocincias, UFRGS, P.O. Box 15001, 91501-970 Porto Alegre (Brazil); Vasconcellos, M.A.Z. [Instituto de Fsica, UFRGS, P.O. Box 15051, 91501-970 Porto Alegre (Brazil); Zoz, H.; Yigit, D.; Sun, X. [Zoz Group, 57482 Wenden (Germany)

2013-12-15T23:59:59.000Z

417

Handbook of Reactor Physics  

Science Journals Connector (OSTI)

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

PETER W. MUMMERY

1956-08-25T23:59:59.000Z

418

Fast reactor safety  

Science Journals Connector (OSTI)

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

R.D. SMITH

1979-08-23T23:59:59.000Z

419

Instrumentation of Nuclear Reactors  

Science Journals Connector (OSTI)

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

1953-03-07T23:59:59.000Z

420

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

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

Why is the High Flux Beam Reactor Being Decommissioned? Why is the High Flux Beam Reactor Being Decommissioned? HFBR The High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is being decommissioned because the Department of Energy (DOE) decided in 1999 that it would be permanently closed. The reactor was shut down in 1997 after tritium from a leak in the spent-fuel pool was found in the groundwater. The HFBR, which had operated from 1965 to 1996, was used solely for scientific research, providing neutrons for materials science, chemistry, biology, and physics experiments. The reactor was shut down for routine maintenance in November of 1996. In January 1997, tritium, a radioactive form of hydrogen and a by-product of reactor operations, was found in groundwater monitoring wells immediately south of the HFBR. The tritium

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


421

Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering Division  

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

Analysis Analysis Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Overview Current Projects Software Nuclear Plant Dynamics and Safety Nuclear Data Program Advanced Reactor Development Nuclear Waste Form and Repository Performance Modeling Nuclear Energy Systems Design and Development Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Reactor Physics and Fuel Cycle Analysis Bookmark and Share Reactor physics and fuel cycle analysis is a core competency of the Nuclear Engineering (NE) Division. The Division has played a major role in the design and analysis of advanced reactors, particularly liquid-metal-cooled reactors. NE researchers have concentrated on developing computer codes for

422

Simulation of a marine nuclear reactor  

SciTech Connect (OSTI)

A Nuclear-powered ship Engineering Simulation SYstem (NESSY) has been developed by the Japan Atomic Energy Research Institute as an advanced design tool for research and development of future marine reactors. A marine reactor must respond to changing loads and to the ship`s motions because of the ship`s maneuvering and its presence in a marine environment. The NESSY has combined programs for the reactor plant behavior calculations and the ship`s motion calculations. Thus, it can simulate reactor power fluctuations caused by changing loads and the ship`s motions. It can also simulate the behavior of water in the pressurizer and steam generators. This water sloshes in response to the ship`s motions. The performance of NESSY has been verified by comparing the simulation calculations with the measured data obtained by experiments performed using the nuclear ship Mutsu. The effects of changing loads and the ship`s motions on the reactor behavior can be accurately simulated by NESSY.

Kusunoki, Tsuyoshi; Kyouya, Masahiko; Kobayashi, Hideo; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Office of Nuclear Ship Research and Development

1995-02-01T23:59:59.000Z

423

Reactor & Nuclear Systems Publications | ORNL  

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

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

424

E-Print Network 3.0 - argonaut bilbao reactor Sample Search Results  

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

reprocessing and transmutation... with industry, national laboratories, and universities; upgrading and sharing of research reactors Source: Laughlin, Robert B. - Department of...

425

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

E-Print Network [OSTI]

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

Jiltchenkov, Dmitri Victorovich

2012-01-01T23:59:59.000Z

426

E-Print Network 3.0 - advanced reactor study Sample Search Results  

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

Hotel bargains... to iron out their differences over the site of the International Thermonuclear Experimental Reactor Source: Fusiongnition Research Experiment (FIRE) Collection:...

427

Radiochemical characteristics of tritium to be considered in fusion reactor facility design  

Science Journals Connector (OSTI)

The results of research and development related to radiochemical characteristics of tritium to be considered in a fusion reactor facility design are summarized. Reactions induced by...

S. Ohira; T. Hayashi; W. Shu; T. Yamanishi

2007-06-01T23:59:59.000Z

428

Improved Prediction of the Temperature Feedback in TRISO-Fueled Reactors  

SciTech Connect (OSTI)

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.

Javier Ortensi; Abderrafi M. Ougouag

2009-08-01T23:59:59.000Z

429

Method of producing exfoliated graphite composite compositions for fuel cell flow field plates  

DOE Patents [OSTI]

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.

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

2014-04-08T23:59:59.000Z

430

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

E-Print Network [OSTI]

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

Tsvetkova, Galina Valeryevna

2004-09-30T23:59:59.000Z

431

Strengthening the nuclear-reactor fuel cycle against proliferation  

SciTech Connect (OSTI)

Argonne National Laboratory (ANL) conducts several research programs that serve to reduce the risks of fissile-material diversion from the nuclear-reactor fuel cycle. The objectives are to provide economical and efficient neutron or power generation with the minimum of inherent risks, and to further minimize risks by utilizing sophisticated techniques to detect attempts at material diversion. This paper will discuss the Reduced Enrichment Research and Test Reactor (RERTR) Program, the Isotope Correlation Technique (ICT), and Proliferation-Resistant Closed-Cycle Reactors. The first two are sponsored by the DOE Office of Arms Control and Nonproliferation.

Travelli, A.; Snelgrove, J.; Persiani, P. [Argonne National Lab., IL (United States). Arms Control and Nonproliferation Program

1992-12-31T23:59:59.000Z

432

Nuclear reactor control column  

DOE Patents [OSTI]

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

Bachovchin, Dennis M. (Plum Borough, PA)

1982-01-01T23:59:59.000Z

433

Laboratory Directed Research and Development (LDRD)  

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

concepts - including advanced reactor modeling, nuclear waste reduction and fuel recycling - to develop DOE-NE's needs. INL's LDRD research stimulates exploration at the...

434

01-02253B_OR_Knox_map.ai  

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

OAK RIDGE INN & SUITES THE RIDGE INN and THE SUPER 8 MOTEL ORNL GRAPHITE REACTOR NATIONAL TRANSPORTATION RESEARCH CENTER ETTP OBSERVATION CENTER MELTON HILL DAM HAMPTON INN BULL...

435

Environmental/Radiological Assistance Directory (ERAD) | Department...  

Office of Environmental Management (EM)

Brookhaven Graphite Research Reactor (BGRR), D&D Authorized Limits for Portsmith Oil Inventory Clearance of Real and Personal Property RESRAD Family of Codes Knowledge...

436

E-Print Network 3.0 - alternatives study volume Sample Search...  

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

of Engineering, University of Cambridge Collection: Engineering 53 Brookhaven Graphite Research Reactor Decommissioning Project Summary: of this Feasibility Study (FS) is...

437

Nuclear reactor reflector  

DOE Patents [OSTI]

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

1994-01-01T23:59:59.000Z

438

Spherical torus fusion reactor  

DOE Patents [OSTI]

The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

Martin Peng, Y.K.M.

1985-10-03T23:59:59.000Z

439

Graphite-ceramic rf Faraday-thermal shield and plasma limiter  

DOE Patents [OSTI]

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.

Hwang, D.L.Q.; Hosea, J.C.

1983-05-05T23:59:59.000Z

440

The role of hydrogen in room-temperature ferromagnetism at graphite surfaces  

SciTech Connect (OSTI)

We present a x-ray dichroism study of graphite surfaces that addresses the origin and magnitude of ferromagnetism in metal-free carbon. We find that, in addition to carbon {pi} states, also hydrogen-mediated electronic states exhibit a net spin polarization with significant magnetic remanence at room temperature. The observed magnetism is restricted to the top {approx}10 nm of the irradiated sample where the actual magnetization reaches {approx_equal} 15 emu/g at room temperature. We prove that the ferromagnetism found in metal-free untreated graphite is intrinsic and has a similar origin as the one found in proton bombarded graphite.

Ohldag, Hendrik

2011-08-12T23:59:59.000Z

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441

Effect of oxidizing environment on the strength and oxidation kinetics of HTGR graphites. Part I. Reactivity and strength loss of H451, PGX and IG-11 graphites  

SciTech Connect (OSTI)

The effects of oxidizing atmosphere and temperature on the reactivities and strengths of PGX, H451, and IG-11 were examined. Preliminary measurements of the oxidation kinetics of these graphites in H/sub 2/O-, CO/sub 2/- and O/sub 2/-containing atmospheres indicated that the reactivities of H451 graphite toward O/sub 2/ and H/sub 2/O are quite similar to those of IG-11 graphite. The apparent activation energy for oxidation of these in O/sub 2/ were estimated to be approx. 175 kJ/mol while that in H/sub 2/O is probably approx. 200 kJ/mol. The apparent activation energy of IG-11 graphite oxidized in CO/sub 2/ is 255 +- 18 kJ/mol. PGX graphite was found to be quite variable in its reactivity toward H/sub 2/O. A linear dependence with (Fe) was determined, but other intrinsic properties were found to affect its absolute reactivity by as much as a factor of X50.

Eto, M.; Growcock, F.B.

1981-09-01T23:59:59.000Z

442

Physics characteristics of a large, passive, pressure tube light water reactor with voided calandria  

SciTech Connect (OSTI)

A light water cooled and moderated pressure tube reactor concept has been developed that can survive loss-of-coolant accidents (LOCAs) without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tube. The reactor employs a solid SiC-coated graphite fuel matrix in the pressure tubes and a calandria tank containing a low-pressure gas, surrounded by a graphite reflector. This normally voided calandria is connected to a light water heat sink. The cover gas displaces light water from the calandria during normal operation, while during LOCAs it allows passive calandria flooding. It is shown that such a system, with high void fraction in the core region, exhibits a high degree of neutron thermalization and a large prompt neutron lifetime, similar to D{sub 2}O moderated cores, although light water is used as both coolant and moderator. Moreover, the extremely large neutron migration length results in a strongly coupled core with a flat thermal flux profile and inherent stability against xenon spatial oscillations. The heterogeneous arrangement of the fuel and moderator ensures a negative void coefficient under all circumstances. Flooding of the calandria space with light water results in redundant reactor shutdown. Use of particle fuel allows attainment of high burnups.

Hejzlar, P.; Driscoll, M.J.; Todreas, N.E. [Massachusetts Inst. of Technology, Cambridge, MA (United States). Dept. of Nuclear Engineering

1995-11-01T23:59:59.000Z

443

Progress in xenon stability analysis for a flux-flattened reactor  

SciTech Connect (OSTI)

A power oscillation induced by a xenon transient is common to all large thermal reactors. The Hanford N Reactor, operated by UNC Nuclear Industries for the US Dept. of Energy, is a large graphite-moderated horizontal pressure tube reactor whose dimensions are approx. 10 x 10 x 12 m. To preclude xenon instability, the N Reactor was designed to have a large negative power coefficient of reactivity. Previous analyses and observations made over 23 yr of operation have confirmed that the reactor is in fact very stable. Currently an effort is under way to introduce axial flux-flattening to improve the operating and long-term safety margins. Safety evaluations associated with the flux-flattening program require a complete review of the xenon stability question. To achieve the level of accuracy necessary to make an unambiguous analysis of the xenon stability characteristics in the flux-flattened mode, it is necessary to employ sophisticated methods. To this end, it was decided to write a three-dimensional nodal code and to couple xenon and temperature feedbacks to this code. This paper summarizes the progress made in developing such a code for xenon stability analysis related to the N Reactor.

Wu, R.M.; Lan, J.S.; Albrecht, R.W.; Toffer, H.; Omberg, R.P.

1987-01-01T23:59:59.000Z

444

Reactor Thermal-Hydraulics  

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

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

445

Reactor hot spot analysis  

SciTech Connect (OSTI)

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

Vilim, R.B.

1985-08-01T23:59:59.000Z

446

Gas Reactor Technology R&D  

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

U.S. Department of Energy to Invest U.S. Department of Energy to Invest up to $7.3 Million for "Deep-Burn" Gas-Reactor Technology R&D Artist's rendering of Nuclear Plant An artist's rendering of the Next Generation Nuclear Plant concept. The U.S. Department of Energy today announced a Funding Opportunity Announcement (FOA) valued at $7.3 million for universities, commercial entities, National Laboratories with expertise in the concept of nuclear fuel "Deep-Burn" in which plutonium and higher transuranics recycled from spent nuclear fuel are destroyed. The funding opportunity seeks to establish the technological foundations that will support the role of the very-high-temperature, gas-cooled reactor (VHTR) in the nuclear fuel cycle -- which is one of the prototype reactors being researched/developed under

447

Molten metal reactors  

DOE Patents [OSTI]

A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

2013-11-05T23:59:59.000Z

448

F Reactor Inspection  

ScienceCinema (OSTI)

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-11-24T23:59:59.000Z

449

F Reactor Inspection  

SciTech Connect (OSTI)

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29T23:59:59.000Z

450

Vitrification of surrogate mixed wastes in a graphite electrode arc melter  

SciTech Connect (OSTI)

Demonstration tests for vitrifying mixed wastes and contaminated soils have been conducted using a small (800 kVA), industrial-scale, three-phase AC, graphite electrode furnace located at the Albany Research Center of the United States Bureau of Mines (USBM). The feed mixtures were non-radioactive surrogates of various types of mixed (radioactive and hazardous), transuranic-contaminated wastes stored and buried at the Idaho National Engineering Laboratory (INEL). The feed mixtures were processed with added soil from the INEL. Objectives being evaluated include (1) equipment capability to achieve desired process conditions and vitrification products for different feed compositions, (2) slag and metals tapping capability, (3) partitioning of transuranic elements and toxic metals among the furnace products, (4) slag, fume, and metal products characteristics, and (5) performance of the feed, furnace and air pollution control systems. The tests were successfully completed in mid-April 1995. A very comprehensive process monitoring, sampling and analysis program was included in the test program. Sample analysis, data reduction, and results evaluation are currently underway. Initial results indicate that the furnace readily processed around 20,000 lb of widely ranging feed mixtures at feedrates of up to 1,100 lb/hr. Continuous feeding and slag tapping was achieved. Molten metal was also tapped twice during the test program. Offgas emissions were efficiently controlled as expected by a modified air pollution control system.

Soelberg, N.R.; Chambers, A.G.; Ball, L. [and others

1995-11-01T23:59:59.000Z

451

Low-cost and durable catalyst support for fuel cells: graphite submicronparticles  

SciTech Connect (OSTI)

Low-cost graphite submicronparticles (GSP) are employed as a possible catalyst support for polymer electrolyte membrane (PEM) fuel cells. Platinum nanoparticles are deposited on Vulcan XC-72 carbon black (XC-72), carbon nanotubes (CNT), and GSP via ethylene glycol (EG) reduction method. The morphologies and the crystallinity of Pt/XC-72, Pt/CNT, and Pt/GSP are characterized with X-ray diffraction and transmission electron microscope, which shows that Pt nanoparticles (~ 3.5 nm) are uniformly dispersed on GSP support. Pt/GSP exhibits the highest activity towards oxygen reduction reactions. The durability study indicates that Pt/GSP is 2 ~ 3 times durable than Pt/CNT and Pt/XC-72. The enhanced durability of Pt/GSP catalyst is attributed to the higher corrosion resistance of graphite submicronparticles, which results from higher graphitization degree of GSP support. Considering its low production cost, graphite submicronparticles are promising electrocatalyst support for fuel cells.

Zhang, Sheng; Shao, Yuyan; Li, Xiaohong; Nie, Zimin; Wang, Yong; Liu, Jun; Yin, Geping; Lin, Yuehe

2010-01-01T23:59:59.000Z

452

The addition of a calender machine to a pyrolytic graphite sheet production plant  

E-Print Network [OSTI]

This thesis documents the process and challenges of adding a new calender machine to AvCarb Material Solutions' pyrolytic graphite production plant. Before the machine could be used for mass production, several experiments ...

Svenson, Ernest Knute

2014-01-01T23:59:59.000Z

453

Enhancing thermal conductivity of fluids with graphite nanoparticles and carbon nanotube  

DOE Patents [OSTI]

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.

Zhang, Zhiqiang (Lexington, KY); Lockwood, Frances E. (Georgetown, KY)

2008-03-25T23:59:59.000Z

454

Induced crystallization of single-chain polyethylene on a graphite surface: Molecular dynamics simulation  

Science Journals Connector (OSTI)

Molecular dynamics (MD) simulations have been carried out on the crystallization of single-chain polyethylene (PE) which was adsorbed on a graphite (001) surface on one side and exposed to vacuum on the other at different temperatures. The MD simulation data have been analyzed to provide information about the crystallization process of polymer adsorbed on the solid substrate. The isothermal crystallization of PE proceeds in two steps: