Science.gov

Sample records for graphite reactor physics

  1. Graphite for nuclear reactors

    SciTech Connect

    Virgiliev, Yu.S.; Kalyagina, I.P.

    1993-12-31

    Relative dimensional changes and physical properties of structural graphites - {Gamma}p-280 (nuclear graphite) and {Gamma}p{Pi}-2 (modificated variety of nuclear graphite for the rings of elastic contact) irradiated at temperatures ranging from 320 to 1900K with a fluence of about 2.5.10{sup 22}nvt (E {ge} 0.18 MeV) are represented. In order to ensure a long-time serviceability of the VGM - reactor blocks the high-strength graphite of {Gamma}p-1 grade are developed. The properties and its irradiation changes of {Gamma}p-1 graphite are represented. A secondary swelling of the graphite develops similar to the swelling of metals, alloys and high-melting compounds.

  2. CALANDRIA TYPE SODIUM GRAPHITE REACTOR

    DOEpatents

    Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

    1964-02-11

    A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

  3. METHOD OF FABRICATING A GRAPHITE MODERATED REACTOR

    DOEpatents

    Kratz, H.R.

    1963-05-01

    S>A nuclear reactor formed of spaced bodies of uranium and graphite blocks is improved by diffusing helium through the graphite blocks in order to replace the air in the pores of the graphite with helium. The helium-impregnated graphite conducts heat better, and absorbs neutrons less, than the original air- impregnated graphite. (AEC)

  4. SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

    DOEpatents

    Dickinson, R.W.

    1963-03-01

    This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

  5. Optimized Conditioning of Activated Reactor Graphite

    SciTech Connect

    Tress, G.; Doehring, L.; Pauli, H.; Beer, H.-F.

    2002-02-25

    The research reactor DIORIT at the Paul Scherrer Institute was decommissioned in 1993 and is now being dismantled. One of the materials to be conditioned is activated reactor graphite, approximately 45 tons. A cost effective conditioning method has been developed. The graphite is crushed to less than 6 mm and added to concrete and grout. This graphite concrete is used as matrix for embedding dismantling waste in containers. The waste containers that would have been needed for separate conditioning and disposal of activated reactor graphite are thus saved. Applying the new method, the cost can be reduced from about 55 SFr/kg to about 17 SFr/kg graphite.

  6. US graphite reactor D&D experience

    SciTech Connect

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

    1997-02-01

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

  7. WORKER STACKS GRAPHITE BLOCKS AGAINST INNER SOUTH WALL OF REACTOR. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    WORKER STACKS GRAPHITE BLOCKS AGAINST INNER SOUTH WALL OF REACTOR. INL NEGATIVE NO. 3925. Unknown Photographer, 12/14/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  8. JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

    DOEpatents

    Szilard, L.; Wigner, E.P.; Creutz, E.C.

    1959-05-12

    Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

  9. Criticality Studies of Graphite Moderated Production Reactors

    DTIC Science & Technology

    1980-01-01

    o Tn II 1.4’, 1 I 1 1 eIc (ei pwr I rea -II 1Ii67 I , . I II II I I++ + Fi If67 It + : i ’’I if I if I If II l ~ l I.. . . I -- . . . I... tritium ). Moderator grade graphites also are designed for minimum distortion and gas evolution under irradiation. These physical characteristics are

  10. Examination of reactor grade graphite using neutron powder diffraction

    NASA Astrophysics Data System (ADS)

    DiJulio, D. D.; Hawari, A. I.

    2009-07-01

    Graphite is of principal interest in Generation IV nuclear reactor concepts. In particular, graphite will be the moderator for the Very High Temperature Reactor. In support of experimental and computational investigations that aim at understanding the behavior of reactor grade graphite under operating conditions, neutron powder diffraction experiments have been performed at the North Carolina State University PULSTAR reactor. The collected diffraction patterns exhibit intense broadening of several of the reflections, characteristic of turbostratic stacking. In order to quantify this disorder structurally, a model combined with a Rietveld-like refinement approach was implemented, which includes several refinable parameters that aim at describing this type of structure. Stacking parameters representing the probabilities of a random and registered shift between stacking packages were defined. The results indicate that the studied reactor grade graphite specimens contain a small fraction of layer disorder. The inferred interlayer spacing for the specimens is slightly larger than the theoretical value for graphite of 0.335 nm and the lattice constant is slightly less than 0.246 nm. The developed methodology is found to be successful in fitting the neutron diffraction patterns of reactor grade graphite.

  11. Dismantling of the DIORIT research reactor - Conditioning of activated graphite.

    PubMed

    Sierra Perler, Isabel Cecilia; Beer, Hans-Frieder; Müth, Joachim; Kramer, Andreas

    2017-08-16

    The research reactor DIORIT at the Paul Scherrer Institute was a natural uranium reactor moderated by D2O. It was put in operation in 1960 and finally shut down in August 1977. The dismantling project started in 1982 and could be successfully finished on September 11th, 2012. About 40 tons of activated reactor graphite had to be conditioned during the dismantling of this research reactor. The problem of conditioning of activated reactor graphite had not been solved so far worldwide. Therefore a conditioning method considering radiation protection and economic aspects had to be developed. As a result, the graphite was crushed to a particle size smaller than 5 mm and added as sand substitute to a specially developed grout. The produced graphite concrete was used as a matrix for embedding dismantling waste in containers. By conditioning the graphite conventionally, about 58.5 m(3) (13 containers) of waste volume would have been generated. The new PSI invention resulted in no additional waste caused by graphite. Consequently, the resulting waste volume, as well as the costs, were substantially reduced. Copyright © 2017. Published by Elsevier Ltd.

  12. Bulk-bronzied graphites for plasma-facing components in ITER (International Thermonuclear Experimental Reactor)

    SciTech Connect

    Hirooka, Y.; Conn, R.W.; Doerner, R.; Khandagle, M. . Inst. of Plasma and Fusion Research); Causey, R.; Wilson, K. ); Croessmann, D.; Whitley, J. ); Holland, D.; Smolik, G. ); Matsuda, T.; Sogabe, T. (Toyo Tanso Co. Ltd., O

    1990-06-01

    Newly developed bulk-boronized graphites and boronized C-C composites with a total boron concentration ranging from 1 wt % to 30 wt % have been evaluated as plasma-facing component materials for the International Thermonuclear Experimental Reactor (ITER). Bulk-boronized graphites have been bombarded with high-flux deuterium plasmas at temperatures between 200 and 1600{degree}C. Plasma interaction induced erosion of bulk-boronized graphites is observed to be a factor of 2--3 smaller than that of pyrolytic graphite, in regimes of physical sputtering, chemical sputtering and radiation enhanced sublimation. Postbombardment thermal desorption spectroscopy indicates that bulk-boronized graphites enhance recombinative desorption of deuterium, which leads to a suppression of the formation of deuterocarbon due to chemical sputtering. The tritium inventory in graphite has been found to decrease by an order of magnitude due to 10 wt % bulk-boronization at temperatures above 1000{degree}C. The critical heat flux to induce cracking for bulk-boronized graphites has been found to be essentially the same as that for non-boronized graphites. Also, 10 wt % bulk-boronization of graphite hinders air oxidation nearly completely at 800{degree}C and reduces the steam oxidation rate by a factor of 2--3 at around 1100 and 1350{degree}C. 38 refs., 5 figs.

  13. TSX graphite for extended use in the N-Reactor

    SciTech Connect

    Kennedy, C.R.

    1985-08-01

    This report reviews the limited amount of irradiation data available for grade TSX graphite with the purpose of obtaining reasonable estimates of material behavior. The results are enhanced by obtaining generalized behavior characteristics demonstrated by similar grades of graphite, such as CSF, AGOT, and PGA. Intent of this work is to furnish the necessary coefficients to describe the material behavior for inclusion in the constitutive equations for the anisotropic graphite grade TSX. Estimates of the free-dimensional changes of TSX graphite as a function of temperature and fluence have been made and shown to be in good agreement with the data. The effects of irradiation on other physical properties, such as elastic moduli, conductivity, and coefficient of thermal expansion, are also described. The irradiation creep characteristics of TSX graphite are also estimated on the basis of data for similar grades of graphite in the US and Europe. Crude approximations of stresses generated in the keyed structure were made to demonstrate the magnitude of the problem. The results clearly predict that the filler-block keys will fail and the tube-block keys will not. It is also indicated that the overall stack height growth will be increased by 25 to 38 mm (1 to 1.5 in.) because of creep.

  14. Alloying of steel and graphite by hydrogen in nuclear reactor

    NASA Astrophysics Data System (ADS)

    Krasikov, E.

    2017-02-01

    In traditional power engineering hydrogen may be one of the first primary source of equipment damage. This problem has high actuality for both nuclear and thermonuclear power engineering. Study of radiation-hydrogen embrittlement of the steel raises the question concerning the unknown source of hydrogen in reactors. Later unexpectedly high hydrogen concentrations were detected in irradiated graphite. It is necessary to look for this source of hydrogen especially because hydrogen flakes were detected in reactor vessels of Belgian NPPs. As a possible initial hypothesis about the enigmatical source of hydrogen one can propose protons generation during beta-decay of free neutrons поскольку inasmuch as protons detected by researches at nuclear reactors as witness of beta-decay of free neutrons.

  15. MODELING THE ELECTROLYTIC DECHLORINATION OF TRICHLOROETHYLENE IN A GRANULAR GRAPHITE-PACKED REACTOR

    EPA Science Inventory

    A comprehensive reactor model was developed for the electrolytic dechlorination of trichloroethylene (TCE) at a granular-graphite cathode. The reactor model describes the dynamic processes of TCE dechlorination and adsorption, and the formation and dechlorination of all the major...

  16. MODELING THE ELECTROLYTIC DECHLORINATION OF TRICHLOROETHYLENE IN A GRANULAR GRAPHITE-PACKED REACTOR

    EPA Science Inventory

    A comprehensive reactor model was developed for the electrolytic dechlorination of trichloroethylene (TCE) at a granular-graphite cathode. The reactor model describes the dynamic processes of TCE dechlorination and adsorption, and the formation and dechlorination of all the major...

  17. Thermo-mechanical stresses in the lumps of laying of thermal neutron pulsed graphite reactor

    SciTech Connect

    Boyko, V.I.; Guralev, S.S.; Koshelev, F.P.

    1993-12-31

    The research thermal neutron pulsed graphite reactor (PGR) is intended to get powerful neutron and gamma radiation streams. The reactor is the homogeneous carbon-uranium reactor with a graphite reflector. The reactor laying consists of a number of columns and it`s sizes are 2400* 2400*4500mm. The shape of the active zone is almost cubic, it`s sizes are 1400*1400*1330mm. There is a vertical experimental channel in the reactor laying for irradiation of test samples. The operation of the reactor is briefly described. Evaluations about the workability of the reactor laying lumps in the neutron flashout regime are made.

  18. Physical aging in graphite epoxy composites

    NASA Technical Reports Server (NTRS)

    Kong, E. S. W.

    1981-01-01

    The matrix dominated mechanical behavior of a graphite epoxy composite was found to be affected by sub Tg annealing. Postcured + or - 45 deg 4S specimens of Thornel 300 graphite/Narmco 5208 epoxy were quenched from above Tg and given a sub Tg annealing at 140 C for times up to 10 to the 5th power min. The ultimate tensile strength, strain to break, and toughness of the composite material were found to decrease as functions of sub Tg annealing time. No weight loss was observed during the sub Tg annealing. The time dependent change in mechanical behavior is explained on the basis of free volume changes that are related to the physical aging of the nonequilibrium glassy network epoxy. The results imply possible changes in composite properties with service time.

  19. Implications of Graphite Radiation Damage on the Neutronic, Operational, and Safety Aspects of Very High Temperature Reactors

    SciTech Connect

    Hawari, Ayman I

    2011-08-30

    In both the prismatic and pebble bed designs of Very High Temperature Reactors (VHTR), the graphite moderator is expected to reach exposure levels of 1021 to 1022 n/cm2 over the lifetime of the reactor. This exposure results in damage to the graphite structure. In this work, molecular dynamic and ab initio molecular static calculations will be used to: 1) simulate radiation damage in graphite under various irradiation and temperature conditions, 2) generate the thermal neutron scattering cross sections for damaged graphite, and 3) examine the resulting microstructure to identify damage formations that may produce the high-temperature Wigner effect. The impact of damage on the neutronic, operational and safety behavior of the reactor will be assessed using reactor physics calculations. In addition, tests will be performed on irradiated graphite samples to search for the high-temperature Wigner effect, and phonon density of states measurements will be conducted to quantify the effect on thermal neutron scattering cross sections using these samples.

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

    SciTech Connect

    Strizak, Joe P; Burchell, Timothy D; Windes, Will

    2011-12-01

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

  1. Initial Comparison of Baseline Physical and Mechanical Properties for the VHTR Candidate Graphite Grades

    SciTech Connect

    Carroll, Mark C.

    2014-09-01

    High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is establishing accurate as-manufactured mechanical and physical property distributions in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered “candidate” grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding

  2. Reactor antineutrinos and nuclear physics

    NASA Astrophysics Data System (ADS)

    Balantekin, A. B.

    2016-11-01

    Short-baseline reactor neutrino experiments successfully measured the neutrino parameters they set out to measure, but they also identified a shape distortion in the 5-7 MeV range as well as a reduction from the predicted value of the flux. Nuclear physics input into the calculations of reactor antineutrino spectra needs to be better refined if this anomaly is to be interpreted as due to sterile neutrino states.

  3. Physical aging in graphite/epoxy composites

    NASA Technical Reports Server (NTRS)

    Kong, E. S. W.

    1983-01-01

    Sub-Tg annealing has been found to affect the properties of graphite/epoxy composites. The network epoxy studied was based on the chemistry of tetraglycidyl 4,4'-diamino-diphenyl methane (TGDDM) crosslinked by 4,4'-diamino-diphenyl sulfone (DDS). Differential scanning calorimetry, thermal mechanical analysis, and solid-state cross-polarized magic-angle-spinning nuclear magnetic resonance spectroscopy have been utilized in order to characterize this process of recovery towards thermodynamic equilibrium. The volume and enthalpy recovery as well as the 'thermoreversibility' aspects of the physical aging are discussed. This nonequilibrium and time-dependent behavior of network epoxies are considered in view of the increasingly wide applications of TGDDM-DDS epoxies as matrix materials of structural composites in the aerospace industry.

  4. Physical aging in graphite/epoxy composites

    NASA Technical Reports Server (NTRS)

    Kong, E. S. W.

    1983-01-01

    Sub-Tg annealing has been found to affect the properties of graphite/epoxy composites. The network epoxy studied was based on the chemistry of tetraglycidyl 4,4'-diamino-diphenyl methane (TGDDM) crosslinked by 4,4'-diamino-diphenyl sulfone (DDS). Differential scanning calorimetry, thermal mechanical analysis, and solid-state cross-polarized magic-angle-spinning nuclear magnetic resonance spectroscopy have been utilized in order to characterize this process of recovery towards thermodynamic equilibrium. The volume and enthalpy recovery as well as the 'thermoreversibility' aspects of the physical aging are discussed. This nonequilibrium and time-dependent behavior of network epoxies are considered in view of the increasingly wide applications of TGDDM-DDS epoxies as matrix materials of structural composites in the aerospace industry.

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

    SciTech Connect

    Mcwilliams, A. J.

    2015-09-08

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

  6. Light water reactor health physics.

    PubMed

    Prince, Robert J; Bradley, Scott E

    2004-11-01

    In this article an overview of the historical development of light water reactor health physics programs is presented. Operational health physics programs have developed and matured as experience in operating and maintaining light water reactors has been gained. Initial programs grew quickly in both size and complexity with the number and size of nuclear units under construction and in operation. Operational health physics programs evolved to face various challenges confronted by the nuclear industry, increasing the effectiveness of radiological safety measures. Industry improvements in radiological safety performance have resulted in significant decreases in annual collective exposures from a high value of 790 person-rem in 1980 to 117 person-rem per reactor in 2002. Though significant gains have been made, the continued viability of the nuclear power industry is confronted with an aging workforce, as well as the challenges posed by deregulation and the need to maintain operational excellence.

  7. Light water reactor health physics.

    PubMed

    Prince, Robert J; Bradley, Scott E

    2005-06-01

    In this article an overview of the historical development of light water reactor health physics programs is presented. Operational health physics programs have developed and matured as experience in operating and maintaining light water reactors has been gained. Initial programs grew quickly in both size and complexity with the number and size of nuclear units under construction and in operation. Operational health physics programs evolved to face various challenges confronted by the nuclear industry, increasing the effectiveness of radiological safety measures. Industry improvements in radiological safety performance have resulted in significant decreases in annual collective exposures from a high value of 790 person-rem in 1980 to 117 person-rem per reactor in 2002. Though significant gains have been made, the continued viability of the nuclear power industry is confronted with an aging workforce, as well as the challenges posed by deregulation and the need to maintain operational excellence.

  8. EFFECTS OF REACTOR CONDITIONS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODE.

    EPA Science Inventory

    Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas pr...

  9. EFFECTS OF REACTOR CONDITIONS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODE

    EPA Science Inventory

    Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas pr...

  10. EFFECTS OF REACTOR CONDITIONS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODE.

    EPA Science Inventory

    Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas pr...

  11. EFFECTS OF REACTOR CONDITIONS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODE

    EPA Science Inventory

    Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas pr...

  12. REACTOR HAVING NaK-UO$sub 2$ SLURRY HELICALLY POSITIONED IN A GRAPHITE MODERATOR

    DOEpatents

    Rodin, M.B.; Carter, J.C.

    1962-05-15

    A reactor utilizing 20% enriched uranium consists of a central graphite island in cylindrical form, with a spiral coil of tubing fitting against the central island. An external graphite moderator is placed around the central island and coil. A slurry of uranium dioxide dispersed in alkali metal passes through the coil to transfer heat externally to the reactor. There are also conventional controls for regulating the nuclear reaction. (AEC)

  13. Graphite Technology Development Plan

    SciTech Connect

    W. Windes; T. Burchell; M.Carroll

    2010-10-01

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

  14. A safety assessment of the use of graphite in nuclear reactors licensed by the US NRC

    SciTech Connect

    Schweitzer, D.G.; Gurinsky, D.H.; Kaplan, E.; Sastre, C.

    1987-09-01

    This report reviews existing literature and knowledge on graphite burning and on stored energy accumulation and releases in order to assess what role, if any, a stored energy release can have in initiating or contributing to hypothetical graphite burning scenarios in research reactors. It also addresses the question of graphite ignition and self-sustained combustion in the event of a loss-of-coolant accident (LOCA). The conditions necessary to initiate and maintain graphite burning are summarized and discussed. From analyses of existing information it is concluded that only stored energy accumulations and releases below the burning temperature (650/sup 0/C) are pertinent. After reviewing the existing knowledge on stored energy it is possible to show that stored energy releases do not occur spontaneously, and that the maximum stored energy that can be released from any reactor containing graphite is a very small fraction of the energy produced during the first few minutes of a burning incident. The conclusions from these analyses are that the potential to initiate or maintain a graphite burning incident is essentially independent of the stored energy in the graphite, and depends on other factors that are unique for these reactors, research reactors, and for Fort St. Vrain. In order to have self-sustained rapid graphite oxidation in any of these reactors, certain necessary conditions of geometry, temperature, oxygen supply, reaction product removal, and a favorable heat balance must be maintained. There is no new evidence associated with either the Windscale Accident or the Chernobyl Accident that indicates a credible potential for a graphite burning accident in any of the reactors considered in this review.

  15. Development of graphite/metals bondings for fusion reactor applications

    NASA Astrophysics Data System (ADS)

    Brossa, F.; Franconi, E.; Schiller, P.

    1992-09-01

    The need to join graphite and metal components is particularly important in the nuclear industry, especially in the first wall components of fusion machines. The aim of this work was to find brazing agents which can wet steel and graphite without producing excessive alloying, steel dissolution or the formation of fragile components and to determine the best thermal brazing cycles. Direct steel-graphite junctions are very sensitive to thermal cycling. Multiple brazings with three components (AISI 316L-Cu-graphite or AISI 316L-Mo-graphite) and four components (AISI 316L-Cu-W-graphite) were thus made. Layers of W on the graphite were produced using chemical vapour deposition (CVD), while the Mo and some brazing agents were deposited using vacuum plasma spray (VPS). It was found that multiple brazings were clearly more resistant to thermal shocks than simple junctions.

  16. Treatment of irradiated graphite from French Bugey reactor

    SciTech Connect

    Stevens, Howard; Laurent, Gerard

    2013-07-01

    In 2008, following the general French plan for nuclear waste management, Electricite de France attempted to find for irradiated graphite an alternative solution to direct storage at the low-activity long-life storage center in France managed by the national agency for wastes (ANDRA). EDF management requested that its engineering arm, EDF CIDEN, study the graphite treatment alternatives to direct storage. In mid-2008, this study revealed the potential advantage for EDF to use a steam reforming process known as Thermal Organic Reduction, 'THOR' (owned by Studsvik, Inc., USA), to treat or destroy the graphite matrix and limit the quantity of secondary waste to be stored. In late 2009, EDF began a test program with Studsvik to determine if the THOR steam reforming process could be used to destroy the graphite. The program also sought to determine if the graphite could be treated to release the bulk of activity while minimizing the gasification of the bulk mass of the graphite. In October 2009, tests with non-irradiated graphite were completed and demonstrated destruction of a graphite matrix by the THOR process at satisfactory rates. After gasifying the graphite, focus shifted to the effect of roasting graphite at high temperatures in inert gases with low concentrations of oxidizing gases to preferentially remove volatile radionuclides while minimizing the graphite mass loss to 5%. A radioactive graphite sleeve was imported from France to the US for these tests. Completed in April 2010, 'Phase I' of testing showed that the process removed >99% of H-3 and 46% of C-14 with <6% mass loss. Completed in September 2011, 'Phase II' testing achieved increased removals as high as 80% C-14. During Phase II, it was also discovered that roasting in a reducing atmosphere helped to limit the oxidation of the graphite. Future work seeks to explore the effects of reducing gases to limit the bulk oxidation of graphite. If the graphite could be decontaminated of long-lived radionuclides

  17. Seismic study of high-temperature engineering test reactor core graphite structures

    SciTech Connect

    Iyoku, T.; Inagaki, Y.; Shiozawa, S. . Oarai Research Establishment); Nishiguchi, I. )

    1992-08-01

    This paper discusses the High-Temperature Engineering Test Reactor (HTTR) a 30-MW (thermal) helium gas-cooled reactor with a core composed of prismatic graphite blocks piled on core support structures. Safety analyses have been made for the seismic design of the HTTR core using a two-dimensional seismic analysis code called SONATINA-2V, which was developed by the Japan Atomic Energy Research Institute. To evaluate the validity of the SONATINA-2V code and confirm the structural integrity of the core graphite blocks, large-scale seismic tests are conducted using a half-scale vertical section model and a full-scale seven-column model of the core graphite blocks and the core support structures. The test results are in good agreement with the analytical ones, and the validity of the analysis code is confirmed. The structural integrity of the core graphite blocks is confirmed by both analytical and test results.

  18. Neutronic reactor

    DOEpatents

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  19. Requirements for Reactor Physics Design

    SciTech Connect

    Diamond,D.J.

    2008-04-11

    It has been recognized that there is a need for requirements and guidance for design and operation of nuclear power plants. This is becoming more important as more reactors are being proposed to be built. In parallel with activities in individual countries are norms established by international organizations. This paper discusses requirements/guidance for neutronic design and operation as promulgated by the U.S. Nuclear Regulatory Commission (NRC). As an example, details are given for one reactor physics parameter, namely, the moderator temperature reactivity coefficient. The requirements/guidance from the NRC are discussed in the context of those generated for the International Atomic Energy Agency. The requirements/guidance are not identical from the two sources although they are compatible.

  20. Design of Modern Reactors for Synthesis of Thermally Expanded Graphite

    NASA Astrophysics Data System (ADS)

    Strativnov, Eugene V.

    2015-05-01

    One of the most progressive trends in the development of modern science and technology is the creation of energy-efficient technologies for the synthesis of nanomaterials. Nanolayered graphite (thermally exfoliated graphite) is one of the key important nanomaterials of carbon origin. Due to its unique properties (chemical and thermal stability, ability to form without a binder, elasticity, etc.), it can be used as an effective absorber of organic substances and a material for seal manufacturing for such important industries as gas transportation and automobile. Thermally expanded graphite is a promising material for the hydrogen and nuclear energy industries. The development of thermally expanded graphite production is resisted by high specific energy consumption during its manufacturing and by some technological difficulties. Therefore, the creation of energy-efficient technology for its production is very promising.

  1. Design of Modern Reactors for Synthesis of Thermally Expanded Graphite.

    PubMed

    Strativnov, Eugene V

    2015-12-01

    One of the most progressive trends in the development of modern science and technology is the creation of energy-efficient technologies for the synthesis of nanomaterials. Nanolayered graphite (thermally exfoliated graphite) is one of the key important nanomaterials of carbon origin. Due to its unique properties (chemical and thermal stability, ability to form without a binder, elasticity, etc.), it can be used as an effective absorber of organic substances and a material for seal manufacturing for such important industries as gas transportation and automobile. Thermally expanded graphite is a promising material for the hydrogen and nuclear energy industries. The development of thermally expanded graphite production is resisted by high specific energy consumption during its manufacturing and by some technological difficulties. Therefore, the creation of energy-efficient technology for its production is very promising.

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

    SciTech Connect

    Brown, Thomas; Poncet, Bernard

    2013-07-01

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

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

    SciTech Connect

    Blaine Grover

    2010-10-01

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

  4. Activation analysis of concrete and graphite in the experimental reactor RUS.

    PubMed

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

    2005-01-01

    The decommissioning and dismantling of nuclear installations after their service life involves the necessary disassembling, handling and disposing of a large amount of radioactive equipment and structures. In particular, the concrete that has been used as a biological reactor shield and graphite that has been used as a moderator-reflector represent the majority of waste, requiring geological disposal. To reduce this undesirable volume to the minimum and to successfully plan the dismantling and disposal of radioactive materials to storage facilities, the activations of the structures should be accurately evaluated. In the framework of the decommissioning and the dismantling of the experimental reactor of the University of Strasbourg, detailed activation estimates have been conducted to characterise the graphite and the structural materials present in the reactor environment. For this purpose, the chemical compositions of fresh graphite samples and different types of concrete have been determined by activation analysis in the research reactors OSIRIS and ORPHEE of CEA Saclay (France). Then, the activations of graphite, concrete and other materials have been calculated in the whole reactor, as a function of the three main nuclear data libraries, i.e. ENDF, JEF and JENDL. In parallel, the activations of representative graphite and concrete samples have been measured experimentally. The comparison of theoretical predictions with experimental values validates the approach and the methodology used in the present study and tests the consistency and the reliability of the nuclear data used for activation analysis. We believe that a similar approach could also be used for the decommissioning of industrial nuclear reactors.

  5. Probing Unparticle Physics in Reactor Neutrinos

    SciTech Connect

    Bolanos, A.

    2008-11-13

    Unparticle physics is studied by using reactor neutrino data. We obtain limits to the scalar unparticle couplings depending on different values for the parameter d. We found that, as has been already noticed, reactor neutrino data is a good tool to put constraints on unparticle physics. Thanks to a detailed analysis of the experimental characteristics of reactor data we find better constraints than the previously reported.

  6. Deformation and fracture of irradiated polygranular pile grade A reactor core graphite

    NASA Astrophysics Data System (ADS)

    Heard, P. J.; Wootton, M. R.; Moskovic, R.; Flewitt, P. E. J.

    2011-11-01

    Pile grade A (PGA) graphite is used as a moderator in UK gas cooled nuclear reactors. This is a polygranular, aggregate material with quasi-brittle behaviour. When exposed to the service environment the material is subject to radiolytic oxidation that results in mass loss and an attendant increase in porosity. In the present work both unirradiated and irradiated small specimens of PGA graphite have been subjected to diametral compression. A novel trench-probe loading method is also described that allows micro-scale specimens prepared by focused ion beam milling to be fractured in a focused ion beam work station. This allows the fracture characteristics of selected regions of the graphite microstructure to be interrogated. The load-displacement and fracture characteristics of both the unirradiated and irradiated PGA graphite are compared and shown to be consistent with quasi-brittle behaviour. In addition, surface features consistent with elastically induced twins are observed associated with filler particles of the graphite. The results are discussed with respect to the quasi-brittle behaviour of this polygranular graphite.

  7. Graphite irradiation testing for HTR technology at the High Flux Reactor in Petten

    NASA Astrophysics Data System (ADS)

    Vreeling, J. A.; Wouters, O.; Laan, J. G. van der

    2008-10-01

    In 2001 the Nuclear Research and Consultancy Group started a large graphite irradiation program for the development of High Temperature Reactor technology in the European framework. The irradiation experiments, containing present day available graphite grades, are performed at the High Flux Reactor in Petten. The grades are NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, LPEB, IG-110 and IG-430. In the fifth framework programme (2001-2004) and sixth framework programme (2005-2009) four irradiation experiments are foreseen, resulting in design curves at irradiation temperatures between 650 °C and 950 °C. The post-irradiation testing is focused on dimensional changes, dynamic Young's modulus, coefficient of thermal expansion and coefficient of thermal conductivity. The irradiation programme and preliminary results from the first irradiation experiment at 750 °C to 8 dpa will be discussed in this paper.

  8. Analysis of Wigner energy release process in graphite stack of shut-down uranium-graphite reactor

    NASA Astrophysics Data System (ADS)

    Bespala, E. V.; Pavliuk, A. O.; Kotlyarevskiy, S. G.

    2015-10-01

    Data, which finding during thermal differential analysis of sampled irradiated graphite are presented. Results of computational modeling of Winger energy release process from irradiated graphite staking are demonstrated. It's shown, that spontaneous combustion of graphite possible only in adiabatic case.

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

    SciTech Connect

    Poncet, Bernard

    2013-07-01

    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)

  10. Thermal, Structural, and Radiological Properties of Irradiated Graphite from the ASTRA Research Reactor - Implications for Disposal

    SciTech Connect

    Lexa, D.; Kropf, A.J.

    2006-07-01

    There is currently no consensus regarding the disposal of nuclear graphite. The two main problems include high activities of C-14 and H-3 as well as accumulation of Wigner energy (responsible for the Windscale Pile 1 fire in 1957). The release of Wigner energy from the graphite of the inner thermal column of the ASTRA research reactor has been studied by differential scanning calorimetry and simultaneous differential scanning calorimetry / synchrotron powder x-ray diffraction between 25 deg. C and 725 deg. C at a heating rate of 10 deg. C.min{sup -1}. The graphite, having been subject to a fast-neutron fluence from {approx}10{sup 17}' to {approx}10{sup 20} n.cm{sup -2} over the life time of the reactor at temperatures not exceeding 100 deg. C, exhibits Wigner energies ranging from 25 to 572 J.g{sup -1} and a Wigner energy accumulation rate of {approx}7 x 10{sup -17} J.g{sup -1}/n.cm{sup -2}. The shape of the rate-of-heat-release curves, e.g., maximum at ca. 200 deg. C and a fine structure at higher temperatures, varies with sample position within the inner thermal column, i.e., the distance from the reactor core. Crystal structure of samples closest to the reactor core (fast-neutron fluence > 1.5 - 5.0 x 10{sup 19} n.cm{sup -2}) is destroyed while that of samples farther from the reactor core (fast-neutron fluence < 1.5 - 5.0 x 10{sup 19} n.cm{sup -2}) is intact. The dependence of the c lattice parameter on temperature between 25 deg. C and 200 deg. C as determined by Rietveld refinement for the non-amorphous samples leads to the expected microscopic thermal expansion coefficient along the c axis of {approx}26 x 10{sup -6} deg. C{sup -1}. However, at 200 deg. C, coinciding with the maximum in the rate-of-heat-release curves, the rate of thermal expansion abruptly decreases indicating a crystal lattice relaxation. The C-14 activity in the inner thermal column graphite ranges from 6 to 467 kBq.g{sup -1}. Prior to interim storage or final disposal, thermal treatment

  11. Irradiation experiments on high temperature gas-cooled reactor fuels and graphites at the high flux reactor petten

    NASA Astrophysics Data System (ADS)

    Ahlf, J.; Conrad, R.; Cundy, M.; Scheurer, H.

    1990-04-01

    Because of its favourable design and operational characteristics and the availability of dedicated experimental equipment the High Flux Reactor at Petten has been extensively used as a test bed for HTR fuel and graphite irradiations for more than 20 years. Earlier fuel testing programmes contributed to the development of the coated fuel particle concept by extended screening tests. Now these programmes concentrate on performance testing of reference coated fuel particles and reference fuel elements for the German HTR-Module, the HTR-500 and to a lesser extent for the US HTGR concepts. It is shown with representative examples that these fuels have excellent fission product retention capabilities under normal and anticipated off-normal operating conditions. Extended irradiation programmes in the HFR Petten have significantly contributed to the database for the design of HTR graphite structures. The programmes not only comprise radiation damage accumulation in the temperature range from 570 to 1570 K up to very high fast neutron fluences and its influence on technological properties, but also irradiations under specified load conditions to investigate the irradiation creep behaviour of various graphites in the temperature range 570 to 1170 K.

  12. New Concept of a Small Passive-Safety Reactor with UO{sub 2}-Graphite-Water Core

    SciTech Connect

    Tetsuo Matsumura; Takanori Kameyama; Yasushi Nauchi; Izumi Kinoshita

    2002-07-01

    New concept of a passive-safety reactor with I/O:-graphite-water core is proposed, which has negligible possibility of core melting and, flexibility of total reactor power. Present concept has simple plant system design without a reactor pressure vessel, ECCS, recirculation systems (of BWR) and others. Therefore construction cost per electric power generation is expected to be slightly low comparing with conventional large scale WRs. (authors)

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

    SciTech Connect

    S. Blaine Grover

    2014-05-01

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

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

    SciTech Connect

    Romano, T.

    1997-09-29

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

  15. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO2-cooled reactors and for the decontamination of irradiated graphite waste

    NASA Astrophysics Data System (ADS)

    Le Guillou, M.; Toulhoat, N.; Pipon, Y.; Moncoffre, N.; Khodja, H.

    2015-06-01

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO2-cooled nuclear fission reactors (called UNGG for "Uranium Naturel-Graphite-Gaz") to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D+ ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO2) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the total amount produced

  16. Physics: A New Reactor Physics Analysis Toolkit

    SciTech Connect

    C. Rabiti; Y. Wang; G. Palmiotti; H. Hiruta; J. Cogliati; A. Alfonsi

    2011-06-01

    In the last year INL has internally pursued the development of a new reactor analysis tool: PHISICS. The software is built in a modular approach to simplify the independent development of modules by different teams and future maintenance. Most of the modules at the time of this summary are still under development (time dependent transport driver, depletion, cross section I/O and interpolation, generalized perturbation theory), while the transport solver INSTANT (Intelligent Nodal and Semi-structured Treatment for Advanced Neutron Transport) has already been widely used1, 2, 3, 4. For this reason we will focus mainly on the presentation of the transport solver INSTANT

  17. Reactor Physics Characterization of the HTR Module with UCO Fuel

    SciTech Connect

    Gerhard Strydom

    2011-01-01

    ABSTRACT The HTR Module [1] is a graphite-moderated, helium cooled pebble bed High Temperature Reactor (HTR) design that has been extensively used as a reference template for the former South African and current Chinese HTR [2] programs. This design utilized spherical fuel elements packed into a dynamic pebble bed, consisting of TRISO coated uranium oxide (UO2) fuel kernels with a U-235 enrichment of 7.8% and a Heavy Metal loading of 7 grams per pebble. The main objective of this study is to compare several important reactor physics and core design parameters for the HTR Module and an identical design utilizing UCO fuel kernels. Fuel kernels of this type are currently being tested in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) as part of the larger Next Generation Nuclear Plant (NGNP) project. The PEBBED-THERMIX [3] code, which was developed specifically for the analysis of pebble bed HTRs, was used to compare the coupled neutronic and thermal fluid performance of the two designs.

  18. Examination of Surface Deposits on Oldbury Reactor Core Graphite to Determine the Concentration and Distribution of 14C.

    PubMed

    Payne, Liam; Heard, Peter J; Scott, Thomas B

    2016-01-01

    Pile Grade A graphite was used as a moderator and reflector material in the first generation of UK Magnox nuclear power reactors. As all of these reactors are now shut down there is a need to examine the concentration and distribution of long lived radioisotopes, such as 14C, to aid in understanding their behaviour in a geological disposal facility. A selection of irradiated graphite samples from Oldbury reactor one were examined where it was observed that Raman spectroscopy can distinguish between underlying graphite and a surface deposit found on exposed channel wall surfaces. The concentration of 14C in this deposit was examined by sequentially oxidising the graphite samples in air at low temperatures (450°C and 600°C) to remove the deposit and then the underlying graphite. The gases produced were captured in a series of bubbler solutions that were analysed using liquid scintillation counting. It was observed that the surface deposit was relatively enriched with 14C, with samples originating lower in the reactor exhibiting a higher concentration of 14C. Oxidation at 600°C showed that the remaining graphite material consisted of two fractions of 14C, a surface associated fraction and a graphite lattice associated fraction. The results presented correlate well with previous studies on irradiated graphite that suggest there are up to three fractions of 14C; a readily releasable fraction (corresponding to that removed by oxidation at 450°C in this study), a slowly releasable fraction (removed early at 600°C in this study), and an unreleasable fraction (removed later at 600°C in this study).

  19. Examination of Surface Deposits on Oldbury Reactor Core Graphite to Determine the Concentration and Distribution of 14C

    PubMed Central

    Payne, Liam; Heard, Peter J.; Scott, Thomas B.

    2016-01-01

    Pile Grade A graphite was used as a moderator and reflector material in the first generation of UK Magnox nuclear power reactors. As all of these reactors are now shut down there is a need to examine the concentration and distribution of long lived radioisotopes, such as 14C, to aid in understanding their behaviour in a geological disposal facility. A selection of irradiated graphite samples from Oldbury reactor one were examined where it was observed that Raman spectroscopy can distinguish between underlying graphite and a surface deposit found on exposed channel wall surfaces. The concentration of 14C in this deposit was examined by sequentially oxidising the graphite samples in air at low temperatures (450°C and 600°C) to remove the deposit and then the underlying graphite. The gases produced were captured in a series of bubbler solutions that were analysed using liquid scintillation counting. It was observed that the surface deposit was relatively enriched with 14C, with samples originating lower in the reactor exhibiting a higher concentration of 14C. Oxidation at 600°C showed that the remaining graphite material consisted of two fractions of 14C, a surface associated fraction and a graphite lattice associated fraction. The results presented correlate well with previous studies on irradiated graphite that suggest there are up to three fractions of 14C; a readily releasable fraction (corresponding to that removed by oxidation at 450°C in this study), a slowly releasable fraction (removed early at 600°C in this study), and an unreleasable fraction (removed later at 600°C in this study). PMID:27706228

  20. Radiation Effects in Graphite

    SciTech Connect

    Burchell, Timothy D

    2012-01-01

    The requirements for a solid moderator are reviewed and the reasons that graphite has become the solid moderator of choice discussed. The manufacture and properties of some currently available near-isotropic and isotropic grades are described. The major features of a graphite moderated reactors are briefly outlined. Displacement damage and the induced structural and dimensional changes in graphite are described. Recent characterization work on nano-carbons and oriented pyrolytic graphites that have shed new light on graphite defect structures are reviewed, and the effect of irradiation temperature on the defect structures is highlighted. Changes in the physical properties of nuclear graphite caused by neutron irradiation are reported. Finally, the importance of irradiation induced creep is presented, along with current models and their deficiencies.

  1. Gas-phase impregnation of porous media with pyrocarbon as a promising trend in the manufacturing technology of carbon-graphite materials and products for reactor engineering

    SciTech Connect

    Gurin, V.A.; Zelensky, V.F.; Konotop, Yu.F.

    1993-12-31

    This paper presents some special features of producing carbon-graphite materials by gas-phase methods. Main differences between these methods and the traditional one of graphite fabrication are discussed; basic characteristics of the equipment available at the KIPT are given. The HTGR type reactors need radiation resistant grades of graphite with a normal operation guaranteed for a specified time.

  2. Graphite Technology Development Plan

    SciTech Connect

    W. Windes; T. Burchell; R. Bratton

    2007-09-01

    This technology development plan is designed to provide a clear understanding of the research and development direction necessary for the qualification of nuclear grade graphite for use within the Next Generation Nuclear Plant (NGNP) reactor. The NGNP will be a helium gas cooled Very High Temperature Reactor (VHTR) with a large graphite core. Graphite physically contains the fuel and comprises the majority of the core volume. Considerable effort will be required to ensure that the graphite performance is not compromised during operation. Based upon the perceived requirements the major data needs are outlined and justified from the perspective of reactor design, reatcor performance, or the reactor safety case. The path forward for technology development can then be easily determined for each data need. How the data will be obtained and the inter-relationships between the experimental and modeling activities will define the technology development for graphite R&D. Finally, the variables affecting this R&D program are discussed from a general perspective. Factors that can significantly affect the R&D program such as funding, schedules, available resources, multiple reactor designs, and graphite acquisition are analyzed.

  3. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2009-05-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any

  4. Rapid analysis of 14C and 3H in graphite and concrete for decommissioning of nuclear reactor.

    PubMed

    Hou, Xiaolin

    2005-06-01

    A rapid oxidizing combustion method using a commercial Sample Oxidizer has been investigated to determine separately the 14C and 3H activities in graphite and concrete. By this method the sample preparation time can be reduced to 2-3 min. The detection limits for 3H and 14C are 0.96 and 0.58 Bq/g graphite and 0.11 and 0.06 Bq/g concrete, respectively. The cross contamination of 14C and tritium in the preparation of samples is less than 0.2%. The interference of other radionuclides in the determination of 14C and tritium in graphite is insignificant. The analytical accuracy, investigated by the standard addition method, is better than 95%. In addition, an acid digestion method has also been used to examine the graphite and concrete activities, to allow comparison with the method developed herein. The two methods show good agreement for graphite samples. Graphite samples were collected from the Danish Reactors DR-2 and DR-3, in addition to two concrete cores drilled in the Danish reactor DR-2; these were analysed for 3H and 14C using the method that has been developed.

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

    SciTech Connect

    Blaine Grover

    2012-10-01

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

  6. ICP-MS analysis of fission product diffusion in graphite for High-Temperature Gas-Cooled Reactors

    NASA Astrophysics Data System (ADS)

    Carter, Lukas M.

    Release of radioactive fission products from nuclear fuel during normal reactor operation or in accident scenarios is a fundamental safety concern. Of paramount importance are the understanding and elucidation of mechanisms of chemical interaction, nuclear interaction, and transport phenomena involving fission products. Worldwide efforts to reduce fossil fuel dependence coupled with an increasing overall energy demand have generated renewed enthusiasm toward nuclear power technologies, and as such, these mechanisms continue to be the subjects of vigorous research. High-Temperature Gas-Cooled Reactors (HTGRs or VHTRs) remain one of the most promising candidates for the next generation of nuclear power reactors. An extant knowledge gap specific to HTGR technology derives from an incomplete understanding of fission product transport in major core materials under HTGR operational conditions. Our specific interest in the current work is diffusion in reactor graphite. Development of methods for analysis of diffusion of multiple fission products is key to providing accurate models for fission product release from HTGR core components and the reactor as a whole. In the present work, a specialized diffusion cell has been developed and constructed to facilitate real-time diffusion measurements via ICP-MS. The cell utilizes a helium gas-jet system which transports diffusing fission products to the mass spectrometer using carbon nanoparticles. The setup was designed to replicate conditions present in a functioning HTGR, and can be configured for real-time release or permeation measurements of single or multiple fission products from graphite or other core materials. In the present work, we have analyzed release rates of cesium in graphite grades IG-110, NBG-18, and a commercial grade of graphite, as well as release of iodine in IG-110. Additionally we have investigated infusion of graphite samples with Cs, I, Sr, Ag, and other surrogate fission products for use in release or

  7. Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

    SciTech Connect

    Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

    2015-03-01

    In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.

  8. Leaching of /sup 14/C and /sup 36/Cl from Hanford reactor graphite

    SciTech Connect

    Gray, W.J.; Morgan, W.C.

    1988-12-01

    The leach rates of /sup 14C/ and /sup 36Cl/ were measured on solid cylindrical samples of graphite prepared from a bar retrieved from one of the surplus Hanford production reactors. Static leach tests were conducted in deionized water and Hanford ground water at temperatures of 20/degree/C to 90/degree/C for 8 weeks. The graphite samples were completely submerged in the leachant, and the entire volume of leachant was changed and analyzed weekly. The leach rates of both /sup 14C/ and /sup 36Cl/ decreased with time and appeared to approach steady-state values that were independent of temperature in the case of /sup 36Cl/ but decreased with temperature in the case of /sup 14C/. Both radionuclides leached more slowly in Hanford ground water. The data are compared with previously measured and estimated leach rates. Implications of the data regarding possible rate-limiting mechanisms are also discussed. 4 refs., 18 figs., 4 tabs.

  9. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    NASA Astrophysics Data System (ADS)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-10-01

    The study evaluates the possible use of graphite foam as the bonding material between U-Pu-Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U-15Pu-6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600-660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

  10. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  11. Effect of oxidation on physical properties of nuclear grade graphites

    NASA Astrophysics Data System (ADS)

    Matsuo, Hideto; Fujii, Kimio; Imai, Hisashi; Kurosawa, Takeshi

    1985-11-01

    Changes in thermal conductivity, coefficient of thermal expansion (CTE). electrical resistivity, and Young's modulus were measured for the two nuclear grade graphites IG-11 and H451 oxidized thermally in air at 723 K and by water vapor at 1123 K. Thermal conductivity, CTE and Young's modulus decreased and electrical resistivity increased owing to the oxidation. The changes were dependent on the graphites as well as the oxidation condition. The results were analyzed and discussed. Furthermore the changes in thermal shock resistance were discussed for the thermally oxidized nuclear grade graphites.

  12. Reactor Physics Methods and Analysis Capabilities in SCALE

    SciTech Connect

    DeHart, Mark D; Bowman, Stephen M

    2011-01-01

    The TRITON sequence of the SCALE code system provides a powerful, robust, and rigorous approach for performing reactor physics analysis. This paper presents a detailed description of TRITON in terms of its key components used in reactor calculations. The ability to accurately predict the nuclide composition of depleted reactor fuel is important in a wide variety of applications. These applications include, but are not limited to, the design, licensing, and operation of commercial/research reactors and spent-fuel transport/storage systems. New complex design projects such as next-generation power reactors and space reactors require new high-fidelity physics methods, such as those available in SCALE/TRITON, that accurately represent the physics associated with both evolutionary and revolutionary reactor concepts as they depart from traditional and well-understood light water reactor designs.

  13. Reactor Physics Methods and Analysis Capabilities in SCALE

    SciTech Connect

    Mark D. DeHart; Stephen M. Bowman

    2011-05-01

    The TRITON sequence of the SCALE code system provides a powerful, robust, and rigorous approach for performing reactor physics analysis. This paper presents a detailed description of TRITON in terms of its key components used in reactor calculations. The ability to accurately predict the nuclide composition of depleted reactor fuel is important in a wide variety of applications. These applications include, but are not limited to, the design, licensing, and operation of commercial/research reactors and spent-fuel transport/storage systems. New complex design projects such as next-generation power reactors and space reactors require new high-fidelity physics methods, such as those available in SCALE/TRITON, that accurately represent the physics associated with both evolutionary and revolutionary reactor concepts as they depart from traditional and well-understood light water reactor designs.

  14. Health physics research reactor reference dosimetry

    SciTech Connect

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

    1987-06-01

    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.

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

    SciTech Connect

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

    2002-02-26

    The Brookhaven Graphite Research Reactor (BGRR) Historical Site Assessment (BNL 1999) identified contamination inside the Below Grade Ducts (BGD) resulting from the deposition of fission and activation products from the pile on the inner carbon steel liner during reactor operations. Due to partial flooding of the BGD since shutdown, some of this contamination may have leaked out of the ducts into the surrounding soils. The baseline remediation plan for cleanup of contaminated soils beneath the BGD involves complete removal of the ducts, followed by surveying the underlying and surrounding soils, then removing soil that has been contaminated above cleanup goals. Alternatively, if soil contamination around and beneath the BGD is either non-existent/minimal (below cleanup goals) or is very localized and can be ''surgically removed'' at a reasonable cost, the BGD can be decontaminated and left in place. The focus of this Department of Energy Accelerated Site Technology Deployment (DOE ASTD) project was to determine the extent (location, type, and level) of soil contamination surrounding the BGD and to present this data to the stakeholders as part of the Engineering Evaluation/Cost Analysis (EE/CA) process. A suite of innovative characterization tools was used to complete the characterization of the soil surrounding the BGD in a cost-effective and timely fashion and in a manner acceptable to the stakeholders. The tools consisted of a tracer gas leak detection system that was used to define the gaseous leak paths out of the BGD and guide soil characterization studies, a small-footprint Geoprobe to reach areas surrounding the BGD that were difficult to access, two novel, field-deployed, radiological analysis systems (ISOCS and BetaScint) and a three-dimensional (3D) visualization system to facilitate data analysis/interpretation. All of the technologies performed as well or better than expected and the characterization could not have been completed in the same time or at

  16. Thermal-hydraulic analysis of N Reactor graphite and shield cooling system performance

    SciTech Connect

    Low, J.O.; Schmitt, B.E.

    1988-02-01

    A series of bounding (worst-case) calculations were performed using a detailed hydrodynamic RELAP5 model of the N Reactor graphite and shield cooling system (GSCS). These calculations were specifically aimed to answer issues raised by the Westinghouse Independent Safety Review (WISR) committee. These questions address the operability of the GSCS during a worst-case degraded-core accident that requires the GDCS to mitigate the consequences of the accident. An accident scenario previously developed was designed as the hydrogen-mitigation design-basis accident (HMDBA). Previous HMDBA heat transfer analysis,, using the TRUMP-BD code, was used to define the thermal boundary conditions that the GSDS may be exposed to. These TRUMP/HMDBA analysis results were used to define the bounding operating conditions of the GSCS during the course of an HMDBA transient. Nominal and degraded GSCS scenarios were investigated using RELAP5 within or at the bounds of the HMDBA transient. 10 refs., 42 figs., 10 tabs.

  17. Advances in reactor physics education: Visualization of reactor parameters

    SciTech Connect

    Snoj, L.; Kromar, M.; Zerovnik, G.

    2012-07-01

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

  18. Finite Element Based Stress Analysis of Graphite Component in High Temperature Gas Cooled Reactor Core Using Linear and Nonlinear Irradiation Creep Models

    SciTech Connect

    Mohanty, Subhasish; Majumdar, Saurindranath

    2015-01-01

    Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

  19. Surface chemistry and physics of deuterium retention in lithiated graphite

    SciTech Connect

    Taylor, C. N.; Krstic, Predrag S; Allain, J. P.; Heim, B.; Skinner, C. H.; Kugel, H.

    2011-01-01

    Lithium wall conditioning in TFTR, CDX-U, T-11M, TJ-II and NSTX is found to yield enhanced plasma performance manifest, in part, through improved deuterium particle control. X-ray photoelectron spectroscopy (XPS) experiments examine the affect of D irradiation on lithiated graphite and show that the surface chemistry of lithiated graphite after D ion bombardment (500 eV/amu) is fundamentally different from that of non-Li conditioned graphite. Instead of simple LiD bonding seen in pure liquid Li, graphite introduces additional complexities. XPS spectra show that Li-O-D (533.0 {+-} 0.6 eV) and Li-C-D (291.4 {+-} 0.6 eV) bonds, for a nominal Li dose of 2 {micro}m, become 'saturated' with D at fluences between 3.8 and 5.2 x 10{sup 17} cm{sup -2}. Atomistic modeling indicate that Li-O-D-C interactions may be a result of multibody effects as opposed to molecular bonding.

  20. OVERVIEW OF NUCLEAR PHYSICS LABORATORY (IMMEDIATELY EAST OF SPSE REACTOR ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    OVERVIEW OF NUCLEAR PHYSICS LABORATORY (IMMEDIATELY EAST OF SP-SE REACTOR ROOM), LEVEL -15’, LOOKING SOUTHWEST. NOTE SLIDING STEEL PLATE DOOR BETWEEN LABORATORY AND REACTOR ROOM - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

  1. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  2. Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

    NASA Astrophysics Data System (ADS)

    Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

    2017-07-01

    Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10-6 K-1 (α∥) and 6.15 × 10-6 K-1 (α⊥) at the temperature range of 25-700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

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

    SciTech Connect

    Vaghetto, Rodolfo; Capone, Luigi; Hassan, Yassin A

    2011-05-31

    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.

  4. Advanced reactor physics methods for heterogeneous reactor cores

    NASA Astrophysics Data System (ADS)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  5. REACTOR PHYSICS MODELING OF SPENT NUCLEAR RESEARCH REACTOR FUEL FOR SNM ATTRIBUTION AND NUCLEAR FORENSICS

    SciTech Connect

    Sternat, M.; Beals, D.; Webb, R.; Nichols, T.

    2010-06-09

    Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle. The

  6. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    SciTech Connect

    Not Available

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

  7. Physics methods development for the NCSU PULSTAR reactor

    SciTech Connect

    Perez, P.B.; Mayo, C.W.; Giavedoni, E.

    1996-12-31

    The safety analysis reports (SARs) of all university research reactors include analyses that determine reactor physics parameters. The initial SAR analyses utilized numerical models, codes, cross-section libraries, and computing platforms available at the time. Advances and updates in all of these contributing areas make it difficult or impractical to resort to the earlier methodologies for meeting current analysis needs. Many facilities updated their physics methods during the high-enrichment uranium (HEU) to low-enrichment uranium (LEU) conversion effort. These facilities updated their SAR with current methodologies. The North Carolina State University`s (NCSU`s) PULSTAR research reactor was designed to use low-enrichment (4%) fuel, and as a result, the facility did not update the reactor physics analyses during the HEU-to-LEU conversion program. An effort is currently under way at NCSU to develop new and updated methods for reactor physics calculations. Currently planned physics calculations for the PULSTAR reactor support both reactor licensing and experimental facility development goals. These goals include the following: 1. Increase excess reactivity by introducing beryllium reflector assemblies and a mixed-enrichment core. 2. Characterize various experimental facilities in support of neutron transmutation doping, prompt gamma analysis, and neutron depth profiling. 3. Establish core loading patterns that optimize characteristics for experimental facilities. Two and three-dimensional, multigroup models utilizing the DANT-SYS and MCNP codes have been developed in support of these goals. Results and lessons learned with the DANT-SYS code are presented in this paper.

  8. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    NASA Technical Reports Server (NTRS)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  9. NGNP Graphite Selection and Acquisition Strategy

    SciTech Connect

    Burchell, T.; Bratton, R.; Windes, W.

    2007-09-30

    The nuclear graphite (H-451) previously used in the United States for High-Temperature Reactors (HTRs) is no longer available. New graphites have been developed and are considered suitable candidates for the Next-Generation Nuclear Plant (NGNP). A complete properties database for these new, available, candidate grades of graphite must be developed to support the design and licensing of NGNP core components. Data are required for the physical, mechanical (including radiation-induced creep), and oxidation properties of graphites. Moreover, the data must be statistically sound and take account of in-billet, between billets, and lot-to-lot variations of properties. These data are needed to support the ongoing development1 of the risk-derived American Society of Mechanical Engineers (ASME) graphite design code (a consensus code being prepared under the jurisdiction of the ASME by gas-cooled reactor and NGNP stakeholders including the vendors). The earlier Fort St. Vrain design of High-Temperature Reactor (HTRs) used deterministic performance models for H-451, while the NGNP will use new graphite grades and risk-derived (probabilistic) performance models and design codes, such as that being developed by the ASME. A radiation effects database must be developed for the currently available graphite materials, and this requires a substantial graphite irradiation program. The graphite Technology Development Plan (TDP)2 describes the data needed and the experiments planned to acquire these data in a timely fashion to support NGNP design, construction, and licensing. The strategy for the selection of appropriate grades of graphite for the NGNP is discussed here. The final selection of graphite grades depends upon the chosen reactor type and vendor because the reactor type (pebble bed or prismatic block) has a major influence on the graphite chosen by the designer. However, the time required to obtain the needed irradiation data for the selected NGNP graphite is sufficiently

  10. Effects of prehydrostatic loading on mechanical and physical properties of nuclear-grade graphites

    NASA Astrophysics Data System (ADS)

    Yoda, S.; Eto, M.; Ioka, I.

    1984-06-01

    The effects of prehydrostatic loading on mechanical and physical properties were examined for two isostatistically-molded graphites, grades IG-11 and Iso-20. The increase in loading levels caused the monotonic decrease in specimen volume of both graphites, the change of which was much larger in the case of IG-11 graphite than of Iso-20. The electrical resistivity increased continuously with an increase in prehydrostatic loading levels, where the effect was more pronounced for IG-11 than for Iso-20 graphite. The manner in which the Young's moduli of the two materials changed after prehydrostatic loading was clearly different from each other, i.e., for IG-11 it decreased continuously with increasing pressure, whereas it increased at small hydrostatic pressures and leveled off above 300 MPa for Iso-20. A clear difference was also observed for the flexural strength of both materials: Its continuous decrease was caused by prehydrostatic loading for IG-11 graphite, whereas for Iso-20 it remained at the original value up to 500 MPa and decreased abruptly above that pressure level. An explanation of these features was provided from the viewpoint that microcracks are induced by hydrostatic loading.

  11. Computational prediction of dust production in graphite moderated pebble bed reactors

    NASA Astrophysics Data System (ADS)

    Rostamian, Maziar

    The scope of the work reported here, which is the computational study of graphite wear behavior, supports the Nuclear Engineering University Programs project "Experimental Study and Computational Simulations of Key Pebble Bed Thermomechanics Issues for Design and Safety" funded by the US Department of Energy. In this work, modeling and simulating the contact mechanics, as anticipated in a PBR configuration, is carried out for the purpose of assessing the amount of dust generated during a full power operation year of a PBR. A methodology that encompasses finite element analysis (FEA) and micromechanics of wear is developed to address the issue of dust production and its quantification. Particularly, the phenomenon of wear and change of its rate with sliding length is the main focus of this dissertation. This work studies the wear properties of graphite by simulating pebble motion and interactions of a specific type of nuclear grade graphite, IG-11. This study consists of two perspectives: macroscale stress analysis and microscale analysis of wear mechanisms. The first is a set of FEA simulations considering pebble-pebble frictional contact. In these simulations, the mass of generated graphite particulates due to frictional contact is calculated by incorporating FEA results into Archard's equation, which is a linear correlation between wear mass and wear length. However, the experimental data by Johnson, University of Idaho, revealed that the wear rate of graphite decreases with sliding length. This is because the surfaces of the graphite pebbles become smoother over time, which results in a gradual decrease in wear rate. In order to address the change in wear rate, a more detailed analysis of wear mechanisms at room temperature is presented. In this microscale study, the wear behavior of graphite at the asperity level is studied by simulating the contact between asperities of facing surfaces. By introducing the effect of asperity removal on wear rate, a nonlinear

  12. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    SciTech Connect

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical

  13. Computational mathematics and physics of fusion reactors

    PubMed Central

    Garabedian, Paul R.

    2003-01-01

    Theory has contributed significantly to recent advances in magnetic fusion research. New configurations have been found for a stellarator experiment by computational methods. Solutions of a free-boundary problem are applied to study the performance of the plasma and look for islands in the magnetic surfaces. Mathematical analysis and numerical calculations have been used to study equilibrium, stability, and transport of optimized fusion reactors. PMID:14614129

  14. Modelleing the Multiaxial Strength of Nuclear Graphite

    SciTech Connect

    Burchell, Timothy D; Yahr, Terry; Battiste, Rick

    2006-01-01

    The core of a prismatic High Temperature Reactor (HTR) is constructed from an array of nuclear graphite components including replaceable fuel blocks, replaceable and permanent moderator blocks, and core support posts. Similarly, the core of a Pebble Bed Reactor is confined by large graphite blocks which define the (annular) core geometry. In both HTR designs (prismatic and pebble bed) the large graphite components act as neutron moderator and reflector as well as providing mechanical support to the active core. During reactor operation the graphite components of the core are subjected to complex stress states arising from structural loads, thermal gradients, neutron irradiation damage, and seismic events. Structural design of HTR cores requires that the designer have a suitable theory of failure. Both deterministic (e.g. maximum principal stress theory) and probabilistic (e.g., Weibull failure theory) have been considered as candidates. To test candidate theories a multiaxial testing program was conducted at Oak Ridge National Laboratory on grade H-451 graphite, the fuel element and moderator graphite used in the Fort St. Vrain high temperature reactor in the USA. Large test specimens ({approx}27 cm length) were subjected to combined axial stress (both tension and compression) and internal pressure. A total of 59 specimens were tested at 9 stress ratios in the first and fourth stress quadrants. In a parallel effort a physically based fracture model of graphite was developed. The model used a fracture mechanics based failure criteria and has been shown to predict the tensile failure probability of several graphites of widely ranging texture. Here we report the basis and performance of the fracture model and multiaxial strength data for grade H-451 graphite. Moreover, we report the successful extension of the model to predict the failure envelope for H-451 graphite in the first and fourth multiaxial quadrants. The model's predictions are compared to experimental

  15. Actinide nuclear data for reactor physics calculations

    SciTech Connect

    Brady, M.C.; Wright, R.Q. ); England, T.R. )

    1991-07-01

    Calculational methodologies and data sources used to predict and recommend fission-product yields and delayed neutron and prompt neutron data for a number of actinide nuclides are presented and discussed. This compilation of nuclear data is the result of a nearly three-year effort under the Japan/US Actinide Program (JUSAP) at Oak Ridge National Laboratory to provide nuclear data supporting the preliminary design of an actinide burner reactor. In this type of reactor, minor actinides are the major components of the fuel. Nuclear data for these minor actinides are, therefore, essential in the design of such reactors. Fission yield, delayed neutron, and prompt neutron data are presented in the report for the following nuclides: Neptumium-237, Plutonium-238, -240, and -242, Americium-241 and -243, and Curium-242, -243, -244, -246, and -248. Additionally, prompt neutron data are also presented for these nuclides (except Plutonium-240, -242 and Curium-242) and for Curium-245 and -247. As in all compilations of nuclear data, the information in this report is subject to change as newer data become available. Most of the data presented here are based on calculational methodologies and should be revised as experimental data become available. The release of Version 6 of the Evaluated Nuclear Data Files (ENDF/B-6) is expected to be completed in 1991 and should replace this evaluation in areas of overlap although no serious discrepancies are expected between this compilation and ENDF/B-6. Because of the large amount of data comprising this compilation and limitations in publishing such a voluminous report, a complete listing of the explicit data is not included in this report. The data are, however, available from the authors on 5 {1/2}-in. high-density (1.2-Mbyte) diskettes. The file contents and formats are described in the text, and examples are given in the appendices. 34 refs., 18 tabs.

  16. Neutrino Physics with Accelerator Driven Subcritical Reactors

    NASA Astrophysics Data System (ADS)

    Ciuffoli, Emilio

    2017-09-01

    Accelerator Driven Subcritical System (ADS) reactors are being developed around the world, to produce energy and, at the same time, to provide an efficient way to dispose of and to recycle nuclear waste. Used nuclear fuel, by itself, cannot sustain a chain reaction; however in ADS reactors the additional neutrons which are required will be supplied by a high-intensity accelerator. This accelerator will produce, as a by-product, a large quantity of {\\bar{ν }}μ via muon Decay At Rest (µDAR). Using liquid scintillators, it will be possible to to measure the CP-violating phase δCP and to look for experimental signs of the presence of sterile neutrinos in the appearance channel, testing the LSND and MiniBooNE anomalies. Even in the first stage of the project, when the beam energy will be lower, it will be possible to produce {\\bar{ν }}e via Isotope Decay At Rest (IsoDAR), which can be used to provide competitive bounds on sterile neutrinos in the disappearance channel. I will consider several experimental setups in which the antineutrinos are created using accelerators that will be constructed as part of the China-ADS program.

  17. Reactor physics verification of the MCNP6 unstructured mesh capability

    SciTech Connect

    Burke, T. P.; Kiedrowski, B. C.; Martz, R. L.; Martin, W. R.

    2013-07-01

    The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)

  18. Critical Configuration and Physics Mesaurements for Graphite Reflected Assemblies of U(93.15)O2 Fuel Rods (1.27-CM Pitch)

    SciTech Connect

    Margaret A. Marshall

    2011-09-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory's Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950's efforts were made to study 'power plants for the production of electrical power in space vehicles'. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in FY 1964, 1965, and 1966. A summary of the program's effort was compiled in 1967. The delayed critical experiments served as a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated 253 stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. 'The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.' The experiment studied within this evaluation was the first of the series and had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Information for this evaluation was compiled from Reference 1 and 2, reports on subsequent experiments in the series, and the experimental logbook as well as from communication with the experimenter, John T. Mihalczo.

  19. IAEA coordinated research projects on core physics benchmarks for high temperature gas-cooled reactors

    SciTech Connect

    Methnani, M.

    2006-07-01

    High-temperature Gas-Cooled Reactor (HTGR) designs present special computational challenges related to their core physics characteristics, in particular neutron streaming, double heterogeneities, impurities and the random distribution of coated fuel particles in the graphite matrix. In recent years, two consecutive IAEA Coordinated Research Projects (CRP 1 and CRP 5) have focused on code-to-code and code-to-experiment comparisons of representative benchmarks run by several participating international institutes. While the PROTEUS critical HTR experiments provided the test data reference for CRP-1, the more recent CRP-5 data has been made available by the HTTR, HTR-10 and ASTRA test facilities. Other benchmark cases are being considered for the GT-MHR and PBMR core designs. This paper overviews the scope and some sample results of both coordinated research projects. (authors)

  20. Current Reactor Physics Benchmark Activities at the Idaho National Laboratory

    SciTech Connect

    John D. Bess; Margaret A. Marshall; Mackenzie L. Gorham; Joseph Christensen; James C. Turnbull; Kim Clark

    2011-11-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) [1] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) [2] were established to preserve integral reactor physics and criticality experiment data for present and future research. These valuable assets provide the basis for recording, developing, and validating our integral nuclear data, and experimental and computational methods. These projects are managed through the Idaho National Laboratory (INL) and the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD-NEA). Staff and students at the Department of Energy - Idaho (DOE-ID) and INL are engaged in the development of benchmarks to support ongoing research activities. These benchmarks include reactors or assemblies that support Next Generation Nuclear Plant (NGNP) research, space nuclear Fission Surface Power System (FSPS) design validation, and currently operational facilities in Southeastern Idaho.

  1. An Overview of the International Reactor Physics Experiment Evaluation Project

    SciTech Connect

    Briggs, J. Blair; Gulliford, Jim

    2014-10-09

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

  2. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  3. Plasma Physics Lab and the Tokamak Fusion Test Reactor, 1989

    ScienceCinema

    None

    2016-07-12

    From the Princeton University Archives: Promotional video about the Plasma Physics Lab and the new Tokamak Fusion Test Reactor (TFTR), with footage of the interior, machines, and scientists at work. This film is discussed in the audiovisual blog of the Seeley G. Mudd Manuscript Library, which holds the archives of Princeton University.

  4. 10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal....

  5. 10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal....

  6. NEUTRONIC REACTOR

    DOEpatents

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  7. Effects of Spatial Variations in Packing Fraction on Reactor Physics Parameters in Pebble-Bed Reactors

    SciTech Connect

    William K. Terry; A. M. Ougouag; Farzad Rahnema; Michael Scott McKinley

    2003-04-01

    The well-known spatial variation of packing fraction near the outer boundary of a pebble-bed reactor core is cited. The ramifications of this variation are explored with the MCNP computer code. It is found that the variation has negligible effects on the global reactor physics parameters extracted from the MCNP calculations for use in analysis by diffusion-theory codes, but for local reaction rates the effects of the variation are naturally important. Included is some preliminary work in using first-order perturbation theory for estimating the effect of the spatial variation of packing fraction on the core eigenvalue and the fision density distribution.

  8. Effects of Spatial Variations in Packing Fraction of Reactor Physics Parameters in Pebble-Bed Reactors

    SciTech Connect

    Terry, W K; Ougouag, A M; Rahnema, F; Mckinley, M S

    2003-06-11

    The well-known spatial variation of packing fraction near the outer boundary of a pebble-bed reactor core is cited. The ramifications of this variation are explored with the MCNP computer code. It is found that the variation has negligible effects on the global reactor physics parameters extracted from the MCNP calculations for use in analysis by diffusion-theory codes, but for local reaction rates the effects of the variation are naturally important. Included is some preliminary work in using first-order perturbation theory for estimating the effect of the spatial variation of packing fraction on the core eigenvalue and the fission density distribution.

  9. Graphite for fusion energy applications

    SciTech Connect

    Eatherly, W.P.; Clausing, R.E.; Strehlow, R.A.; Kennedy, C.R.; Mioduszewski, P.K.

    1987-03-01

    Graphite is in widespread and beneficial use in present fusion energy devices. This report reflects the view of graphite materials scientists on using graphite in fusion devices. Graphite properties are discussed with emphasis on application to fusion reactors. This report is intended to be introductory and descriptive and is not intended to serve as a definitive information source. (JDH)

  10. Effect of physical properties of Cu-Ni-graphite composites on tribological characteristics by grey correlation analysis

    NASA Astrophysics Data System (ADS)

    Wang, Yiran; Gao, Yimin; Sun, Liang; Li, YeFei; Zheng, Baochao; Zhai, Wenyan

    Cu-Ni-graphite composites are intended for using as switch slide baseplates materials in high-speed railways industrials. The tribological characteristics of Cu-Ni-graphite composites with different graphite content were affected by physical properties and were generated variation correspondingly. It is difficult to study the correlation degree between tribological characteristics and physical properties by the means of experiments. However, grey correlation analysis (GCA) is a suitable mathematic method for researching the correlation with each factor. In this study, Cu-Ni-graphite composites were prepared by powder metallurgy and the correlation degree between tribological characteristics (Cu-Ni-graphite composites slid against U75V steel) and physical properties were calculated by GCA method. The results showed that through the calculating by GCA method, compared with physical properties of Cu-Ni-graphite composites, the most effective way to reduce the friction coefficient is to improve the relative density as well as the increasing of hardness can also bring down the friction coefficient usefully. The best way to reduce wear rate is the increasing of work of rupture and flexural strength, when these two properties have a small amount of ascension, the wear rate would be rapidly decreasing. The GCA results could help to improve materials' performance and give reasonable advices to the following studies.

  11. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    SciTech Connect

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

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

    SciTech Connect

    Heeger, Karsten M.

    2014-09-13

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

  13. Lessons Learned from the Application of Bulk Characterization to Individual Containers on the Brookhaven Graphite Research Reactor Decommissioning Project at Brookhaven National Laboratory - 12056

    SciTech Connect

    Kneitel, Terri; Rocco, Diane

    2012-07-01

    When conducting environmental cleanup or decommissioning projects, characterization of the material to be removed is often performed when the material is in-situ. The actual demolition or excavation and removal of the material can result in individual containers that vary significantly from the original bulk characterization profile. This variance, if not detected, can result in individual containers exceeding Department of Transportation regulations or waste disposal site acceptance criteria. Bulk waste characterization processes were performed to initially characterize the Brookhaven Graphite Research Reactor (BGRR) graphite pile and this information was utilized to characterize all of the containers of graphite. When the last waste container was generated containing graphite dust from the bottom of the pile, but no solid graphite blocks, the material contents were significantly different in composition from the bulk waste characterization. This error resulted in exceedance of the disposal site waste acceptance criteria. Brookhaven Science Associates initiated an in-depth investigation to identify the root causes of this failure and to develop appropriate corrective actions. The lessons learned at BNL have applicability to other cleanup and demolition projects which characterize their wastes in bulk or in-situ and then extend that characterization to individual containers. (authors)

  14. Effect of hexagonal boron nitride and graphite nanoparticles on the mechanical and physical properties of magnesium

    NASA Astrophysics Data System (ADS)

    Drozd, Z.; Trojanová, Z.; Arlic, U.; Kasakewitsch, A.; Dash, K.

    2017-07-01

    Magnesium nanocomposites reinforced with 3 vol.% of hexagonal boron nitride (hBN) and ultrafine-grained magnesium with 3 vol.% of graphite (Gr) nanoparticles were prepared by milling followed by hot extrusion. The influence of the hBN and Gr nanoparticles on the mechanical and physical properties of magnesium were subsequently investigated. The microstructure of the nanocomposites was analysed by electron microscopy. The nanocomposites were deformed in compression over a wide temperature range from room temperature to 300 °C and true stress-true strain curves were determined. The linear thermal expansion coefficient was measured over a wide temperature range from room temperature up to 400 °C. Pre-deformation in compression was used to estimate the influence of twins on the thermal expansion coefficient. The results are analysed and possible physical mechanisms for the observed mechanical and physical properties are discussed.

  15. Summary of ORSphere critical and reactor physics measurements

    NASA Astrophysics Data System (ADS)

    Marshall, Margaret A.; Bess, John D.

    2017-09-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

  16. Summary of ORSphere Critical and Reactor Physics Measurements

    SciTech Connect

    Marshall, Margaret A.; Bess, John D.

    2016-09-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J. J. Lynn, and J. R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is summary summarize all the critical and reactor physics measurements evaluations and, when possible, to compare them to GODIVA experiment results.

  17. On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

    SciTech Connect

    Chauvin, J.P.; Blaise, P.; Lyoussi, A.

    2015-07-01

    The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physics calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)

  18. 78 FR 31821 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-28

    ... COMMISSION 10 CFR Part 73 RIN 3150-AI64 Physical Protection of Shipments of Irradiated Reactor Fuel AGENCY... (NRC) is issuing Revision 2 of NUREG-0561, ``Physical Protection of Shipments of Irradiated Reactor... regulations for the transport of irradiated reactor fuel at Sec. 73.37 of Title 10 of the Code of...

  19. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

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

  20. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

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

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

    SciTech Connect

    Not Available

    1992-04-01

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

  2. Brazing graphite to graphite

    DOEpatents

    Peterson, George R.

    1976-01-01

    Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of virtually graphite.

  3. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    SciTech Connect

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  4. CRITICAL CONFIGURATION AND PHYSICS MEASUREMENTS FOR GRAPHITE REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH)

    SciTech Connect

    Margaret A. Marshall

    2012-03-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of 253 unmoderated stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.” The first experiment in the series was evaluated in HEU-COMP-FAST-001. It had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank (References 1 and 2). The second experiment in the series, performed in early 1963, which is studied in this evaluation, had the 253 fuel tubes at a 1.506-cm triangular lattice in a 25.96 cm OD core tank and graphite reflectors on all sides. The experiment has been determined to represent an acceptable benchmark experiment. Information for this evaluation was compiled from published reports on all three parts of the experimental series (Reference 1-5) and the experimental logbook as

  5. BENCHMARK EVALUATION OF THE START-UP CORE REACTOR PHYSICS MEASUREMENTS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR

    SciTech Connect

    John Darrell Bess

    2010-05-01

    The benchmark evaluation of the start-up core reactor physics measurements performed with Japan’s High Temperature Engineering Test Reactor, in support of the Next Generation Nuclear Plant Project and Very High Temperature Reactor Program activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Evaluation Project Handbook. Results provided include updated evaluation of the initial six critical core configurations (five annular and one fully-loaded). The calculated keff eigenvalues agree within 1s of the benchmark values. Reactor physics measurements that were evaluated include reactivity effects measurements such as excess reactivity during the core loading process and shutdown margins for the fully-loaded core, four isothermal temperature reactivity coefficient measurements for the fully-loaded core, and axial reaction rate measurements in the instrumentation columns of three core configurations. The calculated values agree well with the benchmark experiment measurements. Fully subcritical and warm critical configurations of the fully-loaded core were also assessed. The calculated keff eigenvalues for these two configurations also agree within 1s of the benchmark values. The reactor physics measurement data can be used in the validation and design development of future High Temperature Gas-cooled Reactor systems.

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

    SciTech Connect

    Evan Harpenau

    2011-07-15

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

  7. Physics of reactor safety. Quarterly report, October-December 1980. Volume IV

    SciTech Connect

    Not Available

    1981-02-01

    The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

  8. Physics of reactor safety. Volume II. Quarterly report, April-June 1980

    SciTech Connect

    Not Available

    1980-08-01

    The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

  9. Physics of reactor safety. Quarterly report, January-March 1980. Volume I

    SciTech Connect

    Not Available

    1980-05-01

    The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

  10. Reactor physics methods for the preconceptual core design of the advanced neutron source

    SciTech Connect

    Ryskamp, J.M.; Difilippo, F.C.; Primm, R.T. III

    1987-01-01

    Idaho National Engineering Laboratory (INEL) and Oak Ridge National Laboratory (ORNL) have been jointly working to develop and evaluate preconceptual reactor core configurations for the advanced neutron source. This paper reviews the reactor physics methods used to compute reactor parameters and demonstrates that the laboratories achieve good agreement on these parameters.

  11. Graphite for the nuclear industry

    SciTech Connect

    Burchell, T.D.; Fuller, E.L.; Romanoski, G.R.; Strizak, J.P.

    1991-01-01

    Graphite finds applications in both fission and fusion reactors. Fission reactors harness the energy liberated when heavy elements, such as uranium or plutonium, fragment or fission''. Reactors of this type have existed for nearly 50 years. The first nuclear fission reactor, Chicago Pile No. 1, was constructed of graphite under a football stand at Stagg Field, University of Chicago. Fusion energy devices will produce power by utilizing the energy produced when isotopes of the element hydrogen are fused together to form helium, the same reaction that powers our sun. The role of graphite is very different in these two reactor systems. Here we summarize the function of the graphite in fission and fusion reactors, detailing the reasons for their selection and discussing some of the challenges associated with their application in nuclear fission and fusion reactors. 10 refs., 15 figs., 1 tab.

  12. Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor

    SciTech Connect

    Pope, M. A.; Sen, R. S.; Ougouag, A. M.; Youinou, G.; Boer, B.

    2012-07-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)

  13. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    SciTech Connect

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  14. Modeling Fission Product Sorption in Graphite Structures

    SciTech Connect

    Szlufarska, Izabela; Morgan, Dane; Allen, Todd

    2013-04-08

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

  15. Optimization of an ionized metal physical vapor deposition reactor

    SciTech Connect

    Lu, J.; Kushner, M.J.

    1998-12-31

    Conventional sputtering for microelectronic fabrication produces poorly collimated neutral atom fluxes. Ion fluxes, however, can be accelerated and collimated by using a conventional dc or rf substrate bias. Hence, magnetron ionized metal physical vapor deposition (IMPVD) can produce highly ionized metal fluxes that can be used to fill high-aspect-ratio vias and trenches in microelectronic devices. Hopwood and Qian have examined design issues in IMPVD systems. In this study, a Design of Experiment (DOE) has been numerically performed for an IMPVD reactor using an inductively coupled plasma and a capacitively biased substrate. Gas pressure, reactor geometry, ICP power, and number of inductive coils are the design variables. Uniformity, magnitude, and ionization fraction of the depositing fluxes are the response variables. The influence of the design variables on the response variables is examined, with the goals of obtaining high uniformity, high magnitude, and high ionization fraction of the depositing metal fluxes. The computational tool used in this study is the two-dimensional Hybrid Plasma Equipment Model (HPEM). The aspect ratio of the reactor (height/radius) ranges from 0.5 to 1.0, the gas pressure ranges from 10 to 40 mTorr, the ICP power ranges from 0.5 to 2.0 kW, and the number of ICP coils ranges from 2 to 6. It was found that: (a) uniformity maximizes at high aspect ratio, low power, and high pressure; (b) flux magnitude maximizes at low aspect ratio, high power, and low pressure; (c) ionization fraction maximizes at high aspect ratio, high power, and high pressure.

  16. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    SciTech Connect

    Cheng, L.; Diamond, D.; Xu, J.; Carew, J.; Rorer, D.

    2004-03-31

    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

  17. The past, present, and future of test and research reactor physics

    SciTech Connect

    Ryskamp, J.M. )

    1992-01-01

    Reactor physics calculations have been performed on research reactors since the first one was built 50 yr ago under the University of Chicago stadium. Since then, reactor physics calculations have evolved from Fermi-age theory calculations performed with slide rules to three-dimensional, continuous-energy, coupled neutron-photon Monte Carlo computations performed with supercomputers and workstations. Such enormous progress in reactor physics leads us to believe that the next 50 year will be just as exciting. This paper reviews this transition from the past to the future.

  18. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    SciTech Connect

    Carew, J.; Hanson, A.; Xu, J.; Rorer, D.; Diamond, D.

    2003-08-26

    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

  19. Chapter 20: Graphite

    SciTech Connect

    Burchell, Timothy D

    2012-01-01

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

  20. Effect of specimen size and grain orientation on the mechanical and physical properties of NBG-18 nuclear graphite

    NASA Astrophysics Data System (ADS)

    Vasudevamurthy, G.; Byun, T. S.; Pappano, P.; Snead, L. L.; Burchell, T. D.

    2015-07-01

    We present here a comparison of the measured baseline mechanical and physical properties of with grain (WG) and against grain (AG) non-ASTM size NBG-18 graphite. The objectives of the experiments were twofold: (1) assess the variation in properties with grain orientation; (2) establish a correlation between specimen tensile strength and size. The tensile strength of the smallest sized (4 mm diameter) specimens were about 5% higher than the standard specimens (12 mm diameter) but still within one standard deviation of the ASTM specimen size indicating no significant dependence of strength on specimen size. The thermal expansion coefficient and elastic constants did not show significant dependence on specimen size. Experimental data indicated that the variation of thermal expansion coefficient and elastic constants were still within 5% between the different grain orientations, confirming the isotropic nature of NBG-18 graphite in physical properties.

  1. Effect of specimen size and grain orientation on the mechanical and physical properties of NBG-18 nuclear graphite

    SciTech Connect

    Vasudevamurthy, Gokul; Byun, Thak Sang; Pappano, Pete; Snead, Lance L.; Burchell, Tim D.

    2015-03-13

    We present here a comparison of the measured baseline mechanical and physical properties of with grain (WG) and against grain (AG) non-ASTM size NBG-18 graphite. The objectives of the experiments were twofold: (1) assess the variation in properties with grain orientation; (2) establish a correlation between specimen tensile strength and size. The tensile strength of the smallest sized (4 mm diameter) specimens were about 5% higher than the standard specimens (12 mmdiameter) but still within one standard deviation of the ASTM specimen size indicating no significant dependence of strength on specimen size. The thermal expansion coefficient and elastic constants did not show significant dependence on specimen size. Experimental data indicated that the variation of thermal expansion coefficient and elastic constants were still within 5% between the different grain orientations, confirming the isotropic nature of NBG-18 graphite in physical properties.

  2. Effect of specimen size and grain orientation on the mechanical and physical properties of NBG-18 nuclear graphite

    SciTech Connect

    Vasudevamurthy, G.; Byun, T. S.; Pappano, Pete; Snead, Lance L.; Burchell, Tim D.

    2015-03-13

    Here we present a comparison of the measured baseline mechanical and physical properties of with grain (WG) and against grain (AG) non-ASTM size NBG-18 graphite. The objectives of the experiments were twofold: (1) assess the variation in properties with grain orientation; (2) establish a correlation between specimen tensile strength and size. The tensile strength of the smallest sized (4 mm diameter) specimens were about 5% higher than the standard specimens (12 mm diameter) but still within one standard deviation of the ASTM specimen size indicating no significant dependence of strength on specimen size. The thermal expansion coefficient and elastic constants did not show significant dependence on specimen size. Lastly, experimental data indicated that the variation of thermal expansion coefficient and elastic constants were still within 5% between the different grain orientations, confirming the isotropic nature of NBG-18 graphite in physical properties.

  3. Effect of specimen size and grain orientation on the mechanical and physical properties of NBG-18 nuclear graphite

    DOE PAGES

    Vasudevamurthy, G.; Byun, T. S.; Pappano, Pete; ...

    2015-03-13

    Here we present a comparison of the measured baseline mechanical and physical properties of with grain (WG) and against grain (AG) non-ASTM size NBG-18 graphite. The objectives of the experiments were twofold: (1) assess the variation in properties with grain orientation; (2) establish a correlation between specimen tensile strength and size. The tensile strength of the smallest sized (4 mm diameter) specimens were about 5% higher than the standard specimens (12 mm diameter) but still within one standard deviation of the ASTM specimen size indicating no significant dependence of strength on specimen size. The thermal expansion coefficient and elastic constantsmore » did not show significant dependence on specimen size. Lastly, experimental data indicated that the variation of thermal expansion coefficient and elastic constants were still within 5% between the different grain orientations, confirming the isotropic nature of NBG-18 graphite in physical properties.« less

  4. Solid State Reactor Final Report

    SciTech Connect

    Mays, G.T.

    2004-03-10

    were undertaken: (1) establishing the design and safety-related basis via neutronic and reactor control assessments with the graphite foam as heat transfer medium; (2) evaluating the thermal performance of the graphite foam for heat removal, reactor stability, reactor operations, and overall core thermal characteristics; (3) characterizing the physical properties of the graphite foam under normal and irradiated conditions to determine any effects on structure, dimensional stability, thermal conductivity, and thermal expansion; and (4) developing a power conversion system design to match the reactor operating parameters.

  5. Opportunities for applied measurements using the PROSPECT antineutrino detector: reactor physics and safeguards

    NASA Astrophysics Data System (ADS)

    Bowden, Nathaniel; Prospect Collaboration

    2015-10-01

    Disagreement of reactor antineutrino spectrum and flux measurements with updated predictions indicates that we have much to learn about the complicated processes underlying antineutrino production in reactors, as well as hinting at new physics. A number of new efforts seek to address these questions, including the PROSPECT experiment planned at the HFIR research reactor. In addition to greatly advancing our understanding of reactor antineutrino emissions, PROSPECT can support a rich applied physics program. The detection technology developed for PROSPECT will enable precision antineutrino spectrum measurements close to essentially any reactor type. Here we describe how such measurements provide opportunities to probe fissile isotope and fission daughter distributions, and their potential use for reactor physics and safeguards applications. LLNL-ABS-673983. Prepared by LLNL under Contract DE-AC52-07NA27344.

  6. Recent improvements of reactor physics codes in MHI

    SciTech Connect

    Kosaka, Shinya Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki

    2015-12-31

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO’s Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

  7. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    SciTech Connect

    Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

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

    SciTech Connect

    Not Available

    1982-06-01

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

  9. Physically-based reduced-order capacity loss model for graphite anodes in Li-ion battery cells

    NASA Astrophysics Data System (ADS)

    Jin, Xing; Vora, Ashish; Hoshing, Vaidehi; Saha, Tridib; Shaver, Gregory; García, R. Edwin; Wasynczuk, Oleg; Varigonda, Subbarao

    2017-02-01

    Physically-based Li-ion electrochemical cell models have been shown capable of predicting cell performance and degradation, but are computationally expensive for optimization-oriented design applications. Faster empirical models have been developed from experimental data, but are not generalizable to operating conditions outside of the range established by the calibration data. In this paper, a reduced-order capacity-loss model for graphite anodes is derived based upon the salient physical loss mechanisms to improve computational efficiency without sacrificing model fidelity. This model captures the two primary degradation mechanisms that occur in the graphite anode of a typical lithium ion cell: a) capacity loss due to Solid Electrolyte Interface (SEI) layer growth, and b) capacity loss due to isolation of active material. The model is calibrated and validated for a commercial 2.3-Ah cell with a Lithium Iron Phosphate (LFP) cathode and graphite anode. One data set is used for calibration, another four experimental data sets are used for validation. The model matches experimental capacity degradation results within a 20% error. Moreover, the reported model is 2400× faster than currently existing more complex physically-based electrochemical models that are only slightly more accurate (in some cases).

  10. Physics of reactor safety. Quarterly report, July-September 1980. Volume III

    SciTech Connect

    Not Available

    1980-11-01

    This Quarterly progress report summarizes work done during the months of July-September 1980 in Argonne National Laboratory's Applied Physics and Components Technology Divisions for the Division of Reactor Safety Research of the US Nuclear Regulatory Commission. The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions. An executive summary is provided including a statement of the findings and recommendations of the report.

  11. Physical properties of nanofluid suspension of ferromagnetic graphite with high Zeta potential

    NASA Astrophysics Data System (ADS)

    Souza, N. S.; Rodrigues, A. D.; Cardoso, C. A.; Pardo, H.; Faccio, R.; Mombru, A. W.; Galzerani, J. C.; de Lima, O. F.; Sergeenkov, S.; Araujo-Moreira, F. M.

    2012-01-01

    We report on the magnetic properties and stability of nanofluid ferromagnetic graphite (NFMG) studied through the measurements of its magnetization hysteresis curves, Raman spectrum and the so-called Zeta potential. The obtained results suggest a robust ferromagnetic behavior of NFMG even at room temperature along with a good stability of the dispersed solution (with Zeta potential around 41.3 mV) and a good reactivity between magnetic graphite and CTAB type cationic surfactant.

  12. Consistent neutron-physical and thermal-physical calculations of fuel rods of VVER type reactors

    NASA Astrophysics Data System (ADS)

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

    2017-09-01

    For modeling the isotopic composition of fuel, and maximum temperatures at different moments of time, one can use different algorithms and codes. In connection with the development of new types of fuel assemblies and progress in computer technology, the task makes important to increase accuracy in modeling of the above characteristics of fuel assemblies during the operation. Calculations of neutronphysical characteristics of fuel rods are mainly based on models using averaged temperature, thermal conductivity factors, and heat power density. In this paper, complex approach is presented, based on modern algorithms, methods and codes to solve separate tasks of thermal conductivity, neutron transport, and nuclide transformation kinetics. It allows to perform neutron-physical and thermal-physical calculation of the reactor with detailed temperature distribution, with account of temperature-depending thermal conductivity and other characteristics. It was applied to studies of fuel cell of the VVER-1000 reactor. When developing new algorithms and programs, which should improve the accuracy of modeling the isotopic composition and maximum temperature in the fuel rod, it is necessary to have a set of test tasks for verification. The proposed approach can be used for development of such verification base for testing calculation of fuel rods of VVER type reactors

  13. Conceptual physics design of a compact torus fusion reactor (CTOR)

    SciTech Connect

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

    1980-07-01

    The general approach to fusion power embodied by field-reversed plasmoid configurations is reviewed within the context of a power reactor. A simple analytic formulation is developed and applied to the field-reversed theta pinch as a core plasma for a thermonuclear reactor. These calculations and results are based on a minimum power constraint and will serve as a basis for more exact and detailed reactor modeling.

  14. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    NASA Astrophysics Data System (ADS)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  15. Health physics aspects of advanced reactor licensing reviews

    SciTech Connect

    Hinson, C.S.

    1995-03-01

    The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.

  16. Simple Coupling of Reactor Physics Effects and Uncertain Nuances

    SciTech Connect

    Bays, Samuel

    2012-08-27

    The "Simple Coupling of Reactor Physics Effects and Uncertain Nuances" (SCORPEUN) code is a simple r-z 1-group neutron diffusion code where each r-mesh is coupled to a single-flow-channel model that represents all flow-channels in that r-mesh. This 1-D model assesses q=m*Cp*deletaT for each z-mesh in that channel. This flow channel model is then coupled to a simple 1-D heat conduction model for ascertaining the peak center-line fuel temperature in a hypothetical pin assigned to that flow channel. The code has property lookup capability for water, Na, Zirc, HT9, metalic fuel, oxide fuel, etc. It has linear interpolation features for micro-scopic cross-sections with respect to coolant density and fuel temperature. ***This last feature has not been fully tested and may need development***. The interpolated microscopic cross-sections are then combined (using the water density from the T/H calculation) to generate macroscopic diffusion coefficient, removal cross-section and nu-sigmaF for each r-z mesh of the neutron diffusion code.

  17. Measurement of the enthalpy and specific heat of a Be2C-graphite-UC2 reactor fuel material to 1980 K

    NASA Astrophysics Data System (ADS)

    Roth, E. P.

    1982-03-01

    The enthalpy and specific heat of a Be2C-Graphite-UC2 composite nuclear fuel material have been measured over the temperature range 298 1980 K using both differential scanning calorimetry and liquid argon vaporization calorimetry. The fuel material measured was developed at Sandia National Laboratories for use in pulsed test reactors. The material is a hot-pressed composite consisting of 40 vol% Be2C, 49.5 vol% graphite, 3.5 vol% UC2, and 7.0 vol% void. The specific heat was measured with the differential scanning calorimeter over the temperature range 298 950 K, while the enthalpy was measured over the range 1185 1980 K with the liquid argon vaporization calorimeter. The normal spectral emittance at a wavelength of 6.5×10-5 cm was also measured over the experimental temperature range. The combined experimental enthalpy data were fit using a spline routine and differentiated to give the specific heat. Comparison of the measured specific heat of the composite to the specific heat calculated by summing the contributions of the individual components indicates that the specific heat of the Be2C component differs significantly from literature values and is approximately 0.56 cal · g-1 · K -1 (2.3×103J · kg-1 · K -1) for temperatures above 1000 K.

  18. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.

    1960-11-22

    A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

  19. Synthesis, physical and chemical properties, and potential applications of graphite fluoride fibers

    NASA Technical Reports Server (NTRS)

    Hung, Ching-Cheh; Long, Martin; Stahl, Mark

    1987-01-01

    Graphite fluoride fibers can be produced by fluorinating pristine or intercalated graphite fibers. The higher the degree of graphitization of the fibers, the higher the temperature needed to reach the same degree of fluorination. Pitched based fibers were fluorinated to flourine-to-carbon atom rations between 0 and 1. The graphite fluoride fibers with a fluorine-to-carbon atom ration near 1 have extensive visible structural damage. On the other hand, fluorination of fibers pretreated with bromine or fluorine and bromine result in fibers with a fluorine-to-carbon atom ratio nearly equal to 0.5 with no visible structural damage. The electrical resistivity of the fibers is dependent upon the fluorine to carbon atom ratio and ranged from .01 to 10 to the 11th ohm/cm. The thermal conductivity of these fibers ranged from 5 to 73 W/m-k, which is much larger than the thermal conductivity of glass, which is the regular filler in epoxy composites. If graphite fluoride fibers are used as a filler in epoxy or PTFE, the resulting composite may be a high thermal conductivity material with an electrical resistivity in either the insulator or semiconductor range. The electrically insulating product may provide heat transfer with lower temperature gradients than many current electrical insulators. Potential applications are presented.

  20. Preliminary Reactor Physics Assessment of the HTR Module with 14% Enriched UCO Fuel

    SciTech Connect

    Gerhard Strydom; Hans D. Gougar

    2013-03-01

    The high temperature reactor (HTR) Module (Lohnert, 1990) is a graphite-moderated, helium cooled pebble bed design that has been extensively used as a reference template for the former South African (Matzner, 2004) and current Chinese (Zhang et al., 2009) HTR programs. This design utilizes spherical fuel elements packed into a dynamic pebble bed, consisting of tri-structural isotropic (TRISO) coated uranium oxide (UO2) 500 µm fuel kernels with a U-235 enrichment of 7.8% and a heavy metal loading of 7 g per pebble. This fuel type was previously qualified for use in Germany for pebble bed HTRs, as well as undergoing re-qualification in South Africa for the PBMR project. It is also the fuel type being tested for use in the high temperature reactor (HTR-PM) under construction in China. In the United States, however, a different TRISO fuel form is the subject of a qualification program. The U.S. experience with HTRs has been focused upon the batch-fueled prismatic reactor in which TRISO particles are embedded in cylindrical compacts and stacked inside the graphite blocks which comprise the core. Under this type of operating regime, a smaller TRISO with a different composition and enrichment performs better than the fuel historically used in PBRs. Fuel kernels and compacting techniques more suited to prismatic core duty are currently being developed and qualified under the U.S. Department of Energy's Advanced Gas Reactor (AGR) fuel development program and in support of the Next Generation Nuclear Plant project. Interest in the pebble bed concept remains high, however, and a study was undertaken by the authors to assess the viability of using AGR fuel in a pebble bed reactor. Using the German HTR Module as the reference plant, key neutronic and thermal-hydraulic parameters were compared between the nominal design and one fueled with the fuel that is the focus of the AGR program.

  1. Oxidation Resistant Graphite Studies

    SciTech Connect

    W. Windes; R. Smith

    2014-07-01

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

  2. Characterization of Epoxy Functionalized Graphite Nanoparticles and the Physical Properties of Epoxy Matrix Nanocomposites

    NASA Technical Reports Server (NTRS)

    Miller, Sandi G.; Bauer, Jonathan L.; Maryanski, Michael J.; Heimann, Paula J.; Barlow, Jeremy P.; Gosau, Jan-Michael; Allred, Ronald E.

    2010-01-01

    This work presents a novel approach to the functionalization of graphite nanoparticles. The technique provides a mechanism for covalent bonding between the filler and matrix, with minimal disruption to the sp2 hybridization of the pristine graphene sheet. Functionalization proceeded by covalently bonding an epoxy monomer to the surface of expanded graphite, via a coupling agent, such that the epoxy concentration was measured as approximately 4 wt.%. The impact of dispersing this material into an epoxy resin was evaluated with respect to the mechanical properties and electrical conductivity of the graphite-epoxy nanocomposite. At a loading as low as 0.5 wt.%, the electrical conductivity was increased by five orders of magnitude relative to the base resin. The material yield strength was increased by 30% and Young s modulus by 50%. These results were realized without compromise to the resin toughness.

  3. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    SciTech Connect

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  4. Thermal and physical properties of graphite intercalation compounds. Final report 15 July 77-30 June 82

    SciTech Connect

    Onn, D.G.

    1982-01-01

    Five major contributions to research in graphite intercalation compounds (GIC's) have been made. They are (1) the discovery of superconductivity in the mercurographitides (KHgC8 and RbHgC8) which was first seen in low temperature specific heat (Cp) studies, (2) that low-energy phonon states appear to play a role in suppressing Tc for superconducting GIC's and may suppress superconductivity altogether, (3) the re-awakening of interest in magnetic graphite intercalation compounds arising in part from our specific heat studies which suggest the possibility of a magnetic spin-glass state in FeCl3 and NiCl2 compounds, and (4) the confirmation that a low density of electronic states is common to a wide class of acceptor intercalation compounds. In addition, (5) it permitted completion of research that showed for the first time the university of 'twin' phase transitions in donor alkali metal GIC's below stage 1. Of the above, (1), (3) and (5) have led to a wealth of further research by other groups in recent years and have had lasting influence in this research area. Research performed under this grant was devoted primarily to the determination of the low temperature physical properties of a wide range of graphite intercalation compounds (GIC's) and the interpretation of these properties. In addition some new GIC's were synthesized and transport studies initiated elsewhere were completed.

  5. Inspection methods for physical protection Task III review of other agencies' physical security activities for research reactors

    SciTech Connect

    1980-01-01

    In Task I of this project, the current Nuclear Regulatory Commission (NRC) position-on physical security practices and procedures at research reactors were reviewed. In the second task, a sampling of the physical security plans was presented and the three actual reactor sites described in the security plans were visited. The purpose of Task III is to review other agencies' physical security activities for research reactors. During this phase, the actions, procedures and policies of two domestic and two foreign agencies other than the NRC that relate to the research reactor community were examined. The agencies examined were: International Atomic Energy Agency; Canadian Atomic Energy Control Board; Department of Energy; and American Nuclear Insurers.

  6. 78 FR 69139 - Physical Security-Design Certification and Operating Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-18

    ... Nuclear Energy Institute (NEI) submitted a letter on October 9, 2013 (Agencywide Documents Access and... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Physical Security--Design Certification and Operating Reactors AGENCY: Nuclear...

  7. Code System for Reactor Physics and Fuel Cycle Simulation.

    SciTech Connect

    TEUCHERT, E.

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.

  8. The Treatment of PPCP-Containing Sewage in an Anoxic/Aerobic Reactor Coupled with a Novel Design of Solid Plain Graphite-Plates Microbial Fuel Cell

    PubMed Central

    Chang, Yi-Tang; Yang, Chu-Wen; Chang, Yu-Jie; Chang, Ting-Chieh; Wei, Da-Jiun

    2014-01-01

    Synthetic sewage containing high concentrations of pharmaceuticals and personal care products (PPCPs, mg/L level) was treated using an anoxic/aerobic (A/O) reactor coupled with a microbial fuel cell (MFC) at hydraulic retention time (HRT) of 8 h. A novel design of solid plain graphite plates (SPGRPs) was used for the high surface area biodegradation of the PPCP-containing sewage and for the generation of electricity. The average CODCr and total nitrogen removal efficiencies achieved were 97.20% and 83.75%, respectively. High removal efficiencies of pharmaceuticals, including acetaminophen, ibuprofen, and sulfamethoxazole, were also obtained and ranged from 98.21% to 99.89%. A maximum power density of 532.61 mW/cm2 and a maximum coulombic efficiency of 25.20% were measured for the SPGRP MFC at the anode. Distinct differences in the bacterial community were presented at various locations including the mixed liquor suspended solids and biofilms. The bacterial groups involved in PPCP biodegradation were identified as Dechloromonas spp., Sphingomonas sp., and Pseudomonas aeruginosa. This design, which couples an A/O reactor with a novel design of SPGRP MFC, allows the simultaneous removal of PPCPs and successful electricity production. PMID:25197659

  9. The treatment of PPCP-containing sewage in an anoxic/aerobic reactor coupled with a novel design of solid plain graphite-plates microbial fuel cell.

    PubMed

    Chang, Yi-Tang; Yang, Chu-Wen; Chang, Yu-Jie; Chang, Ting-Chieh; Wei, Da-Jiun

    2014-01-01

    Synthetic sewage containing high concentrations of pharmaceuticals and personal care products (PPCPs, mg/L level) was treated using an anoxic/aerobic (A/O) reactor coupled with a microbial fuel cell (MFC) at hydraulic retention time (HRT) of 8 h. A novel design of solid plain graphite plates (SPGRPs) was used for the high surface area biodegradation of the PPCP-containing sewage and for the generation of electricity. The average CODCr and total nitrogen removal efficiencies achieved were 97.20% and 83.75%, respectively. High removal efficiencies of pharmaceuticals, including acetaminophen, ibuprofen, and sulfamethoxazole, were also obtained and ranged from 98.21% to 99.89%. A maximum power density of 532.61 mW/cm(2) and a maximum coulombic efficiency of 25.20% were measured for the SPGRP MFC at the anode. Distinct differences in the bacterial community were presented at various locations including the mixed liquor suspended solids and biofilms. The bacterial groups involved in PPCP biodegradation were identified as Dechloromonas spp., Sphingomonas sp., and Pseudomonas aeruginosa. This design, which couples an A/O reactor with a novel design of SPGRP MFC, allows the simultaneous removal of PPCPs and successful electricity production.

  10. Physical characteristics of GE (General Electric) BWR (boiling-water reactor) fuel assemblies

    SciTech Connect

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs.

  11. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... fuel in transit. 73.37 Section 73.37 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.37 Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives....

  12. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... fuel in transit. 73.37 Section 73.37 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.37 Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives....

  13. 76 FR 5102 - Draft NUREG-0561, Revision 2; Physical Protection of Shipments of Irradiated Reactor Fuel...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-28

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 73 RIN 3150-AI64 Draft NUREG-0561, Revision 2; Physical...-0561, ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This document provides guidance on implementing the provisions of proposed 10 CFR Part 73.37, ``Requirements for Physical Protection...

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

    SciTech Connect

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

    2012-11-01

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

  15. Analysis and evaluation of ZPPR (Zero Power Physics Reactor) critical experiments for a 100 kilowatt-electric space reactor

    SciTech Connect

    McFarlane, H.F.; Collins, P.J.; Carpenter, S.G.; Olsen, D.N.; Smith, D.M.; Schaefer, R.W. ); Doncals, R.A.; Andre, S.V.; Porter, C.A. ); Cowan, C.L; Stewart, S.L.; Protsik, R. . Astro Space Div.)

    1990-01-01

    ZPPR critical experiments were used for physics testing the reactor design of the SP-100, a 100-kW thermoelectric LMR that is being developed to provide electrical power for space applications. These tests validated all key physics characteristics of the design, including the ultimate safety in the event of a launch or re-entry accident. Both the experiments and the analysis required the use of techniques not previously applied to fast reactor designs. A few significant discrepancies between the experimental and calculated results leave opportunities for further optimization. An initial investigation has been made into application of the ZPPR-20 results, along with those of other relevant integral data, to the SP-100 design. 13 refs., 5 figs., 7 tabs.

  16. Effect of oxidation on physical properties of neutron irradiated nuclear grade graphite

    NASA Astrophysics Data System (ADS)

    Matsuo, Hideto

    1986-04-01

    Changes in thermal conductivity, electrical resistivity and Young's modulus due to thermal oxidation in air were studied for the neutron irradiated nuclear grade graphite IG-11. Samples were irradiated and then oxidized in air at 450°C up to the maximum weight loss of 27%. Neutron irradiation caused the increases of the rate of oxidation in air, electrical resistivity and Young's modulus, and the decrease of thermal conductivity as well. Subsequent oxidation led to the results of the increase of electrical resistivity, and the decreases of Young's modulus and thermal conductivity. An analytical expression was given to the present experimental results, and the tendencies of the changes in the properties of neutron irradiated nuclear grade graphite due to the oxidation were clarified to be the same as those of unirradiated samples.

  17. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    SciTech Connect

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  18. Optimisation and fabrication of a composite pyrolytic graphite monochromator for the Pelican instrument at the ANSTO OPAL reactor

    NASA Astrophysics Data System (ADS)

    Freund, A. K.; Yu, D. H.

    2011-04-01

    The triple monochromator for the TOF neutron spectrometer Pelican at ANSTO has been fully optimised in terms of overall performance, including the determination of the thickness of the pyrolytic graphite crystals. A total of 24 composite crystals were designed and fabricated. The calculated optimum thickness of 1.3 mm and the length of 15 cm of the monochromator crystals, that are not available commercially, were obtained by cleaving and soldering with indium. An extensive characterisation of the crystals using X-ray and neutron diffraction was conducted before and after the cleaving and bonding processes. The results proved that no damage was introduced during fabrication and showed that the design goals were fully met. The measured peak reflectivity and rocking curve widths were indeed in an excellent agreement with theory. In addition to the superior efficiency of the triple monochromator achieved by this novel approach, the amount of the crystal material required could be reduced by 1/3.

  19. An overview of the current status of resonance theory in reactor physics applications

    NASA Astrophysics Data System (ADS)

    Hwang, R. N.

    The neutron resonance phenomena constitute one of the most fundamental subjects in nuclear physics as well as in reactor physics. It is the area where the concepts of nuclear interaction and the treatment of the neutronic balance in reactor lattices become intertwined. The later requires the detailed knowledge of resonance structures of many nuclide of practical interest to the development of nuclear energy. The key issue of the resonance treatment in reactor applications is directly associated with the use of the microscopic cross sections in the macroscopic reactor cells with a wide range of composition, temperature,and geometric configurations. It gives rise to the so called self-shielding effect. The accurate estimations of such an effect are essential not only in the calculation of the criticality of a reactor but also from the point of view of safety considerations. The latter manefests through the Doppler effect particularly crucial to the fast reactor development. The task of accurate treatment of the self-shielding effect, however, is by no means simple. In fact, it is perhaps the most complicated problem in neutron physics which, strictly speaking, requires the dependence of many physical variables. Two important elements of particular interest are: (1) a concise description of the resonance cross sections as a function of energy and temperature; (2) accurate estimation of the corresponding neutron flux where appropriate. These topics will be discussed from both the historical as well as the state-of-art perspectives.

  20. GRAPHITE BLOCKS ARE ARRAYED IN "THERMAL COLUMN" ON NORTH SIDE ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    GRAPHITE BLOCKS ARE ARRAYED IN "THERMAL COLUMN" ON NORTH SIDE OF REACTOR. INL NEGATIVE NO. 4000. Unknown Photographer, 12/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  1. High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

    PubMed Central

    Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

    2014-01-01

    An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

  2. High-Resolution Coupled Physics Solvers for Analysing Fine-Scale Nuclear Reactor Design Problems

    SciTech Connect

    Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

    2014-06-30

    An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.

  3. High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.

    PubMed

    Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

    2014-08-06

    An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.

  4. Evaluation of the HTR-10 Reactor as a Benchmark for Physics Code QA

    SciTech Connect

    William K. Terry; Soon Sam Kim; Leland M. Montierth; Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-09-01

    The HTR-10 is a small (10 MWt) pebble-bed research reactor intended to develop pebble-bed reactor (PBR) technology in China. It will be used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and co-generation of heat, and provide experience in PBR design, operation, and construction. As the only currently operating PBR in the world, the HTR-10 can provide data of great interest to everyone involved in PBR technology. In particular, if it yields data of sufficient quality, it can be used as a benchmark for assessing the accuracy of computer codes proposed for use in PBR analysis. This paper summarizes the evaluation for the International Reactor Physics Experiment Evaluation Project (IRPhEP) of data obtained in measurements of the HTR-10’s initial criticality experiment for use as benchmarks for reactor physics codes.

  5. Graphite technology development plan

    SciTech Connect

    1986-07-01

    This document presents the plan for the graphite technology development required to support the design of the 350 MW(t) Modular HTGR within the US National Gas-Cooled Reactor Program. Besides descriptions of the required technology development, cost estimates, and schedules, the plan also includes the associated design functions and design requirements.

  6. Reactor Physics Characterization of Transmutation Targeting Options in a Sodium Fast Reactor

    SciTech Connect

    Samuel E. Bays

    2007-04-01

    In sodium fast reactor designs, the fuel related inherent negative reactivity feedback is accomplished mainly through parasitic capture in U-238. However for an efficient minor actinide burning system, it is desirable to reduce or eliminate U-238 entirely to suppress further transuranic actinide generation. Consequently, reactivity feedback is accomplished by enhancing axial neutron streaming during a loss of coolant void situation. This is done by flattening “pancake” the active core geometry. Flattening the reactor also increases axial leakage which removes neutrons that could otherwise be used to destroy minor actinides. Therefore, it is important to tailor the neutron spectrum in the core for optimized feedback and minor actinide destruction simultaneously by using minor actinide and fission product targets.

  7. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  8. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    SciTech Connect

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  9. Spent graphite fuel element processing

    SciTech Connect

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

    1981-07-01

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

  10. Nuclear Data Measurements for 21st Century Reactor Physics Applications

    SciTech Connect

    Rahmat Aryaeinejad; Jerald D. Cole; Mark W. Drigert; James K. Jewell; Christopher A. McGrath; David W. Nigg; Edward L. Reber

    2003-03-01

    The United States Department of Energy (DOE), Office of Nuclear Energy (NE) has embarked on a long-term program to significantly advance the science and technology of nuclear energy. This is in response to the overall national plan for accelerated development of domestic energy resources on several fronts, punctuated by recent dramatic events that have emphasized the need for the US to reduce its dependence on foreign petroleum supplies. Key aspects of the DOE-NE agenda are embodied in the Generation-IV (Gen-IV) advanced nuclear energy systems development program and in the Advanced Fuel Cycle (AFC) program. The planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current nuclear power reactor systems as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The success of the overall NE effort will depend not only on sophisticated system development and engineering, but also on the advances in the supporting sciences and technologies. Of these, one of the most important is the improvement of the relevant fundamental nuclear science data bases, especially the evaluated neutron interaction cross section files that serve as the foundation of all reactor system designs, operating strategies, and fuel cycle engineering activities. The new concepts for reactors and fuel cycles involve the use of transuranic nuclides that were previously of little interest, and where experimentally measured information is lacking. The current state of the cross section database for some of these nuclides is such that design computations for advanced fast-spectrum reactor systems and fuel cycles that incorporate such materials in significant quantities are meaningful only for approximate conceptual applications. No actual system could reliably be designed according to currently accepted standards, nor

  11. Fort St. Vrain graphite site mechanical separation concept selection. Final report

    SciTech Connect

    Berry, S.M.

    1993-09-01

    One of the alternatives to the disposal of the Fort St. Vrain (FSV) reactor spent nuclear fuel involves the separation of the fuel rods composed of compacts from the graphite fuel block assembly. After the separation of these two components, the empty graphite fuel blocks would be disposed of as a low level waste (provided the appropriate requirements are met) and the fuel compacts would be treated as high level waste material. This report deals with the mechanical separation aspects concerning physical disassembly of the FSV graphite fuel element into the empty graphite fuel blocks and fuel compacts. This report recommends that a drilling technique is the preferred choice for accessing the, fuel channel holes and that each hole is drilled separately. This report does not cover any techniques or methods to separate the triso fuel particles from the graphite matrix of the fuel compacts.

  12. 10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

    Code of Federal Regulations, 2010 CFR

    2010-01-01

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

  13. Graphite on graphite

    NASA Astrophysics Data System (ADS)

    Volovik, G. E.; Pudalov, V. M.

    2016-12-01

    We propose potential geometry for fabrication of the graphite sheets with atomically smooth edges. For such sheets with Bernal stacking, the electron-electron interaction and topology should cause sufficiently high density of states resulting in the high temperature of either spin ordering or superconducting pairing.

  14. REFLECTOR FOR NEUTRONIC REACTORS

    DOEpatents

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  15. Physical particularities of nuclear reactors using heavy moderators of neutrons

    SciTech Connect

    Kulikov, G. G. Shmelev, A. N.

    2016-12-15

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  16. Physical particularities of nuclear reactors using heavy moderators of neutrons

    NASA Astrophysics Data System (ADS)

    Kulikov, G. G.; Shmelev, A. N.

    2016-12-01

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using 233U as a fissile nuclide and 232Th and 231Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  17. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  18. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  19. Global variance reduction for Monte Carlo reactor physics calculations

    SciTech Connect

    Zhang, Q.; Abdel-Khalik, H. S.

    2013-07-01

    Over the past few decades, hybrid Monte-Carlo-Deterministic (MC-DT) techniques have been mostly focusing on the development of techniques primarily with shielding applications in mind, i.e. problems featuring a limited number of responses. This paper focuses on the application of a new hybrid MC-DT technique: the SUBSPACE method, for reactor analysis calculation. The SUBSPACE method is designed to overcome the lack of efficiency that hampers the application of MC methods in routine analysis calculations on the assembly level where typically one needs to execute the flux solver in the order of 10{sup 3}-10{sup 5} times. It places high premium on attaining high computational efficiency for reactor analysis application by identifying and capitalizing on the existing correlations between responses of interest. This paper places particular emphasis on using the SUBSPACE method for preparing homogenized few-group cross section sets on the assembly level for subsequent use in full-core diffusion calculations. A BWR assembly model is employed to calculate homogenized few-group cross sections for different burn-up steps. It is found that using the SUBSPACE method significant speedup can be achieved over the state of the art FW-CADIS method. While the presented speed-up alone is not sufficient to render the MC method competitive with the DT method, we believe this work will become a major step on the way of leveraging the accuracy of MC calculations for assembly calculations. (authors)

  20. Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations

    DOE PAGES

    Palmiotti, Giuseppe; Salvatores, Massimo

    2012-01-01

    The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described.

  1. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    SciTech Connect

    Shemon, E. R.; Grudzinski, J. J.; Lee, C. H.; Thomas, J. W.; Yu, Y. Q.

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  2. Effect of Graphite Nanoplate Morphology on the Dispersion and Physical Properties of Polycarbonate Based Composites

    PubMed Central

    Müller, Michael Thomas; Hilarius, Konrad; Liebscher, Marco; Lellinger, Dirk; Alig, Ingo; Pötschke, Petra

    2017-01-01

    The influence of the morphology of industrial graphite nanoplate (GNP) materials on their dispersion in polycarbonate (PC) is studied. Three GNP morphology types were identified, namely lamellar, fragmented or compact structure. The dispersion evolution of all GNP types in PC is similar with varying melt temperature, screw speed, or mixing time during melt mixing. Increased shear stress reduces the size of GNP primary structures, whereby the GNP aspect ratio decreases. A significant GNP exfoliation to individual or few graphene layers could not be achieved under the selected melt mixing conditions. The resulting GNP macrodispersion depends on the individual GNP morphology, particle sizes and bulk density and is clearly reflected in the composite’s electrical, thermal, mechanical, and gas barrier properties. Based on a comparison with carbon nanotubes (CNT) and carbon black (CB), CNT are recommended in regard to electrical conductivity, whereas, for thermal conductive or gas barrier application, GNP is preferred. PMID:28772907

  3. Effect of Graphite Nanoplate Morphology on the Dispersion and Physical Properties of Polycarbonate Based Composites.

    PubMed

    Müller, Michael Thomas; Hilarius, Konrad; Liebscher, Marco; Lellinger, Dirk; Alig, Ingo; Pötschke, Petra

    2017-05-18

    The influence of the morphology of industrial graphite nanoplate (GNP) materials on their dispersion in polycarbonate (PC) is studied. Three GNP morphology types were identified, namely lamellar, fragmented or compact structure. The dispersion evolution of all GNP types in PC is similar with varying melt temperature, screw speed, or mixing time during melt mixing. Increased shear stress reduces the size of GNP primary structures, whereby the GNP aspect ratio decreases. A significant GNP exfoliation to individual or few graphene layers could not be achieved under the selected melt mixing conditions. The resulting GNP macrodispersion depends on the individual GNP morphology, particle sizes and bulk density and is clearly reflected in the composite's electrical, thermal, mechanical, and gas barrier properties. Based on a comparison with carbon nanotubes (CNT) and carbon black (CB), CNT are recommended in regard to electrical conductivity, whereas, for thermal conductive or gas barrier application, GNP is preferred.

  4. Micro-physical properties of carbonaceous aerosol particles generated by laser ablation of a graphite target

    NASA Astrophysics Data System (ADS)

    Ajtai, T.; Utry, N.; Pintér, M.; Tápai, Cs.; Kecskeméti, G.; Smausz, T.; Hopp, B.; Bozóki, Z.; Szabó, G.

    2014-09-01

    In this work the authors propose laser ablation as a highly versatile tool for carbonaceous aerosol generation. The generated carbonaceous particles can be used as a model aerosol for atmospheric black carbon. Various microphysical properties including mass concentration, size distribution and morphology of aerosol particles generated by laser ablation of a high purity graphite sample were investigated in detail. These measurements proved that the proposed method can be used to generate both primary particles and fractal aggregates with a high yield. As a further advantage of the method the size distribution of the generated aerosol can cover a wide range, and can be tuned accurately with laser fluence, the ambient composition or with the volumetric flow rate of the carrier gas.

  5. Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit

    SciTech Connect

    Merzari, E.; Shemon, E. R.; Yu, Y. Q.; Thomas, J. W.; Obabko, A.; Jain, Rajeev; Mahadevan, Vijay; Tautges, Timothy; Solberg, Jerome; Ferencz, Robert Mark; Whitesides, R.

    2015-12-21

    This report describes to employ SHARP to perform a first-of-a-kind analysis of the core radial expansion phenomenon in an SFR. This effort required significant advances in the framework Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit used to drive the coupled simulations, manipulate the mesh in response to the deformation of the geometry, and generate the necessary modified mesh files. Furthermore, the model geometry is fairly complex, and consistent mesh generation for the three physics modules required significant effort. Fully-integrated simulations of a 7-assembly mini-core test problem have been performed, and the results are presented here. Physics models of a full-core model of the Advanced Burner Test Reactor have also been developed for each of the three physics modules. Standalone results of each of the three physics modules for the ABTR are presented here, which provides a demonstration of the feasibility of the fully-integrated simulation.

  6. Neutronic reactor

    DOEpatents

    Lewis, Warren R.

    1978-05-30

    A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

  7. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  8. Uranium Oxide Aerosol Transport in Porous Graphite

    SciTech Connect

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

    2012-01-23

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

  9. Evolution of the core physics concept for the Canadian supercritical water reactor

    SciTech Connect

    Pencer, J.; Colton, A.; Wang, X.; Gaudet, M.; Hamilton, H.; Yetisir, M.

    2013-07-01

    The supercritical water cooled reactor (SCWR) is one of the advanced reactor concepts chosen by the GEN-IV International Forum (GIF) for research and development efforts. Canada's contribution is the Canadian SCWR, a heavy water moderated, pressure tube supercritical light water cooled reactor. Recent developments in the SCWR lattice and core concepts, primarily the introduction of a large central flow tube filled with coolant combined with a two-ring fuel assembly, have enabled significant improvements compared to earlier concepts. These improvements include a reduction in coolant void reactivity (CVR) by more than 10 mk, and an almost 40% increase in fuel exit burnup, which is achieved via balanced power distribution between the fuel pins in the fuel assembly. In this paper the evolution of the physics concept is reviewed, and the present lattice and core physics concepts are presented.

  10. Capability assessment for application of clay mixture as barrier material for irradiated zirconium alloy structure elements long-term processing for storage during decommissioning of uranium-graphite nuclear reactors

    NASA Astrophysics Data System (ADS)

    Kotlyarevskiy, S. G.; Pavliuk, A. O.; Zakharova, E. V.; Volkova, A. G.

    2016-06-01

    The radionuclide composition and the activity level of the irradiated zirconium alloy E110, the radionuclide immobilization strength and the retention properties of the mixed clay barrier material with respect to the radionuclides identified in the alloy were investigated to perform the safety assessment of handling structural units of zirconium alloy used for the technological channels in uranium-graphite reactors. The irradiated zirconium alloy waste contained the following activation products: 93mNb and the long-lived 94Nb, 93Zr radionuclides. Radionuclides of 60Co, 137Cs, 90Sr, and actinides were also present in the alloy. In the course of the runs no leaching of niobium and zirconium isotopes from the E110 alloy was detected. Leach rates were observed merely for 60Co and 137Cs present in the deposits formed on the internal surface of technological channels. The radionuclides present were effectively adsorbed by the barrier material. To ensure the localization of radionuclides in case of the radionuclide migration from the irradiated zirconium alloy into the barrier material, the sorption properties were determined of the barrier material used for creating the long-term storage point for the graphite stack from uranium-graphite reactors.

  11. AGC-2 Graphite Preirradiation Data Package

    SciTech Connect

    David Swank; Joseph Lord; David Rohrbaugh; William Windes

    2012-10-01

    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.

  12. Inhibition of Oxidation in Nuclear Graphite

    SciTech Connect

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

    2013-10-01

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

  13. Students' Assessment of Interactive Distance Experimentation in Nuclear Reactor Physics Laboratory Education

    ERIC Educational Resources Information Center

    Malkawi, Salaheddin; Al-Araidah, Omar

    2013-01-01

    Laboratory experiments develop students' skills in dealing with laboratory instruments and physical processes with the objective of reinforcing the understanding of the investigated subject. In nuclear engineering, where research reactors play a vital role in the practical education of students, the high cost and long construction time of research…

  14. Students' Assessment of Interactive Distance Experimentation in Nuclear Reactor Physics Laboratory Education

    ERIC Educational Resources Information Center

    Malkawi, Salaheddin; Al-Araidah, Omar

    2013-01-01

    Laboratory experiments develop students' skills in dealing with laboratory instruments and physical processes with the objective of reinforcing the understanding of the investigated subject. In nuclear engineering, where research reactors play a vital role in the practical education of students, the high cost and long construction time of research…

  15. The integral fast reactor (IFR) concept: Physics of operation and safety

    SciTech Connect

    Wade, D.C.; Chang, Y.I.

    1987-01-01

    The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed.

  16. Direct physical exfoliation of few-layer graphene from graphite grown on a nickel foil using polydimethylsiloxane with tunable elasticity and adhesion

    NASA Astrophysics Data System (ADS)

    Yoo, Kwanghyun; Takei, Yusuke; Kim, Sungjin; Chiashi, Shohei; Maruyama, Shigeo; Matsumoto, Kiyoshi; Shimoyama, Isao

    2013-05-01

    We firstly introduce a facile method for the site-specific direct physical exfoliation of few-layer graphene sheets from cheap and easily enlargeable graphite grown on a Ni foil using an optimized polydimethylsiloxane (PDMS) stamp. By decreasing the PDMS cross-linking time, the PDMS elasticity is reduced to ˜52 kPa, similar to that of a typical gel. As a result of this process, the PDMS becomes more flexible yet remains in a handleable state as a stamp. Furthermore, the PDMS adhesion to a graphite/Ni surface, as measured by the peel strength, increases to ˜5.1 N m-1, which is approximately 17 times greater than that of typical PDMS. These optimized properties allow the PDMS stamp to have improved contact with the graphite/Ni surface, including the graphite wrinkles. This process is verified, and changes in surface morphology are observed using a 3D laser scanning microscope. Under conformal contact, the optimized PDMS stamp demonstrates the site-specific direct physical exfoliation of few-layer graphene sheets including mono- and bi-layer graphene sheets from the graphite/Ni substrate without the use of special equipment, conditions or chemicals. The number of layers of the exfoliated graphene and its high quality are revealed by the measured Raman spectroscopy. The exfoliation method using tunable elasticity and adhesion of the PDMS stamp can be used not only for cost-effective mass production of defect-less few-layer graphene from the graphite substrate for micro/nano device arrays but also for nano-contact printing of various structures, devices and cells.

  17. Direct physical exfoliation of few-layer graphene from graphite grown on a nickel foil using polydimethylsiloxane with tunable elasticity and adhesion.

    PubMed

    Yoo, Kwanghyun; Takei, Yusuke; Kim, Sungjin; Chiashi, Shohei; Maruyama, Shigeo; Matsumoto, Kiyoshi; Shimoyama, Isao

    2013-05-24

    We firstly introduce a facile method for the site-specific direct physical exfoliation of few-layer graphene sheets from cheap and easily enlargeable graphite grown on a Ni foil using an optimized polydimethylsiloxane (PDMS) stamp. By decreasing the PDMS cross-linking time, the PDMS elasticity is reduced to ∼52 kPa, similar to that of a typical gel. As a result of this process, the PDMS becomes more flexible yet remains in a handleable state as a stamp. Furthermore, the PDMS adhesion to a graphite/Ni surface, as measured by the peel strength, increases to ∼5.1 N m⁻¹, which is approximately 17 times greater than that of typical PDMS. These optimized properties allow the PDMS stamp to have improved contact with the graphite/Ni surface, including the graphite wrinkles. This process is verified, and changes in surface morphology are observed using a 3D laser scanning microscope. Under conformal contact, the optimized PDMS stamp demonstrates the site-specific direct physical exfoliation of few-layer graphene sheets including mono- and bi-layer graphene sheets from the graphite/Ni substrate without the use of special equipment, conditions or chemicals. The number of layers of the exfoliated graphene and its high quality are revealed by the measured Raman spectroscopy. The exfoliation method using tunable elasticity and adhesion of the PDMS stamp can be used not only for cost-effective mass production of defect-less few-layer graphene from the graphite substrate for micro/nano device arrays but also for nano-contact printing of various structures, devices and cells.

  18. Reactor physics studies in the GCFR Phase III critical assembly

    SciTech Connect

    Morman, J A

    1980-03-01

    The third phase of the gas cooled fast reactor (GCFR) program, ZPR-9 Assembly 30, is based on a multi-zoned core of PuO/sub 2/-UO/sub 2/ with radial and axial blankets of UO/sub 2/. Studies performed in this assembly will be compared to the previous phases of the GCFR program and will help to define parameters in this power-flattened demonstration plant-type core. Measurements in the Phase III program included small sample reactivity worths of various materials, central reaction rates and reaction rate distributions, absorption-to-fission ratios and the central point conversion ratio and the worth of steam entry into a small central zone. The reactivity change associated with the construction of a central pin zone in the core and axial blanket was measured. Reaction rate and steam entry measurements were repeated in the pin environment. Standard analysis methods using ENDF/B-IV data are described and the results are compared to measurements performed during the program.

  19. 2012 Gordon Research Conference on Graphitic Carbon Materials, Chemistry and Physics of - Formal Schedule and Speaker/Poster Program

    SciTech Connect

    Fertig, Herbert A.

    2012-06-22

    The Gordon Research Conference on GRAPHITIC CARBON MATERIALS, CHEMISTRY AND PHYSICS OF was held at the Davidson College, Davidson, North Carolina, June 17 – 22, 2012. The Conference was well-attended with 95 participants (attendees list attached). The attendees represented the spectrum of endeavor in this field coming from academia, industry, and government laboratories, both U.S. and foreign scientists, senior researchers, young investigators, and students. Of the 95 attendees, 41 voluntarily responded to a general inquiry regarding ethnicity which appears on our registration forms. Of the 41 respondents, 49% were Minorities – 5% Hispanic, 44% Asian and 0% African American. Approximately 2% of the participants at the 2012 meeting were women. In designing the formal speakers program, emphasis was placed on current unpublished research and discussion of the future target areas in this field. There was a conscious effort to stimulate lively discussion about the key issues in the field today. Time for formal presentations was limited in the interest of group discussions. In order that more scientists could communicate their most recent results, poster presentation time was scheduled. Attached is a copy of the formal schedule and speaker program and the poster program. In addition to these formal interactions, "free time" was scheduled to allow informal discussions. Such discussions are fostering new collaborations and joint efforts in the field. Carbon materials play an extremely important role in our society. They not only constitute the largest supply of energy we use today (i.e., coal) but also are the bases of many important technologies ranging from pencils, adsorbents, and metal strengtheners, to batteries and many others. Recent studies on graphitic carbon, including fullerenes, carbon nanotubes, and graphene, have further revealed novel optical and electrical properties, making it possible to use them for new applications in renewable energy as well as

  20. 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

    1986-05-01

    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.

  1. Physics considerations in the design of liquid metal reactors for transuranium element consumption

    SciTech Connect

    Khalil, H.; Hill, R.; Fujita, E.; Wade, D.

    1992-09-01

    The management of transuranic nuclides in liquid metal reactors (LMR`s) is considered based on the use of the Integral Fast Reactor (IFR) concept. Unique features of the IFR fuel cycle with respect to transuranic management are identified. These features are exploited together with the hard spectrum of LMR`s to demonstrate the neutronic feasibility of a wide range of transuranic management options ranging from efficient breeding to pure consumption. Core physics aspects of the development of a low sodium void worth transuranic burner concept are described. Neutronics performance parameters and reactivity feedback characteristics estimated for this core concept are presented.

  2. Physics considerations in the design of liquid metal reactors for transuranium element consumption

    SciTech Connect

    Khalil, H.; Hill, R.; Fujita, E.; Wade, D.

    1992-01-01

    The management of transuranic nuclides in liquid metal reactors (LMR's) is considered based on the use of the Integral Fast Reactor (IFR) concept. Unique features of the IFR fuel cycle with respect to transuranic management are identified. These features are exploited together with the hard spectrum of LMR's to demonstrate the neutronic feasibility of a wide range of transuranic management options ranging from efficient breeding to pure consumption. Core physics aspects of the development of a low sodium void worth transuranic burner concept are described. Neutronics performance parameters and reactivity feedback characteristics estimated for this core concept are presented.

  3. Applied antineutrino measurements: synergies between short-baseline physics and reactor safeguards

    NASA Astrophysics Data System (ADS)

    Classen, Timothy

    2016-09-01

    The experimental neutrino community has put forth an incredible effort into designing a multitude of detection systems to meet the needs of the short-baseline neutrino program. There is significant overlap between the requirements for these detectors, and those of a detector designed for reactor monitoring through antineutrinos. Here we describe how such technology improvements provide opportunities to probe fissile isotope and fission daughter distributions, and their potential use for reactor physics and safeguards applications. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

  4. Mechanical and physical properties of carbon-graphite fiber-reinforced polymers intended for implant suprastructures.

    PubMed

    Segerström, Susanna; Ruyter, I Eystein

    2007-09-01

    Mechanical properties and quality of fiber/matrix adhesion of poly(methyl methacrylate) (PMMA)-based materials, reinforced with carbon-graphite (CG) fibers that are able to remain in a plastic state until polymerization, were examined. Tubes of cleaned braided CG fibers were treated with a sizing resin. Two resin mixtures, resin A and resin B, stable in the fluid state and containing different cross-linking agents, were reinforced with CG fiber loadings of 24, 36, and 47 wt% (20, 29, and 38 vol.%). In addition, resin B was reinforced with 58 wt% (47 vol.%). After heat-polymerization, flexural strength and modulus were evaluated, both dry and after water storage. Coefficient of thermal expansion, longitudinally and in the transverse direction of the specimens, was determined. Adhesion between fibers and matrix was evaluated with scanning electron microscopy (SEM). Flexural properties and linear coefficient of thermal expansion were similar for both fiber composites. With increased fiber loading, flexural properties increased. For 47 wt% fibers in polymer A the flexural strength was 547.7 (28.12) MPa and for polymer B 563.3 (89.24) MPa when water saturated. Linear coefficient of thermal expansion was for 47 wt% CG fiber-reinforced polymers; -2.5 x 10(-6) degrees C-1 longitudinally and 62.4 x 10(-6) degrees C-1 in the transverse direction of the specimens. SEM revealed good adhesion between fibers and matrix. More porosity was observed with fiber loading of 58 wt%. The fiber treatment and the developed resin matrices resulted in good adhesion between CG fibers and matrix. The properties observed indicate a potential for implant-retained prostheses.

  5. Irradiation Creep in Graphite

    SciTech Connect

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

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

  6. Baseline Graphite Characterization: First Billet

    SciTech Connect

    Mark C. Carroll; Joe Lords; David Rohrbaugh

    2010-09-01

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

  7. Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

    SciTech Connect

    Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.; Mitchell, Mark R.; Gore, Bryan F.; Faris, Drury K.

    2009-10-09

    The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected to come from increasingly diverse educational and experiential backgrounds.

  8. ReactorHealth Physics operations at the NIST center for neutron research.

    PubMed

    Johnston, Thomas P

    2015-02-01

    Performing health physics and radiation safety functions under a special nuclear material license and a research and test reactor license at a major government research and development laboratory encompasses many elements not encountered by industrial, general, or broad scope licenses. This article reviews elements of the health physics and radiation safety program at the NIST Center for Neutron Research, including the early history and discovery of the neutron, applications of neutron research, reactor overview, safety and security of radiation sources and radioactive material, and general health physics procedures. These comprise precautions and control of tritium, training program, neutron beam sample processing, laboratory audits, inventory and leak tests, meter calibration, repair and evaluation, radioactive waste management, and emergency response. In addition, the radiation monitoring systems will be reviewed including confinement building monitoring, ventilation filter radiation monitors, secondary coolant monitors, gaseous fission product monitors, gas monitors, ventilation tritium monitor, and the plant effluent monitor systems.

  9. Empirical correlation of residual gamma radiation resulting from operation of the Health Physics Research Reactor

    SciTech Connect

    Chou, T.L.; Ragan, G.E.; Sims, C.S.

    1985-04-01

    An empirical equation has been developed which gives gamma dose equivalent rate as a function of time, distance, and fission yield after a pulsed operation of Oak Ridge National Laboratory's (ORNL) unshielded Health Physics Research Reactor (HPRR). A related expression which is applicable to steady-state reactor operation has been mathematically derived from the aforementioned empirical equation. The two relations can be used to predict the gamma dose equivalent rate to within 25% for times between 1 minute and 90 minutes after reactor shutdown. Similar agreement is expected for up to several days. In most cases the relations are expected to overestimate the gamma dose equivalent rate. 5 refs., 4 figs., 1 tab.

  10. Multi-physics design and analyses of long life reactors for lunar outposts

    NASA Astrophysics Data System (ADS)

    Schriener, Timothy M.

    event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete

  11. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  12. Multi-physics nuclear reactor simulator for advanced nuclear engineering education

    SciTech Connect

    Yamamoto, A.

    2012-07-01

    Multi-physics nuclear reactor simulator, which aims to utilize for advanced nuclear engineering education, is being introduced to Nagoya Univ.. The simulator consists of the 'macroscopic' physics simulator and the 'microscopic' physics simulator. The former performs real time simulation of a whole nuclear power plant. The latter is responsible to more detail numerical simulations based on the sophisticated and precise numerical models, while taking into account the plant conditions obtained in the macroscopic physics simulator. Steady-state and kinetics core analyses, fuel mechanical analysis, fluid dynamics analysis, and sub-channel analysis can be carried out in the microscopic physics simulator. Simulation calculations are carried out through dedicated graphical user interface and the simulation results, i.e., spatial and temporal behaviors of major plant parameters are graphically shown. The simulator will provide a bridge between the 'theories' studied with textbooks and the 'physical behaviors' of actual nuclear power plants. (authors)

  13. Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis

    SciTech Connect

    Wilson, Paul; Evans, Thomas; Tautges, Tim

    2012-12-24

    This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well

  14. High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

    DOE PAGES

    Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; ...

    2014-06-30

    An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in ordermore » to reduce the overall numerical uncertainty while leveraging available computational resources. Finally, the coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.« less

  15. Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

    NASA Technical Reports Server (NTRS)

    Butler, C.; Albright, D.

    2007-01-01

    Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

  16. Graphite Gamma Scan Results

    SciTech Connect

    Mark W. Drigert

    2014-04-01

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

  17. Structure of different grades of nuclear graphite

    NASA Astrophysics Data System (ADS)

    Mironov, B. E.; Westwood, A. V. K.; Scott, A. J.; Brydson, R.; Jones, A. N.

    2012-07-01

    Owing to its low neutron absorption cross-section, large scattering cross section and thermal and chemical stability, graphite is a key component of operational nuclear reactors where it is used as a moderator, reflector and as major structural component for 90% of current UK nuclear plants. It is also of interest for use in developing the future high temperature gas-cooled reactors. The properties of the nuclear graphite are influenced by its structural characteristics, which change as a function of neutron irradiation, temperature and oxidation. The principal structural changes during neutron irradiation that affect the integrity and dimensions of nuclear graphite components, thereby affecting service lifetime, are that the a-axis contracts and the c-axis expands in the crystallites. Characterization of virgin graphite structure and of the damage evolution after irradiation of nuclear graphite has an important role to play in the understanding and development of materials used in current and future nuclear reactors, respectively.

  18. High surface area monodispersed Fe3O4 nanoparticles alone and on physical exfoliated graphite for improved supercapacitors

    NASA Astrophysics Data System (ADS)

    Sarno, Maria; Ponticorvo, Eleonora; Cirillo, Claudia

    2016-12-01

    Highly conductive, unsophisticated and easy to be obtained physical exfoliated graphite (PHG) supporting well dispersed magnetite, Fe3O4/PHG nanocomposite, has been prepared by a one-step chemical strategy and physico-chemical characterized. The nanocomposite, favoured by the a-polar nanoparticles (NPs) capping, results in a self-assembled monolayer of monodispersed Fe3O4, covering perfectly the hydrophobic surfaces of PHG. The nanocomposite as an electrode material was fabricated into a supercapacitor and characterized by cyclic voltammetry (CV) and galvanostatic charge-discharge measurements. It shows, after a suitable annealing, significant electrochemical properties (capacitance value of 787 F/g at 0.5 A g-1 and a Fe3O4/PHG weight ratio of 0.31) and good cycling stability (retention 91% after 30,000 cycles). Highly monodispersed very fine Fe3O4 NPs, covered by organic chains, have been also synthesized. The high surface area Fe3O4 NPs, after washing to leave a low content of organic chains able to avoid aggregation without excessively affecting the electrical properties of the material, exhibit remarkable pseudocapacitive activities, including the highest specific capacitance over reported for Fe3O4 (300 F/g at 0.5 A g-1).

  19. NEUTRONIC REACTOR SHIELDING

    DOEpatents

    Borst, L.B.

    1961-07-11

    A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

  20. Parallelisation of MONK with Coupling to Thermal Hydraulics and Gamma Heating Calculations for Reactor Physics Applications

    NASA Astrophysics Data System (ADS)

    Richards, Simon D.; Davies, Nigel; Armishaw, Malcolm J.; Dobson, Geoff P.; Wright, George A.

    2014-06-01

    Monte Carlo methods are increasingly being used for whole core reactor physics modelling. We describe a number of recent developments to the MONK nuclear criticality and reactor physics code to implement parallel processing, mesh-dependent burn-up and coupling to both thermal hydraulics and gamma transport codes. Results are presented which demonstrate the e_ects of gamma heating in a MONK calculation coupled to the MCBEND Monte Carlo shielding code. Experimental validation of the mesh-dependent tracking and gamma coupling methods is provided by comparison with the results of the NESSUS experiment. The gamma heating calculated by coupled MONK-MCBEND, and the neutron heating calculated by MONK, both compare well against measurement. Finally results are presented from a parallel MONK calculation of a highly detailed PWR benchmark model, which show encouraging speed-up factors on a small development cluster.

  1. Osiris: A Modern, High-Performance, Coupled, Multi-Physics Code For Nuclear Reactor Core Analysis

    SciTech Connect

    Procassini, R J; Chand, K K; Clouse, C J; Ferencz, R M; Grandy, J M; Henshaw, W D; Kramer, K J; Parsons, I D

    2007-02-26

    To meet the simulation needs of the GNEP program, LLNL is leveraging a suite of high-performance codes to be used in the development of a multi-physics tool for modeling nuclear reactor cores. The Osiris code project, which began last summer, is employing modern computational science techniques in the development of the individual physics modules and the coupling framework. Initial development is focused on coupling thermal-hydraulics and neutral-particle transport, while later phases of the project will add thermal-structural mechanics and isotope depletion. Osiris will be applicable to the design of existing and future reactor systems through the use of first-principles, coupled physics models with fine-scale spatial resolution in three dimensions and fine-scale particle-energy resolution. Our intent is to replace an existing set of legacy, serial codes which require significant approximations and assumptions, with an integrated, coupled code that permits the design of a reactor core using a first-principles physics approach on a wide range of computing platforms, including the world's most powerful parallel computers. A key research activity of this effort deals with the efficient and scalable coupling of physics modules which utilize rather disparate mesh topologies. Our approach allows each code module to use a mesh topology and resolution that is optimal for the physics being solved, and employs a mesh-mapping and data-transfer module to effect the coupling. Additional research is planned in the area of scalable, parallel thermal-hydraulics, high-spatial-accuracy depletion and coupled-physics simulation using Monte Carlo transport.

  2. GRAPHITE EXTRUSIONS

    DOEpatents

    Benziger, T.M.

    1959-01-20

    A new lubricant for graphite extrusion is described. In the past, graphite extrusion mixtures have bcen composed of coke or carbon black, together with a carbonaceous binder such as coal tar pitch, and a lubricant such as petrolatum or a colloidal suspension of graphite in glycerin or oil. Sinee sueh a lubricant is not soluble in, or compatible with the biiider liquid, such mixtures were difficult to extrude, and thc formed pieees lacked strength. This patent teaches tbe use of fatty acids as graphite extrusion lubricants and definite improvemcnts are realized thereby since the fatty acids are soluble in the binder liquid.

  3. Application of the physics of plasma sheaths to the modeling of RF plasma reactors

    NASA Astrophysics Data System (ADS)

    Metze, A.; Ernie, D. W.; Oskam, H. J.

    1986-11-01

    An equivalent circuit model is presented for a planar RF plasma reactor. The physical properties of the plasma sheath adjacent to the electrodes are incorporated in the model. The sheath capacitances and the conduction currents through the sheaths are time varying and have a highly nonlinear dependence on the potentials across the plasma sheaths. The model shows that the waveforms of the voltage differences across the sheaths are highly nonsinusoidal and agree with reported measurements.

  4. Application of a maximum-entropy technique to spherical geometry in radiative transfer and reactor physics

    NASA Astrophysics Data System (ADS)

    Degheidy, A. R.; Madkour, M. A.

    1993-05-01

    The maximum-entropy technique is used to solving three problems in radiative transfer and reactor physics involving spherical geometry. These problems are: (1) luminosity or the total energy emitted by a sphere, (2) neutron capture probability, and (3) the albedo problem. Numerical calculations are done and compared with the exact values as well as with Pade's approximant results. The comparisons show that the maximum-entropy results are very good and converge to the exact results.

  5. 10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule D Appendix D to Part 73 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37...

  6. 10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule D Appendix D to Part 73 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37...

  7. Thermal and neutron-physical features of the nuclear reactor for a power pulsation plant for space applications

    NASA Astrophysics Data System (ADS)

    Gordeev, É. G.; Kaminskii, A. S.; Konyukhov, G. V.; Pavshuk, V. A.; Turbina, T. A.

    2012-05-01

    We have explored the possibility of creating small-size reactors with a high power output with the provision of thermal stability and nuclear safety under standard operating conditions and in emergency situations. The neutron-physical features of such a reactor have been considered and variants of its designs preserving the main principles and approaches of nuclear rocket engine technology are presented.

  8. BFS, a Legacy to the International Reactor Physics, Criticality Safety, and Nuclear Data Communities

    SciTech Connect

    J. Blair Briggs; Anatoly Tsibulya; Yevgeniy Rozhikhin

    2012-03-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. Data provided by these two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades The Russian Federation has been a major contributor to both projects with the Institute of Physics and Power Engineering (IPPE) as the major contributor from the Russian Federation. Included in the benchmark specifications from the BFS facilities are 34 critical configurations from BFS-49, 61, 62, 73, 79, 81, 97, 99, and 101; spectral characteristics measurements from BFS-31, 42, 57, 59, 61, 62, 73, 97, 99, and 101; reactivity effects measurements from BFS-62-3A; reactivity coefficients and kinetics measurements from BFS-73; and reaction rate measurements from BFS-42, 61, 62, 73, 97, 99, and 101.

  9. A bibliography on finite element and related methods analysis in reactor physics computations (1971--1997)

    SciTech Connect

    Carpenter, D.C.

    1998-01-01

    This bibliography provides a list of references on finite element and related methods analysis in reactor physics computations. These references have been published in scientific journals, conference proceedings, technical reports, thesis/dissertations and as chapters in reference books from 1971 to the present. Both English and non-English references are included. All references contained in the bibliography are sorted alphabetically by the first author`s name and a subsort by date of publication. The majority of the references relate to reactor physics analysis using the finite element method. Related topics include the boundary element method, the boundary integral method, and the global element method. All aspects of reactor physics computations relating to these methods are included: diffusion theory, deterministic radiation and neutron transport theory, kinetics, fusion research, particle tracking in finite element grids, and applications. For user convenience, many of the listed references have been categorized. The list of references is not all inclusive. In general, nodal methods were purposely excluded, although a few references do demonstrate characteristics of finite element methodology using nodal methods (usually as a non-conforming element basis). This area could be expanded. The author is aware of several other references (conferences, thesis/dissertations, etc.) that were not able to be independently tracked using available resources and thus were not included in this listing.

  10. Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

    SciTech Connect

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

    2010-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The Japanese government approved construction of the HTTR in the 1989 fiscal year budget; construction began at the Oarai Research and Development Center in March 1991 and was completed May 1996. Fuel loading began July 1, 1998, from the core periphery. The first criticality was attained with an annular core on November 10, 1998 at 14:18, followed by a series of start-up core physics tests until a fully-loaded core was developed on December 16, 1998. Criticality tests were carried out into January 1999. The first full power operation with an average core outlet temperature of 850ºC was completed on December 7, 2001, and operational licensing of the HTTR was approved on March 6, 2002. The HTTR attained high temperature operation at 950 ºC in April 19, 2004. After a series of safety demonstration tests, it will be used as the heat source in a hydrogen production system by 2015. Hot zero-power critical, rise-to-power, irradiation, and safety demonstration testing , have also been performed with the HTTR, representing additional means for computational validation efforts. Power tests were performed in steps from 0 to 30 MW, with various tests performed at each step to confirm

  11. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    SciTech Connect

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  12. Graphite moderated (252)Cf source.

    PubMed

    Sajo-Bohus, Laszlo; Barros, Haydn; Greaves, Eduardo D; Vega-Carrillo, Hector Rene

    2015-06-01

    The Thorium molten-salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid-fuel reactor. The neutron source to run this subcritical reactor is a (252)Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the (252)Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

    SciTech Connect

    Wickham, A.J.; Drace, Z.

    2012-07-01

    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)

  14. DENSITY CONTROL IN A REACTOR

    DOEpatents

    Marshall, J. Jr.

    1961-10-24

    A reactor is described in which natural-uranium bodies are located in parallel channels which extend through the graphite mass in a regular lattice. The graphite mass has additional channels that are out of the lattice and contain no uranium. These additional channels decrease in number per unit volume of graphite from the center of the reactor to the exterior and have the effect of reducing the density of the graphite more at the center than at the exterior, thereby spreading neutron activity throughout the reactor. (AEC)

  15. Carbon-14 Graphitization Chemistry

    NASA Astrophysics Data System (ADS)

    Miller, James; Collon, Philippe; Laverne, Jay

    2014-09-01

    Accelerator Mass Spectrometry (AMS) is a process that allows for the analysis of mass of certain materials. It is a powerful process because it results in the ability to separate rare isotopes with very low abundances from a large background, which was previously impossible. Another advantage of AMS is that it only requires very small amounts of material for measurements. An important application of this process is radiocarbon dating because the rare 14C isotopes can be separated from the stable 14N background that is 10 to 13 orders of magnitude larger, and only small amounts of the old and fragile organic samples are necessary for measurement. Our group focuses on this radiocarbon dating through AMS. When performing AMS, the sample needs to be loaded into a cathode at the back of an ion source in order to produce a beam from the material to be analyzed. For carbon samples, the material must first be converted into graphite in order to be loaded into the cathode. My role in the group is to convert the organic substances into graphite. In order to graphitize the samples, a sample is first combusted to form carbon dioxide gas and then purified and reduced into the graphite form. After a couple weeks of research and with the help of various Physics professors, I developed a plan and began to construct the setup necessary to perform the graphitization. Once the apparatus is fully completed, the carbon samples will be graphitized and loaded into the AMS machine for analysis.

  16. Carbon structural materials for fusion reactors

    SciTech Connect

    Virgiliev, Yu.S.; Kurolenkin, E.I.

    1993-12-31

    This report describes properties of several structural carbon materials being investigated as materials for fusion reactors. Materials include: graphite, graphite doped with boron and titanium; and C-C composites. Radiation effects and additive effects are described.

  17. Neutrino Oscillation Physics with KamLAND: Reactor Antineutrinos and Beyond

    NASA Astrophysics Data System (ADS)

    Heeger, Karsten M.

    The discovery of flavor transformation in atmospheric, solar, and accelerator neutrinos has provided unambiguous evidence that neutrinos have mass and mix flavors. Data obtained in the past decade have revolutionized our understanding of neutrinos and provided the first evidence of physics beyond the Standard Model. In the long history of reactor neutrino physics, KamLAND has added to these recent discoveries the first direct observation of reactor overline{ν}_{e} disappearance, the evidence of spectral distortion as a signature of neutrino oscillation, and provided terrestrial confirmation for neutrino oscillation as the solution to the solar neutrino problem. With its long baseline of 175 km KamLAND makes the most precise determination of the mass splitting Δ m^{2}_{12} and, together with the solar neutrino experiments, has determined under the assumption of CPT invariance the oscillation parameters to unprecedented precision: Δ m^{2} = 7.9^{+0.6}_{-0.5} × 10^{-5} eV^{2}and tan^{2}θ = 0.40^{+0.10}_{-0.07}. Besides the measurement of the reactor overline{ν}_{e} flux, KamLAND has also observed geological antineutrinos from inside the Earth and set limits on the overline{ν}_{e} flux from the Sun. By purifying its liquid scintillator target and reducing internal detector backgrounds, KamLAND is preparing for the direct observation of low-energy, 7Be solar neutrinos that may allow the first direct test of the MSW mechanism of solar neutrino oscillation.

  18. Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

    SciTech Connect

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

    2014-10-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

  19. Newly Available Reactor Physics Benchmark data in the March 2011 Edition of the IRPhEP Handbook

    SciTech Connect

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

    2011-06-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the data are compromised, it is unlikely that any of these measurements would be repeated in the future. The purpose of the IRPhEP is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Several new evaluations have been prepared for inclusion in the March 2011 edition of the IRPhEP Handbook.

  20. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    SciTech Connect

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

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the ICSBEP and the IRPh

  1. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    SciTech Connect

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  2. Constitutive material model for the prediction of stresses in irradiated anisotropic graphite components

    NASA Astrophysics Data System (ADS)

    Tsang, Derek K. L.; Marsden, Barry J.

    2008-10-01

    As well as acting as a moderator and reflector, graphite is used as a structural component in many gas-cooled fission nuclear reactors. Therefore the ability to predict the structural integrity of the many graphite components which make up a graphite reactor core is important in safety case assessments and reactor core life prediction. This involves the prediction of the service life stresses in the individual graphite components. In this paper a material model for the prediction of stresses in anisotropic graphite is presented. The time-integrated non-linear irradiated graphite material model can be used for stress analysis of graphite components subject to both fast neutron irradiation and radiolytic oxidation. As an example a simple stress analysis of a typical reactor graphite component is presented along with a series of sensitivity studies aimed at investigating the importance of the various material property changes involved in graphite component stress prediction.

  3. Neutronic reactor thermal shield

    DOEpatents

    Wende, Charles W. J.

    1976-06-15

    1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

  4. Removal of carbon-14 from irradiated graphite

    NASA Astrophysics Data System (ADS)

    Dunzik-Gougar, Mary Lou; Smith, Tara E.

    2014-08-01

    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. On of the isotopes of great concern for long-term disposal of irradiated graphite is carbon-14 (14C), with a half-life of 5730 years. Study of irradiated graphite from some nuclear reactors indicates 14C is concentrated on the outer 5 mm of the graphite structure. The aim of the research presented here is to develop a practical method by which 14C can be removed. In parallel with these efforts, the same irradiated graphite material is being characterized to identify the chemical form of 14C in irradiated graphite. A nuclear-grade graphite, NBG-18, and a high-surface-area graphite foam, POCOFoam®, were exposed to liquid nitrogen (to increase the quantity of 14C precursor) and neutron-irradiated (1013 neutrons/cm2/s). During post-irradiation thermal treatment, graphite samples were heated in the presence of an inert carrier gas (with or without the addition of an oxidant gas), which carries off gaseous products released during treatment. Graphite gasification occurs via interaction with adsorbed oxygen complexes. Experiments in argon only were performed at 900 °C and 1400 °C to evaluate the selective removal of 14C. Thermal treatment also was performed with the addition of 3 and 5 vol% oxygen at temperatures 700 °C and 1400 °C. Thermal treatment experiments were evaluated for the effective selective removal of 14C. Lower temperatures and oxygen levels correlated to more efficient 14C removal.

  5. (Irradiation creep of graphite)

    SciTech Connect

    Kennedy, C.R.

    1990-12-21

    The traveler attended the Conference, International Symposium on Carbon, to present an invited paper, Irradiation Creep of Graphite,'' and chair one of the technical sessions. There were many papers of particular interest to ORNL and HTGR technology presented by the Japanese since they do not have a particular technology embargo and are quite open in describing their work and results. In particular, a paper describing the failure of Minor's law to predict the fatigue life of graphite was presented. Although the conference had an international flavor, it was dominated by the Japanese. This was primarily a result of geography; however, the work presented by the Japanese illustrated an internal program that is very comprehensive. This conference, a result of this program, was better than all other carbon conferences attended by the traveler. This conference emphasizes the need for US participation in international conferences in order to stay abreast of the rapidly expanding HTGR and graphite technology throughout the world. The United States is no longer a leader in some emerging technologies. The traveler was surprised by the Japanese position in their HTGR development. Their reactor is licensed and the major problem in their graphite program is how to eliminate it with the least perturbation now that most of the work has been done.

  6. Structural graphitic carbon foams

    SciTech Connect

    Kearns, K.M.; Anderson, H.J.

    1998-12-31

    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.

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

    SciTech Connect

    Mark C. Carroll; David T. Rohrbaugh

    2013-08-01

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

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

    SciTech Connect

    J. Blair Briggs; Dr. Enrico Sartori

    2005-09-01

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

  9. Feasibility of Isotopic Measurements: Graphite Isotopic Ratio Method

    SciTech Connect

    Wood, Thomas W.; Gerlach, David C.; Reid, Bruce D.; Morgan, W. C.

    2001-04-30

    This report addresses the feasibility of the laboratory measurements of isotopic ratios for selected trace constituents in irradiated nuclear-grade graphite, based on the results of a proof-of-principal experiment completed at Pacific Northwest National Laboratory (PNNL) in 1994. The estimation of graphite fluence through measurement of isotopic ratio changes in the impurity elements in the nuclear-grade graphite is referred to as the Graphite Isotope Ratio Method (GIRM). Combined with reactor core and fuel information, GIRM measurements can be employed to estimate cumulative materials production in graphite moderated reactors. This report documents the laboratory procedures and results from the initial measurements of irradiated graphite samples. The irradiated graphite samples were obtained from the C Reactor (one of several production reactors at Hanford) and from the French G-2 Reactor located at Marcoule. Analysis of the irradiated graphite samples indicated that replicable measurements of isotope ratios could be obtained from the fluence sensitive elements of Ti, Ca, Sr, and Ba. While these impurity elements are present in the nuclear-grade graphite in very low concentrations, measurement precision was typically on the order of a few tenths of a percent to just over 1 percent. Replicability of the measurements was also very good with measured values differing by less than 0.5 percent. The overall results of this initial proof-of-principal experiment are sufficiently encouraging that a demonstration of GIRM on a reactor scale basis is planned for FY-95.

  10. Initial Testing of the Microscopic Depletion Implementation in the MAMMOTH Reactor Physics Application

    SciTech Connect

    J. Ortensi; Y. Wang; S. Schunert; B.D. Ganapol; F.N. Gleicher; B. Baker; M.D. DeHart

    2016-09-01

    Present and new nuclear fuels that will be tested at the Transient Reactor Test (TREAT) facility will be analyzed with the MAMMOTH reactor physics application, currently under development, at Idaho National Laboratory. MAMMOTH natively couples the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. This system allows the irradiation of fuel from beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing within the same analysis mesh and, thus, retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. The Chebyshev Rational Approximation Method (CRAM) is the current state-of-the-art in the field, as was chosen to be the solver for the decay and transmutation equations. This report shows that the implementation of the CRAM solver within MAMMOTH is correct with various analytic benchmarks for decay and transmutation of nuclides. The results indicate that the solutions with CRAM order 16 achieve the level of precision of the benchmark. The CRAM solutions show little sensitivity to the time step size and consistently produce a high level of accuracy for isotopic decay for time steps of 1x10^11 years. Comparisons to DRAGON5 with 297 isotopes yield comparable results, but some differences need to be further analyzed.

  11. Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures

    NASA Astrophysics Data System (ADS)

    Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R.; Crowhurst, Jonathan C.; Weisz, David G.; Zaug, Joseph M.; Dai, Zurong; Radousky, Harry B.; Chernov, Alex; Ramon, Erick; Stavrou, Elissaios; Knight, Kim; Fabris, Andrea L.; Cappelli, Mark A.; Rose, Timothy P.

    2017-09-01

    We present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 < T < 5000 K) and atmospheric pressure. The reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after they pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.

  12. Reactor physics teaching and research in the Swiss nuclear engineering master

    SciTech Connect

    Chawla, R.

    2012-07-01

    Since 2008, a Master of Science program in Nuclear Engineering (NE) has been running in Switzerland, thanks to the combined efforts of the country's key players in nuclear teaching and research, viz. the Swiss Federal Inst.s of Technology at Lausanne (EPFL) and at Zurich (ETHZ), the Paul Scherrer Inst. (PSI) at Villigen and the Swiss Nuclear Utilities (Swissnuclear). The present paper, while outlining the academic program as a whole, lays emphasis on the reactor physics teaching and research training accorded to the students in the framework of the developed curriculum. (authors)

  13. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    SciTech Connect

    Strydom, Gerhard; Bostelmann, F.

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  14. Effects of space-dependent cross sections on core physics parameters for compact fast spectrum space power reactors

    NASA Astrophysics Data System (ADS)

    Lell, R. M.; Hanan, N. A.

    Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors were examined. Homogeneous and space dependent multigroup cross section set were generated in 11 and 27 groups for a representative fast reactor core. These cross sections were used to compute various reactor physics parameters for the reference core. Coarse group structure and neglect of space dependence in the generation procedure resulted in inaccurate computations of reactor flux and power distributions and in significant errors regarding estimates of core reactivity and control system worth. Delayed neutron fraction was insensitive to cross section treatment, and computed reactivity coefficients were only slightly sensitive. However, neutron lifetime was found to be very sensitive to cross section treatment. Deficiencies in multigroup cross sections are reflected in core nuclear design and, consequently, in system mechanical design.

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

    SciTech Connect

    Rosenthal, Murray Wilford

    2009-08-01

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

  16. A Roadmap and Discussion of Issues for Physics Analyses Required to Support Plutonium Disposition in VVER-1000 Reactors

    SciTech Connect

    Primm, R.T.; Drischler, J.D.; Pavlovichev, A.M. Styrine, Y.A.

    2000-06-01

    The purpose of this report is to document the physics analyses that must be performed to successfully disposition weapons-usable plutonium in VVER-1000 reactors in the Russian Federation. The report is a document to support programmatic and financial planning. It does not include documentation of the technical procedures by which physics analyses are performed, nor are the results of any analyses included.

  17. The New Cold Neutron Radiography Facility (CNRF) at the Mianyang Research Reactor of the China Academy of Engineering Physics

    NASA Astrophysics Data System (ADS)

    Bin, Tang; Heyong, Huo; Ke, Tang; Rogers, John; Haste, Martin; Christodoulou, Marios

    A new cold neutron radiography beamline has been designed and constructed for the Mianyang reactor at the Institute of Nuclear Physics and Chemistry of the China Academy of Engineering Physics. This paper describes the components of the system and demonstrates the achievable image resolution.

  18. Assessment of sensitivity of neutron-physical parameters of fast neutron reactor to purification of reprocessed fuel from minor actinides

    NASA Astrophysics Data System (ADS)

    Cherny, V. A.; Kochetkov, L. A.; Nevinitsa, A. I.

    2013-12-01

    The work is devoted to computational investigation of the dependence of basic physical parameters of fast neutron reactors on the degree of purification of plutonium from minor actinides obtained as a result of pyroelectrochemical reprocessing of spent nuclear fuel and used for manufacturing MOX fuel to be reloaded into the reactors mentioned. The investigations have shown that, in order to preserve such important parameters of a BN-800 type reactor as the criticality, the sodium void reactivity effect, the Doppler effect, and the efficiency of safety rods, it is possible to use the reprocessed fuel without separation of minor actinides for refueling (recharging) the core.

  19. Design and physical studies of fast reactor for bimodal space thermionic system with single-cell TFEs

    SciTech Connect

    Kirillov, E. Ya.; Klimov, A. V.; Ogloblin, B. G.; Radchenko, I. S.; Shumov, D. P.

    1997-01-10

    The paper presents the design studies and results of neutron-physical calculations of a fast nuclear reactor of a bimodal space thermionic system with single-cell thermionic fuel elements (TFE) designed for operation in two modes. These modes are (a) the propulsion mode making possible the system movement in outer space by the use of a reactive thrust generated by hydrogen heated in the reactor and (b) the electric power mode providing power supply to space vehicle-mounted systems with energy consumption level of 40kW(e) for a long time. The paper also discusses the problems of nuclear reactor safeguarding in an emergency.

  20. Surface characterization of TFTR first wall graphite tiles used during DT operations

    SciTech Connect

    Paffett, M. T.; Willms, R. S.; Gentile, C.; Skinner, C.

    2001-01-01

    Surface characterization studies were performed on graphite tiles used as first wall materials during DT operation of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. These ex situ analysis studies revealed a number of interesting and unexpected features. In this work we examined the spatial and (where possible) the depth distribution of impurity species deposited onto the plasma facing surfaces using Xray Photo-electron Spectroscopy (XPS) and Secondary Ion Mass Spectrometry (SIMS).

  1. Graphite Revisited

    NASA Astrophysics Data System (ADS)

    Draine, B. T.

    2016-11-01

    Laboratory measurements are used to constrain the dielectric tensor for graphite, from microwave to X-ray frequencies. The dielectric tensor is strongly anisotropic even at X-ray energies. The discrete dipole approximation is employed for accurate calculations of absorption and scattering by single-crystal graphite spheres and spheroids. For randomly oriented single-crystal grains, the so-called 1/3{--}2/3 approximation for calculating absorption and scattering cross sections is exact in the limit a/λ \\to 0 and provides better than ∼10% accuracy in the optical and UV even when a/λ is not small, but becomes increasingly inaccurate at infrared wavelengths, with errors as large as ∼40% at λ =10 μ {{m}}. For turbostratic graphite grains, the Bruggeman and Maxwell Garnett treatments yield similar cross sections in the optical and ultraviolet, but diverge in the infrared, with predicted cross sections differing by over an order of magnitude in the far-infrared. It is argued that the Maxwell Garnett estimate is likely to be more realistic, and is recommended. The out-of-plane lattice resonance of graphite near 11.5 μm may be observable in absorption with the MIRI spectrograph on James Webb Space Telescope. Aligned graphite grains, if present in the interstellar medium, could produce polarized X-ray absorption and polarized X-ray scattering near the carbon K edge.

  2. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  3. THE NEXT GENERATION NUCLEAR PLANT GRAPHITE PROGRAM

    SciTech Connect

    William E. Windes; Timothy D. Burchell; Robert L. Bratton

    2008-09-01

    Developing new nuclear grades of graphite used in the core of a High Temperature Gas-cooled Reactor (HTGR) is one of the critical development activities being pursued within the Next Generation Nuclear Plant (NGNP) program. Graphite’s thermal stability (in an inert gas environment), high compressive strength, fabricability, and cost effective price make it an ideal core structural material for the HTGR reactor design. While the general characteristics necessary for producing nuclear grade graphite are understood, historical “nuclear” grades no longer exist. New grades must be fabricated, characterized, and irradiated to demonstrate that current grades of graphite exhibit acceptable non-irradiated and irradiated properties upon which the thermo-mechanical design of the structural graphite in NGNP is based. The NGNP graphite R&D program has selected a handful of commercially available types for research and development activities necessary to qualify this nuclear grade graphite for use within the NGNP reactor. These activities fall within five primary areas; 1) material property characterization, 2) irradiated material property characterization, 3) modeling, and 4) ASTM test development, and 5) ASME code development efforts. Individual research and development activities within each area are being pursued with the ultimate goal of obtaining a commercial operating license for the nuclear graphite from the US NRC.

  4. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  5. AGC-2 Graphite Pre-irradiation Data Package

    SciTech Connect

    David Swank; Joseph Lord; David Rohrbaugh; William Windes

    2010-08-01

    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.

  6. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    SciTech Connect

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  7. A physical description of fission product behavior fuels for advanced power reactors.

    SciTech Connect

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

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

    SciTech Connect

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

    2014-06-10

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

  9. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    NASA Astrophysics Data System (ADS)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

  10. Physical and mechanical characteristics and chemical compatibility of aluminum nitride insulator coatings for fusion reactor applications

    SciTech Connect

    Natesan, K.; Rink, D.L.

    1996-04-01

    The blanket system is one of the most important components in a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of various concepts, including liquid metal, molten salt, water, and helium. Based on the requirements for an electrically insulating coating on the first-wall structural material to minimize the MHD pressure drop during the flow of liquid metal in a magnetic field, AlN was selected as a candidate coating material for the Li self-cooled blanket concept. This report discusses the results from an ongoing study of physical and mechanical characteristics and chemical compatibility of AlN electrical insulator coatings in a liquid Li environment. Details are presented on the AlN coating fabrication methods, and experimental data are reported for microstructures, chemistry of coatings, pretreatment of substrate, heat treatment of coatings, hardness data for coatings, coating/lithium interactions, and electrical resistance before and after exposure to lithium. Thermodynamic calculations are presented to establish regions of stability for AlN coatings in an Li environment as a function of O concentration and temperature, which can aid in-situ development of AlN coatings in Li.

  11. Verification of Unstructured Mesh Capabilities in MCNP6 for Reactor Physics Problems

    SciTech Connect

    Burke, Timothy P.; Martz, Roger L.; Kiedrowski, Brian C.; Martin, William R.

    2012-08-22

    New unstructured mesh capabilities in MCNP6 (developmental version during summer 2012) show potential for conducting multi-physics analyses by coupling MCNP to a finite element solver such as Abaqus/CAE[2]. Before these new capabilities can be utilized, the ability of MCNP to accurately estimate eigenvalues and pin powers using an unstructured mesh must first be verified. Previous work to verify the unstructured mesh capabilities in MCNP was accomplished using the Godiva sphere [1], and this work attempts to build on that. To accomplish this, a criticality benchmark and a fuel assembly benchmark were used for calculations in MCNP using both the Constructive Solid Geometry (CSG) native to MCNP and the unstructured mesh geometry generated using Abaqus/CAE. The Big Ten criticality benchmark [3] was modeled due to its geometry being similar to that of a reactor fuel pin. The C5G7 3-D Mixed Oxide (MOX) Fuel Assembly Benchmark [4] was modeled to test the unstructured mesh capabilities on a reactor-type problem.

  12. Physical modelling and adaptive predictive control of diffusion/LPCVD reactors

    NASA Astrophysics Data System (ADS)

    Dewaard, H.

    1992-12-01

    The aim of this study is to design a temperature controller for batch electric diffusion/low pressure chemical vapor deposition (LPCVD) furnaces, that complies with the increasingly more stringent requirements of VLSI processing. A mathematical model has been developed for batch electric diffusion/LPCVD reactors that are currently used in the semiconductor industry for the fabrication of micro-electronic devices. The model has been formulated in terms of partial integro-differential equations, which are derived from the basic energy conservation law of physics. The model takes into account the effects of radiation and conduction. Chapter 2 gives a detailed description of the furnace system and provides some insight into the processes that take place. In chapter 3, the model of the diffusion/LPPCVD furnace is derived. Chapter 4 deals with the design of a temperature control system for the diffusion/LPCVD reactor, that makes use of the model as developed in chapter 3. Chapter 5 gives the results of the control designs, both of simulation and of application on a real furnace. Results of the linear quadratic Gaussian controller, the (non-adaptive) reduced order controller, and the adaptive predictive controller are presented. Finally, in chapter 6, some conclusions are drawn and suggestions for further research are given.

  13. Preliminary Benchmark Evaluation of Japan’s High Temperature Engineering Test Reactor

    SciTech Connect

    John Darrell Bess

    2009-05-01

    A benchmark model of the initial fully-loaded start-up core critical of Japan’s High Temperature Engineering Test Reactor (HTTR) was developed to provide data in support of ongoing validation efforts of the Very High Temperature Reactor Program using publicly available resources. The HTTR is a 30 MWt test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. The benchmark was modeled using MCNP5 with various neutron cross-section libraries. An uncertainty evaluation was performed by perturbing the benchmark model and comparing the resultant eigenvalues. The calculated eigenvalues are approximately 2-3% greater than expected with an uncertainty of ±0.70%. The primary sources of uncertainty are the impurities in the core and reflector graphite. The release of additional HTTR data could effectively reduce the benchmark model uncertainties and bias. Sensitivity of the results to the graphite impurity content might imply that further evaluation of the graphite content could significantly improve calculated results. Proper characterization of graphite for future Next Generation Nuclear Power reactor designs will improve computational modeling capabilities. Current benchmarking activities include evaluation of the annular HTTR cores and assessment of the remaining start-up core physics experiments, including reactivity effects, reactivity coefficient, and reaction-rate distribution measurements. Long term benchmarking goals might include analyses of the hot zero-power critical, rise-to-power tests, and other irradiation, safety, and technical evaluations performed with the HTTR.

  14. AGC-3 Graphite Preirradiation Data Analysis Report

    SciTech Connect

    William Windes; David Swank; David Rohrbaugh; Joseph Lord

    2013-09-01

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

  15. Updating irradiated graphite disposal: Project 'GRAPA' and the international decommissioning network.

    PubMed

    Wickham, Anthony; Steinmetz, Hans-Jürgen; O'Sullivan, Patrick; Ojovan, Michael I

    2017-05-01

    Demonstrating competence in planning and executing the disposal of radioactive wastes is a key factor in the public perception of the nuclear power industry and must be demonstrated when making the case for new nuclear build. This work addresses the particular waste stream of irradiated graphite, mostly derived from reactor moderators and amounting to more than 250,000 tonnes world-wide. Use may be made of its unique chemical and physical properties to consider possible processing and disposal options outside the normal simple classifications and repository options for mixed low or intermediate-level wastes. The IAEA has an obvious involvement in radioactive waste disposal and has established a new project 'GRAPA' - Irradiated Graphite Processing Approaches - to encourage an international debate and collaborative work aimed at optimising and facilitating the treatment of irradiated graphite. Copyright © 2017 Elsevier Ltd. All rights reserved.

  16. An assessment of coupling algorithms for nuclear reactor core physics simulations

    SciTech Connect

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; Pawlowski, Roger; Toth, Alex; Kelley, C.T.; Evans, Thomas; Philip, Bobby

    2016-04-15

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

  17. Passive and Active Radiation Measurements Capability at the INL Zero Power Physics Reactor (ZPPR) Facility

    SciTech Connect

    Robert Neibert; John Zabriskie; Collin Knight; James L. Jones

    2010-12-01

    The Zero Power Physics Reactor (ZPPR) facility is a Department of Energy facility located in the Idaho National Laboratory’s (INL) Materials and Fuels Complex. It contains various nuclear and non-nuclear materials that are available to support many radiation measurement assessments. User-selected, single material, nuclear and non-nuclear materials can be readily utilized with ZPPR clamshell containers with almost no criticality concerns. If custom, multi-material configurations are desired, the ZPPR clamshell or an approved aluminum Inspection Object (IO) Box container may be utilized, yet each specific material configuration will require a criticality assessment. As an example of the specialized material configurations possible, the National Nuclear Security Agency’s Office of Nuclear Verification (NNSA/NA 243) has sponsored the assembly of six material configurations. These are shown in the Appendixes and have been designated for semi-permanent storage that can be available to support various radiation measurement applications.

  18. An assessment of coupling algorithms for nuclear reactor core physics simulations

    NASA Astrophysics Data System (ADS)

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; Pawlowski, Roger; Toth, Alex; Kelley, C. T.; Evans, Thomas; Philip, Bobby

    2016-04-01

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss-Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton-Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

  19. An assessment of coupling algorithms for nuclear reactor core physics simulations

    DOE PAGES

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; ...

    2016-04-01

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Furthermore, numerical simulations demonstrating the efficiency ofmore » JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.« less

  20. An assessment of coupling algorithms for nuclear reactor core physics simulations

    SciTech Connect

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; Pawlowski, Roger; Toth, Alex; Kelley, C. T.; Evans, Thomas; Philip, Bobby

    2016-04-01

    Here we evaluate the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product was developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Finally, both criticality (k-eigenvalue) and critical boron search problems are considered.

  1. Physical characterisation of the sludge produced in a sequencing batch biofilter granular reactor.

    PubMed

    Lotito, Adriana Maria; Di Iaconi, Claudio; Lotito, Vincenzo

    2012-10-15

    Sequencing batch biofilter granular reactor (SBBGR) is a recently developed biological wastewater treatment technology characterised by a very low sludge production, among other numerous advantages. Even if costs for sludge treatment and disposal are mainly dependent on the amount of sludge produced, sludge properties, especially those linked to solid-liquid separation, play a key role as well. In fact, such properties deeply influence the type of treatments sludge has to undergo before disposal and the final achievable solids concentration, strongly affecting treatment and disposal costs. As sludge from SBBGR is a special mixture of biofilm and aerobic granules, no information is available so far on its treatability. This study addresses the characterisation of the sludge produced from SBBGR in terms of some physical properties (settling properties, dewaterability, rheology). The results show that such sludge is characterised by good settling and dewatering properties, adding a new advantage for the full-scale application of SBBGR technology.

  2. An assessment of coupling algorithms for nuclear reactor core physics simulations

    DOE PAGES

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; ...

    2016-02-06

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Furthermore, numerical simulations demonstrating the efficiency ofmore » JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.« less

  3. An assessment of coupling algorithms for nuclear reactor core physics simulations

    SciTech Connect

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; Pawlowski, Roger; Toth, Alex; Kelley, C. T.; Evans, Thomas; Philip, Bobby

    2016-02-06

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Furthermore, numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

  4. Baseline Graphite Initial Mechanical Test Report

    SciTech Connect

    Mark Carroll; Randy Lloyd

    2009-09-01

    The Next Generation Nuclear Plant (NGNP) Project is tasked with selecting a high temperature gas reactor technology that will be capable of generating electricity and supplying large amounts of process heat. The NGNP is presently being designed as a helium-cooled high temperature gas reactor (HTGR) with a large graphite core. The graphite baseline characterization project is conducting the research and development (R&D) activities deemed necessary to fully qualify nuclear-grade graphite for use in the NGNP reactor. One of the major fundamental objectives of the project is establishing nonirradiated thermomechanical and thermophysical properties by characterizing lot-to-lot and billet-to-billet variations (for probabilistic baseline data needs) through extensive data collection and statistical analysis. The reactor core will be made up of stacks of graphite moderator blocks. In order to gain a more comprehensive understanding of the varying characteristics in a wide range of suitable graphites, any of which can be classified as “nuclear grade,” an experimental program has been initiated to develop an extensive database of the baseline characteristics of numerous candidate graphites. Various factors known to affect the properties of graphite will be investigated, including specimen size, spatial location within a graphite billet, specimen orientation within a billet (either parallel to [P] or transverse to [T] the long axis of the as-produced billet), and billet-to-billet variations within a lot or across different production lots. Because each data point is based on a certain position within a given billet of graphite, particular attention must be paid to the traceability of each specimen and its spatial location and orientation within each billet. The evaluation of these properties is discussed in the Graphite Technology Development Plan (Windes et. al 2007). One of the key components in the evaluation of these graphite types will be mechanical testing of

  5. Physics model of a gas-cooled fast reactor: Review and assessment

    SciTech Connect

    Choi, H.

    2012-07-01

    The current physics design and analysis model was reviewed and assessed for its application to a long-life gas-cooled fast reactor (GFR) design. The physics design uses MICROX, BURP, and DIF3D for the cross section generation, depletion calculation, and criticality and flux calculation, respectively. For the application to the long-life GFR, the depletion model was adjusted such that more lumped fission products are included in the burn-up chain to preserve the reaction rate and fuel mass. The performance of the physics design tools including the adjustment of the depletion model was assessed against Monte Carlo depletion calculations. The comparison has shown that the excess reactivity and cycle length of the long-life GFR are reasonably predicted. Some discrepancies were found at the beginning of cycle, which can be attributed to the differences between the nuclear data used in each model. Further studies will be carried out to update the cross section library of the MICROX code for agreement with the latest sets and to expand the fuel burn-up chain for the high burn-up and recycling fuel cycle analysis. (authors)

  6. Design and physical studies of fast reactor for bimodal space thermionic system with single-cell TFEs

    SciTech Connect

    Kirillov, E.Y.; Klimov, A.V.; Ogloblin, B.G.; Radchenko, I.S.; Shumov, D.P.

    1997-01-01

    The paper presents the design studies and results of neutron-physical calculations of a fast nuclear reactor of a bimodal space thermionic system with single-cell thermionic fuel elements (TFE) designed for operation in two modes. These modes are (a) the propulsion mode making possible the system movement in outer space by the use of a reactive thrust generated by hydrogen heated in the reactor and (b) the electric power mode providing power supply to space vehicle-mounted systems with energy consumption level of 40kW(e) for a long time. The paper also discusses the problems of nuclear reactor safeguarding in an emergency. {copyright} {ital 1997 American Institute of Physics.}

  7. Benchmarking of Graphite Reflected Critical Assemblies of UO2

    SciTech Connect

    Margaret A. Marshall; John D. Bess

    2011-11-01

    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

  8. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    NASA Astrophysics Data System (ADS)

    Stankovskiy, Evgeny Yuryevich

    In the recently completed RACE Project of the AFCI, accelerator-driven subcritical systems (ADS) experiments were conducted to develop technology of coupling accelerators to nuclear reactors. In these experiments electron accelerators induced photon-neutron reactions in heavy-metal targets to initiate fission reactions in ADS. Although the Idaho State University (ISU) RACE ADS was constructed only to develop measurement techniques for advanced experiments, many reactor kinetics experiments were conducted there. In the research reported in this dissertation, a method was developed to calculate kinetics parameters for measurement and calculation of the reactivity of ADS, a safety parameter that is necessary for control and monitoring of power production. Reactivity is measured in units of fraction of delayed versus prompt neutron from fission, a quantity that cannot be directly measured in far-subcritical reactors such as the ISU RACE configuration. A new technique is reported herein to calculate it accurately and to predict kinetic behavior of a far-subcritical ADS. Experiments conducted at ISU are first described and experimental data are presented before development of the kinetic theory used in the new computational method. Because of the complexity of the ISU ADS, the Monte-Carlo method as applied in the MCNP code is most suitable for modeling reactor kinetics. However, the standard method of calculating the delayed neutron fraction produces inaccurate values. A new method was developed and used herein to evaluate actual experiments. An advantage of this method is that its efficiency is independent of the fission yield of delayed neutrons, which makes it suitable for fuel with a minor actinide component (e.g. transmutation fuels). The implementation of this method is based on a correlated sampling technique which allows the accurate evaluation of delayed and prompt neutrons. The validity of the obtained results is indicated by good agreement between experimental

  9. 75 FR 67636 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-03

    ... Reactor Fuel AGENCY: Nuclear Regulatory Commission. ACTION: Notice of availability of draft guidance for... regulations pertaining to the transport of irradiated reactor fuel (for purposes of this rulemaking, the terms ``irradiated reactor fuel'' and ``spent nuclear fuel'' (SNF) are used interchangeably). The NRC has prepared...

  10. Nuclear Data - Their Importance and Application in Fission Reactor Physics Calculations

    NASA Astrophysics Data System (ADS)

    Schmidt, J. J.

    1991-01-01

    The following sections are included: * Introduction * Nuclear data - definition * Applications of nuclear data in nuclear fission reactor technology * Doppler temperature coefficient * Radiation damage and neutron dosimetry * Reactor shielding * Build-up and disposal of secondary actinide and fission product isotopes * Evaluated nuclear data files for nuclear fission reactor technology applications * References

  11. 10 CFR 73.35 - Requirements for physical protection of irradiated reactor fuel (100 grams or less) in transit.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... fuel (100 grams or less) in transit. 73.35 Section 73.35 Energy NUCLEAR REGULATORY COMMISSION... Transit § 73.35 Requirements for physical protection of irradiated reactor fuel (100 grams or less) in transit. Each licensee who transports, or delivers to a carrier for transport, in a single shipment,...

  12. METHOD FOR COATING GRAPHITE WITH NIOBIUM CARBIDE

    DOEpatents

    Kane, J.S.; Carpenter, J.H.; Krikorian, O.H.

    1962-01-16

    A method is given for coating graphite with a hard, tenacious layer of niobium carbide up to 30 mils or more thick. The method makes use of the discovery that niobium metal, if degassed and heated rapidly below the carburization temperature in contact with graphite, spreads, wets, and penetrates the graphite without carburization. The method includes the obvious steps of physically contacting niobium powders or other physical forms of niobium with graphite, degassing the assembly below the niobium melting point, e.g., 1400 deg C, heating to about 2200 to 2400 deg C within about 15 minutes while outgassing at a high volume throughput, and thereafter carburizing the niobium. (AEC)

  13. Thermally exfoliated graphite oxide

    NASA Technical Reports Server (NTRS)

    Prud'Homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor); Abdala, Ahmed (Inventor)

    2011-01-01

    A modified graphite oxide material contains a thermally exfoliated graphite oxide with a surface area of from about 300 sq m/g to 2600 sq m/g, wherein the thermally exfoliated graphite oxide displays no signature of the original graphite and/or graphite oxide, as determined by X-ray diffraction.

  14. Designing a TAC thermometer from a VHTR graphite structure

    SciTech Connect

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

    2015-03-31

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

  15. Designing a TAC thermometer from a VHTR graphite structure

    NASA Astrophysics Data System (ADS)

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

    2015-03-01

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

  16. The physics and phenomenology of One Atmosphere Uniform Glow Discharge Plasma (OAUGDP™) reactors for surface treatment applications

    NASA Astrophysics Data System (ADS)

    Reece Roth, J.; Rahel, Jozef; Dai, Xin; Sherman, Daniel M.

    2005-02-01

    In this paper, we present data on the physics and phenomenology of plasma reactors based on the One Atmosphere Uniform Glow Discharge Plasma (OAUGDP™) that are useful in optimizing the conditions for plasma formation, uniformity and surface treatment applications. It is shown that the real (as opposed to reactive) power delivered to a reactor is divided between dielectric heating of the insulating material and power delivered to the plasma available for ionization and active species production. A relationship is given for the dielectric heating power input as a function of the frequency and voltage at which the OAUGDP™ discharge is operated.

  17. Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel

    SciTech Connect

    Rearden, B.T.; Anderson, W.J.; Harms, G.A.

    2005-08-15

    Framatome ANP, Sandia National Laboratories (SNL), Oak Ridge National Laboratory (ORNL), and the University of Florida are cooperating on the U.S. Department of Energy Nuclear Energy Research Initiative (NERI) project 2001-0124 to design, assemble, execute, analyze, and document a series of critical experiments to validate reactor physics and criticality safety codes for the analysis of commercial power reactor fuels consisting of UO{sub 2} with {sup 235}U enrichments {>=}5 wt%. The experiments will be conducted at the SNL Pulsed Reactor Facility.Framatome ANP and SNL produced two series of conceptual experiment designs based on typical parameters, such as fuel-to-moderator ratios, that meet the programmatic requirements of this project within the given restraints on available materials and facilities. ORNL used the Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) to assess, from a detailed physics-based perspective, the similarity of the experiment designs to the commercial systems they are intended to validate. Based on the results of the TSUNAMI analysis, one series of experiments was found to be preferable to the other and will provide significant new data for the validation of reactor physics and criticality safety codes.

  18. Landslides as weathering reactors; links between physical erosion and weathering in rapidly eroding mountain belts

    NASA Astrophysics Data System (ADS)

    Emberson, R.; Hovius, N.; Galy, A.

    2014-12-01

    The link between physical erosion and chemical weathering is generally modelled with a surface-blanketing weathering zone, where the supply of fresh minerals is tied to the average rate of denudation. In very fast eroding environments, however, sediment production is dominated by landsliding, which acts in a stochastic fashion across the landscape, contrasting strongly with more uniform denudation models. If physical erosion is a driver of weathering at the highest erosion rates, then an alternative weathering model is required. Here we show that landslides can be effective 'weathering reactors'. Previous work modelling the effect of landslides on chemical weathering (Gabet 2007) considered the fresh bedrock surfaces exposed in landslide scars. However, fracturing during the landslide motion generates fresh surfaces, the total surface area of which exceeds that of the exposed scar by many orders of magnitude. Moreover, landslides introduce concavity into hillslopes, which acts to catch precipitation. This is funnelled into a deposit of highly fragmented rock mass with large reactive surface area and limited hydraulic conductivity (Lo et al. 2007). This allows percolating water reaction time for chemical weathering; any admixture of macerated organic debris could yield organic acid to further accelerate weathering. In the South island of New Zealand, seepage from recent landslide deposits has systematically high solute concentrations, far outstripping concentration in runoff from locations where soils are present. River total dissolved load in the western Southern Alps is highly correlated with the rate of recent (<35yrs) landsliding, suggesting that landslides are the dominant locus of weathering in this rapidly eroding landscape. A tight link between landsliding and weathering implies that localized weathering migrates through the landscape with physical erosion; this contrasts with persistent and ubiquitous weathering associated with soil production. Solute

  19. Expert systems for the analysis of transients on nuclear reactors: SEXTANT, a general-purpose physical analyzer

    SciTech Connect

    Barbet, N.; Dumas, M.; Mihelich, G.; Souchet, Y.; Thomas, J.B.

    1988-12-01

    Two expert systems for on-line analysis of nuclear reactor transients are reported. During a hypothetical crisis in a nuclear facility, a team of the Institute for Protection and Nuclear Safety must assess the risk to the local population. Expert systems are intended to assist in this analysis. The first deals with the availability of the safety systems of the plant (e.g., emergency core cooling system), depending on the functional state of the support systems. A second expert system will be built to study the physical transient of the reactor (mass and energy balance, pressure, flows). To do this, as in the development of the other expert systems, a physical analyzer is required. This is the aim of SEXTANT, which combines several knowledge bases concerning measurements, models, and qualitative behavior of the plant with a conjecture-refutation mechanism and a set of simplified models of the current physical state. A prototype is being assessed with integral test facility transients.

  20. Physics-Based Multi-State Models of Passive Component Degradation for the R7 Reactor Simulation Environment

    SciTech Connect

    Unwin, Stephen D.; Layton, Robert F.; Johnson, Kenneth I.; Lowry, Peter P.

    2012-06-25

    Abstract: The Next Generation Systems Analysis Code - referred to as R7 - is reactor systems simulation software being developed to support the Risk-Informed Safety Margin Characterization Pathway of the U.S. Department of Energy's Light Water Reactor Sustainability Program. It will provide an integrated multi-physics environment, implemented in an uncertainty quantification (UQ) framework that can produce risk and other performance insights on long-term reactor operations. An element of this simulation environment will be the performance of passive components and materials. Conventional models of component reliability are largely parametric, relying on plant service data to estimate component lifetimes and failure rates. This type of model has limited usefulness in the R7 environment where the intent is to explicitly determine the influence of physical stressors on component degradation. In this paper, we describe a new class of multi-state physics-based component models designed to be R7-compatible. These models capture the physics of materials degradation while also incorporating the effects of interventions and component rejuvenation. The models are implemented in a cumulative damage framework that allows the impact of an evolving physical environment to be addressed without recourse to resampling within the Monte Carlo-based UQ framework. The paper describes an application to stress corrosion cracking in dissimilar metal welds - a principal contributor to potential loss of coolant accidents. So while R7 will have the more conventional capability of reactor simulation codes to model the impact of degraded components and systems on plant performance, the methodology described here allows R7 to model the inverse effect; the impact of the physical environment on component degradation and performance.

  1. Bridged graphite oxide materials

    NASA Technical Reports Server (NTRS)

    Herrera-Alonso, Margarita (Inventor); McAllister, Michael J. (Inventor); Aksay, Ilhan A. (Inventor); Prud'homme, Robert K. (Inventor)

    2010-01-01

    Bridged graphite oxide material comprising graphite sheets bridged by at least one diamine bridging group. The bridged graphite oxide material may be incorporated in polymer composites or used in adsorption media.

  2. Nuclear graphite

    SciTech Connect

    Mercuri, R. A.; Criscione, J. M.

    1985-07-02

    A high strength, high coefficient of thermal expansion fine-grained isotropic graphite article produced from 30% to 70% of attritor milled gilsonite coke or other high CTE carbon filler particles and minor amounts of a binder such a coal tar pitch and petroleum pitch, the article being formed by warm isostatic molding at a temperature of between 50/sup 0/ C. and 70/sup 0/ C. under a pressure between 100 and 1000 psi for a time between 1 and 10 minutes. The particle size of the fillers ranges up to 150 microns.

  3. FUEL ASSAY REACTOR

    DOEpatents

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

    1962-12-25

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

  4. The use of active learning strategies in the instruction of Reactor Physics concepts

    SciTech Connect

    Robinson, Michael A.

    2000-01-01

    Each of the Active Learning strategies employed to teach Reactor Physics material has been or promises to be instructionally successful. The Cooperative Group strategy has demonstrated a statistically significant increase in student performance on the unit exam in teaching conceptually difficult, transport and diffusion theory material. However, this result was achieved at the expense of a modest increase in class time. The Tutorial CBI programs have enabled learning equally as well as classroom lectures without the direct intervention of an instructor. Thus, the Tutorials have been successful as homework assignments, releasing classroom time for other instruction. However, the time required for development of these tools was large, on the order of two hundred hours per hour of instruction. The initial introduction of the Case-Based strategy was roughly as effective as the traditional classroom instruction. Case-Based learning could well, after important modifications, perform better than traditional instruction. A larger percentage of the students prefer active learning strategies than prefer traditional lecture presentations. Student preferences for the active strategies were particularly strong when they believed that the strategies helped them learn the material better than they would have by using a lecture format. In some cases, students also preferred the active strategies because they were different from traditional instruction, a change of pace. Some students preferred lectures to CBI instruction, primarily because the CBI did not afford them the opportunity to question the instructor during the presentation.

  5. Reactor Physics Measurements and Benchmark Specifications for Oak Ridge Highly Enriched Uranium Sphere (ORSphere)

    DOE PAGES

    Marshall, Margaret A.

    2014-11-04

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with themore » GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.« less

  6. Reactor Physics Measurements and Benchmark Specifications for Oak Ridge Highly Enriched Uranium Sphere (ORSphere)

    SciTech Connect

    Marshall, Margaret A.

    2014-11-04

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.

  7. Test tasks for verification of program codes for calculation of neutron-physical characteristics of the BN series reactors

    NASA Astrophysics Data System (ADS)

    Tikhomirov, Georgy; Ternovikh, Mikhail; Smirnov, Anton; Saldikov, Ivan; Bahdanovich, Rynat; Gerasimov, Alexander

    2017-09-01

    System of test tasks is presented with the fast reactor BN-1200 with nitride fuel as prototype. The system of test tasks includes three test based on different geometric models. Model of fuel element in homogeneous and in heterogeneous form, model of fuel assembly in height-heterogeneous and full heterogeneous form, and modeling of the active core of BN-1200 reactor. Cross-verification of program codes was performed. Transition from simple geometry to more complex one allows to identify the causes of discrepancies in the results during the early stage of cross-verification of codes. This system of tests can be applied for certification of engineering programs based on the method of Monte Carlo to the calculation of full-scale models of the reactor core of the BN series. The developed tasks take into account the basic layout and structural features of the reactor BN-1200. They are intended for study of neutron-physical characteristics, estimation of influence of heterogeneous structure and influence of diffusion approximation. The development of system of test tasks allowed to perform independent testing of programs for calculation of neutron-physical characteristics: engineering programs JARFR and TRIGEX, and codes MCU, TDMCC, and MMK based on the method of Monte Carlo.

  8. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

    SciTech Connect

    Moses, David Lewis

    2011-10-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be deployed

  9. MHTGR [modular high-temperature gas-cooled reactor] core physics validation plan

    SciTech Connect

    Baxter, A.; Hackney, R.

    1988-01-01

    This document contains the verification and validation (V&V) plan for analytical methods utilized in the nuclear design for normal and off-normal conditions within the Modular High-Temperature Gas-Cooled Reactor (MHTGR). Regulations, regulatory guides, and industry standards have been reviewed and the approach for V&V has been developed. MHTGR core physics methods are described and the status of previous V&V is summarized within this document. Additional work required to verify and validate these methods is identified. The additional validation work includes comparison of calculations with available experimental data, benchmark comparison of calculations with available experimental data, benchmark comparisons with other validated codes, results from a cooperative program now underway at the Arbeitsgemeinschaft Versuchs-Reaktor GmbH (AVR) facility in Germany, results from a planned series of experiments on the Compact Nuclear Power Source (CNPS) facility at Los Alamos, and detailed documentation of all V&V studies. In addition, information will be obtained from planned international cooperative agreements to provide supplemental data for V&V. The regulatory technology development plan will be revised to include these additional experiments. A work schedule and cost estimate for completing this plan is also provided. This work schedule indicates the timeframe in which major milestones must be performed in order to complete V&V tasks prior to the issuance of preliminary design approval from the NRC. The cost to complete V&V tasks for core physics computational methods is estimated to be $2.2M. 41 refs., 13 figs., 8 tabs.

  10. Eco-friendly exfoliation of graphite into pristine graphene with little defect by a facile physical treatment

    NASA Astrophysics Data System (ADS)

    Chen, Jianping; Shi, Weili; Chen, Yongmei; Yang, Quanling; Wang, Mengkui; Liu, Bin; Tang, Zhen; Jiang, Ming; Fang, De; Xiong, Chuanxi

    2016-02-01

    The superior properties of graphene in applications ranging from electronic devices to composites have been extensively reported. So far, no mass production of defect-free few-layer graphene has been attained. The authors of this study have demonstrated a high-yield method to produce defect-free few-layer graphene by exfoliation of graphite in a degradable water-soluble polymer (I) with cholamine modification, and the obtained intercalated (D-I) chemical structure was confirmed by Fourier transform infrared spectroscopy. The electron donor forms π-π stacking interactions with the graphene sheets during sonication, which prevents the exfoliated graphene from restacking. The method is environment-friendly compared with other liquid exfoliation methods, and the aqueous and ethanolic solutions of graphene are stable for long durations. The authors also confirmed the presence of gossamer graphene sheets, which have typical wrinkled and folded structures, by using high resolution transmission electron microscopy. Atomic force microscopy images revealed that graphene sheets with a thickness of approximately 1 nm were uniformly distributed.

  11. 78 FR 29519 - Physical Protection of Irradiated Reactor Fuel in Transit

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-20

    ...The U.S. Nuclear Regulatory Commission (NRC) is amending its security regulations for the transport of irradiated reactor fuel (the terms ``irradiated reactor fuel'' and ``spent nuclear fuel'' are used interchangeably in this rule). This rulemaking establishes generically applicable security requirements similar to the requirements currently imposed by NRC Order EA-02-109, ``Issuance of Order......

  12. A Special Topic From Nuclear Reactor Dynamics for the Undergraduate Physics Curriculum

    ERIC Educational Resources Information Center

    Sevenich, R. A.

    1977-01-01

    Presents an intuitive derivation of the point reactor equations followed by formulation of equations for inverse and direct kinetics which are readily programmed on a digital computer. Suggests several computer simulations involving the effect of control rod motion on reactor power. (MLH)

  13. A Special Topic From Nuclear Reactor Dynamics for the Undergraduate Physics Curriculum

    ERIC Educational Resources Information Center

    Sevenich, R. A.

    1977-01-01

    Presents an intuitive derivation of the point reactor equations followed by formulation of equations for inverse and direct kinetics which are readily programmed on a digital computer. Suggests several computer simulations involving the effect of control rod motion on reactor power. (MLH)

  14. Variational techniques for reactor physics calculations of heterogeneous reactor cores. Final report for April 15, 1993--April 14, 1995

    SciTech Connect

    Wojtowicz, G.M.; Holloway, J.P.

    1995-06-01

    Variational coarse mesh techniques are developed for the solution of the one group neutron transport equation in one-dimensional reactor lattices. In contrast to conventional nodal lattice applications which discretize diffusion theory and use node homogenized cross sections, the authors retain the spatial dependence of the cross sections and instead employ an alternative flux representation. The initial form of this flux representation (trial function) for the angular flux was inspired by the leading order solution in the asymptotic expansion of the angular flux--namely, the slow modulation of a periodic pin cell flux. The authors called the variational technique based on this form of trial function the Spectral Element Asymptotic Method (SEAM); it is capable of achieving order of magnitude reductions of eigenvalue and pointwise scalar flux errors as compared with diffusion theory. A different trial function can be developed based on the leading order and first order correction terms in the asymptotic expansion of the angular flux. SuperSEAM, the method based on this new trial function, allows the neutron transport equation to be cast into a form whose solution has much slower spatial variation than the SEAM solution; thus, the SuperSEAM result can be accurately described with fewer variables. SuperSEAM is therefore capable of achieving the same high degree of accuracy as SEAM at a cost comparable with homogenized nodal diffusion theory.

  15. Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

    NASA Astrophysics Data System (ADS)

    Krishna, R.; Jones, A. N.; McDermott, L.; Marsden, B. J.

    2015-12-01

    Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated 'D'peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of 'G' and 'D' in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure.

  16. Physical modelling of the composting environment: A review. Part 1: Reactor systems

    SciTech Connect

    Mason, I.G. . E-mail: ian.mason@canterbury.ac.nz; Milke, M.W.

    2005-07-01

    In this paper, laboratory- and pilot-scale reactors used for investigation of the composting process are described and their characteristics and application reviewed. Reactor types were categorised by the present authors as fixed-temperature, self-heating, controlled temperature difference and controlled heat flux, depending upon the means of management of heat flux through vessel walls. The review indicated that fixed-temperature reactors have significant applications in studying reaction rates and other phenomena, but may self-heat to higher temperatures during the process. Self-heating laboratory-scale reactors, although inexpensive and uncomplicated, were shown to typically suffer from disproportionately large losses through the walls, even with substantial insulation present. At pilot scale, however, even moderately insulated self-heating reactors are able to reproduce wall losses similar to those reported for full-scale systems, and a simple technique for estimation of insulation requirements for self-heating reactors is presented. In contrast, controlled temperature difference and controlled heat flux laboratory reactors can provide spatial temperature differentials similar to those in full-scale systems, and can simulate full-scale wall losses. Surface area to volume ratios, a significant factor in terms of heat loss through vessel walls, were estimated by the present authors at 5.0-88.0 m{sup 2}/m{sup 3} for experimental composting reactors and 0.4-3.8 m{sup 2}/m{sup 3} for full-scale systems. Non-thermodynamic factors such as compression, sidewall airflow effects, channelling and mixing may affect simulation performance and are discussed. Further work to investigate wall effects in composting reactors, to obtain more data on horizontal temperature profiles and rates of biological heat production, to incorporate compressive effects into experimental reactors and to investigate experimental systems employing natural ventilation is suggested.

  17. Graphite oxidation and damage under irradiation at high temperatures in an impure helium environment

    NASA Astrophysics Data System (ADS)

    Goodwin, Cameron S.

    The High Temperature Gas-Cooled Reactor (HTGR) is a Generation IV reactor concept that uses a graphite-moderated nuclear reactor with a once-through uranium fuel cycle. In order to investigate the mechanism for corrosion of graphite in HTGRs, the graphite was placed in a similar environment in order to evaluate its resistance to corrosion and oxidation. While the effects of radiation on graphite have been studied in the past, the properties of graphite are largely dependent on the coke used in manufacturing the graphite. There are no longer any of the previously studied graphite types available for use in the HTGR. There are various types of graphite being considered for different uses in the HTGR and all of these graphite types need to be analyzed to determine how radiation will affect them. Extensive characterization of samples of five different types of graphite was conducted. The irradiated samples were analyzed with electron paramagnetic resonance spectroscopy, Raman spectroscopy, x-ray diffraction, x-ray photoelectron spectroscopy and gas chromatography. The results prove a knowledge base for considering the graphite types best suited for use in HTGRs. In my dissertation work graphite samples were gamma irradiated and also irradiated in a mixed field, in order to study the effects of neutron as well as gamma irradiation. Thermal effects on the graphite were also investigated by irradiating the samples at room temperature and at 1000 °C. From the analysi of the samples in this study there is no evidence of substantial damage to the grades of graphite analyzed. This is significant in approving the use of these graphites in nuclear reactors. Should significant damage had occurred to the samples, the use of these grades of graphite would need to be reconsidered. This information can be used to further characterize other grades of nuclear graphite as they become available.

  18. Nondestructive evaluation of nuclear-grade graphite

    NASA Astrophysics Data System (ADS)

    Kunerth, D. C.; McJunkin, T. R.

    2012-05-01

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

  19. Nondestructive evaluation of nuclear-grade graphite

    SciTech Connect

    Kunerth, D. C.; McJunkin, T. R.

    2012-05-17

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

  20. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    SciTech Connect

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  1. Preparation of graphitic articles

    DOEpatents

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

    2010-05-11

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

  2. AGC-2 Graphite Preirradiation Data Analysis Report

    SciTech Connect

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

    2013-08-01

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

  3. Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

    SciTech Connect

    Pope, Michael A.

    2014-10-01

    These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU. Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.

  4. Survey of Dust Production in Pebble Bed Reactors Cores

    SciTech Connect

    Joshua J. Cogliati; Abderafi M. Ougouag; Javier Ortensi

    2011-06-01

    Graphite dust produced via mechanical wear from the pebbles in a pebble bed reactor is an area of concern for licensing. Both the German pebble bed reactors produced graphite dust that contained activated elements. These activation products constitute an additional source term of radiation and must be taken under consideration during the conduct of accident analysis of the design. This paper discusses the available literature on graphite dust production and measurements in pebble bed reactors. Limited data is available on the graphite dust produced from the AVR and THTR-300 pebble bed reactors. Experiments that have been performed on wear of graphite in pebble-bed-like conditions are reviewed. The calculation of contact forces, which are a key driving mechanism for dust in the reactor, are also included. In addition, prior graphite dust predictions are examined, and future areas of research are identified.

  5. Graphite Isotope Ratio Method Development Report: Irradiation Test Demonstration of Uranium as a Low Fluence Indicator

    SciTech Connect

    Reid, B.D.; Gerlach, D.C.; Love, E.F.; McNeece, J.P.; Livingston, J.V.; Greenwood, L.R.; Petersen, S.L.; Morgan, W.C.

    1999-10-20

    This report describes an irradiation test designed to investigate the suitability of uranium as a graphite isotope ratio method (GIRM) low fluence indicator. GIRM is a demonstrated concept that gives a graphite-moderated reactor's lifetime production based on measuring changes in the isotopic ratio of elements known to exist in trace quantities within reactor-grade graphite. Appendix I of this report provides a tutorial on the GIRM concept.

  6. On monitoring and forecasting of graphite stack temperature in transient modes

    NASA Astrophysics Data System (ADS)

    Zagrebaev, A. M.; Ovsyannikova, N. V.; Ramazanov, R. N.

    2017-01-01

    The paper presents a method of monitoring and forecasting of the graphite stack temperature of the RBMK reactor in transient modes. The method is based on processing the in-core information about macro-distribution and mathematical model of distribution of temperature changes of the graphite stack in the reactor core. It is shown that the use of archival neutron field monitoring data allows determining the graphite stack temperature in the on-line mode.

  7. Data Report on the Newest Batch of PCEA Graphite for the VHTR Baseline Graphite Characterization Program

    SciTech Connect

    Carroll, Mark Christopher; Cottle, David Lynn; Rohrbaugh, David Thomas

    2016-08-01

    This report details a comparison of mechanical and physical properties from the first billet of extruded PCEA nuclear-grade graphite from the newest batch of PCEA procured from GrafTech. Testing has largely been completed on three of the billets from the original batch of PCEA, with data distributions for those billets exhibiting a much wider range of values when compared to the distributions of properties from other grades. A higher propensity for extremely low values or specimens that broke while machining or handling was also characteristic of the billets from the first batch, owing to unusually large fissures or disparate flaws in the billets in an as-manufactured state. Coordination with GrafTech prior to placing the order for a second batch of PCEA included discussions on these large disparate flaws and how to prevent them during the manufacturing process. This report provides a comparison of the observed data distributions from properties measured in the first billet from the new batch of PCEA with those observed in the original batch, in order that an evaluation of tighter control of the manufacturing process and the outcome of these controls on final properties can be ascertained. Additionally, this billet of PCEA is the first billet to formally include measurements from two alternate test techniques that will become part of the Baseline Graphite Characterization database – the three-point bend test on sub-sized cylinders and the Brazilian disc splitting tensile strength test. As the program moves forward, property distributions from these two tests will be based on specimen geometries that match specimen geometries being used in the irradiated Advanced Graphite Creep (AGC) program. This will allow a more thorough evaluation of both the utility of the test and expected variability in properties when using those approaches on the constrained geometries of specimens irradiated in the Advanced Test Reactor as part of the AGC experiment.

  8. Recent Developments of JAEA's Monte Carlo Code MVP for Reactor Physics Applications

    NASA Astrophysics Data System (ADS)

    Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa

    2014-06-01

    This paper describes the recent development status of a Monte Carlo code MVP developed at Japan Atomic Energy Agency. The basic features and capabilities of MVP are overviewed. In addition, new capabilities useful for reactor analysis are also described.

  9. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    SciTech Connect

    Powers, Jeffrey J.; George, Nathan; Maldonado, G. Ivan; Worrall, Andrew

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

  10. Fission Product Sorptivity in Graphite

    SciTech Connect

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

    2015-04-01

    Both adsorption and absorption (sorption) of fission product (FP) gases on/into graphite are issues of interest in very high temperature reactors (VHTRs). In the original proposal, we proposed to use packed beds of graphite particles to measure sorption at a variety of temperatures and to use an electrodynamic balance (EDB) to measure sorption onto single graphite particles (a few μm in diameter) at room temperature. The use of packed beds at elevated temperature is not an issue. However, the TPOC requested revision of this initial proposal to included single particle measurements at elevated temperatures up to 1100 °C. To accommodate the desire of NEUP to extend the single particle EDB measurements to elevated temperatures it was necessary to significantly revise the plan and the budget. These revisions were approved. In the EDB method, we levitate a single graphite particle (the size, surface characteristics, morphology, purity, and composition of the particle can be varied) or agglomerate in the balance and measure the sorption of species by observing the changes in mass. This process involves the use of an electron stepping technique to measure the total charge on a particle which, in conjunction with the measured suspension voltages for the particle, allows for determinations of mass and, hence, of mass changes which then correspond to measurements of sorption. Accommodating elevated temperatures with this type of system required a significant system redesign and required additional time that ultimately was not available. These constraints also meant that the grant had to focus on fewer species as a result. Overall, the extension of the original proposed single particle work to elevated temperatures added greatly to the complexity of the proposed project and added greatly to the time that would eventually be required as well. This means that the bulk of the experimental progress was made using the packed bed sorption systems. Only being able to recruit one

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

    SciTech Connect

    Margaret A. Marshall; John D. Bess

    2013-02-01

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

  12. On estimating the fracture probability of nuclear graphite components

    NASA Astrophysics Data System (ADS)

    Srinivasan, Makuteswara

    2008-10-01

    The properties of nuclear grade graphites exhibit anisotropy and could vary considerably within a manufactured block. Graphite strength is affected by the direction of alignment of the constituent coke particles, which is dictated by the forming method, coke particle size, and the size, shape, and orientation distribution of pores in the structure. In this paper, a Weibull failure probability analysis for components is presented using the American Society of Testing Materials strength specification for nuclear grade graphites for core components in advanced high-temperature gas-cooled reactors. The risk of rupture (probability of fracture) and survival probability (reliability) of large graphite blocks are calculated for varying and discrete values of service tensile stresses. The limitations in these calculations are discussed from considerations of actual reactor environmental conditions that could potentially degrade the specification properties because of damage due to complex interactions between irradiation, temperature, stress, and variability in reactor operation.

  13. Time-Dependent Behavior of a Graphite/Thermoplastic Composite and the Effects of Stress and Physical Aging

    NASA Technical Reports Server (NTRS)

    Gates, Thomas S.; Feldman, Mark

    1995-01-01

    Experimental studies were performed to determine the effects of stress and physical aging on the matrix dominated time dependent properties of IM7/8320 composite. Isothermal tensile creep/aging test techniques developed for polymers were adapted for testing of the composite material. Time dependent transverse and shear compliance's for an orthotropic plate were found from short term creep compliance measurements at constant, sub-T(8) temperatures. These compliance terms were shown to be affected by physical aging. Aging time shift factors and shift rates were found to be a function of temperature and applied stress.

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

    SciTech Connect

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

    2008-09-01

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

  15. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    SciTech Connect

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  16. Time dependent behavior of a graphite/thermoplastic composite and the effects of stress and physical aging

    NASA Technical Reports Server (NTRS)

    Gates, Thomas S.; Feldman, Mark

    1993-01-01

    Two complimentary studies were performed to determine the effects of stress and physical aging on the matrix dominated time dependent properties of IM7/8320 composite. The first of these studies, experimental in nature, used isothermal tensile creep/aging test techniques developed for polymers and adapted them for testing of the composite material. From these tests, the time dependent transverse (S22) and shear (S66) compliance's for an orthotropic plate were found from short term creep compliance measurements at constant, sub-T(sub g) temperatures. These compliance terms were shown to be affected by physical aging. Aging time shift factors and shift rates were found to be a function of temperature and applied stress. The second part of the study relied upon isothermal uniaxial tension tests of IM7/8320 to determine the effects of physical aging on the nonlinear material behavior at elevated temperature. An elastic/viscoplastic constitutive model was used to quantify the effects of aging on the rate-independent plastic and rate-dependent viscoplastic response. Sensitivity of the material constants required by the model to aging time were determined for aging times up to 65 hours. Verification of the analytical model indicated that the effects of prior aging on the nonlinear stress/strain/time data of matrix dominated laminates can be predicted.

  17. Electrical Properties and Physical Characteristics of Polycrystalline Diamond Films Deposited in a Microwave Plasma Disk Reactor

    NASA Astrophysics Data System (ADS)

    Huang, Bohr-Ran

    1992-01-01

    This work experimentally investigates techniques for high quality diamond synthesis and develops means for electrical and physical characterization of the films. The films are deposited by plasma assisted chemical vapor deposition using a methane/hydrogen plasma in a microwave plasma disk reactor system. Both a diamond past nucleation method and a diamond powder nucleation method are studied in this research. Although as indicated by Raman spectroscopy both methods produced similar quality diamond films, the powder nucleation method produced fine grain, sub-micron sized crystallite, films whereas the past nucleation method produced large grain, several-micrometer size crystallite, films. For powder polished films, all metallic contacts were ohmic. These samples were used to explore the high electric field properties of diamond. It was discovered that for fields larger than approximately 1 times 10^5 V/cm the electrical properties are dominated by defects, where defect is used generically for either an impurity or a structural defect. For low electric fields, the electrical conductivity was constant which resulted in ohmic behavior. But for high fields, the conductivity was field activated according to Poole's law. This behavior was modeled as being due to ionizable defects and indicates that there is approximately one ionizable defect per 10,000 host atoms. As a result of such defects, the breakdown field for these films was somewhat less than 1 times 10^6 V/cm. A large concentration of defects is compatible with the observation of ohmic contact behavior regardless of metallic work function since contact space charge layers would be sufficiently thin to allow tunneling. Non-ohmic, Schottky barrier contacts were achievable on the past polished films. For Al/diamond/silicon structures diode characteristics were observed. These I-V characteristics were modeled as an ideal Schottky barrier diode in series with bulk diamond, for which the property of the bulk diamond

  18. FERRET-SAND II physics-dosimetry analysis for N Reactor Pressure Tubes 2954, 3053 and 1165 using a WIMS calculated input spectrum

    SciTech Connect

    McElroy, W.N.; Kellogg, L.S.; Matsumoto, W.Y.; Morgan, W.C.; Suski, A.E.

    1988-05-01

    This report is in response to a request from Westinghouse Hanford Company (WHC) that the PNL National Dosimetry Center (NDC) perform physics-dosimetry analyses (E > MeV) for N Reactor Pressure Tubes 2954 and 3053. As a result of these analyses, and recommendations for additional studies, two physics-dosimetry re-evaluations for Pressure Tube 1165 were also accomplished. The primary objective of Pacific Northwest Laboratories' (PNL) National Dosimetry Center (NDC) physics-dosimetry work for N Reactor was to provide FERRET-SAND II physics-dosimetry results to assist in the assessment of neutron radiation-induced changes in the physical and mechanical properties of N Reactor pressure tubes. 15 refs., 6 figs., 5 tabs.

  19. The irradiation dimensional changes of grade TSX graphite

    SciTech Connect

    Kennedy, C R; Woodruff, E M

    1988-01-01

    Grade TSX graphite is used as a moderator in the N Reactor which has operated since 1963. This reactor, designed for a 25-year life, is under study to determine the possibility of significantly extending the operating life. One limiting factor is dimensional growth of the graphite lattice making up the core of the reactor. Since the original demands (25-year life) were modest, the dimensional change behavior was derived from a compendium of irradiation data from other grades and only confirmed by a few low-exposure irradiation experiments. Therefore, to generate actual dimensional change data for grade TSX to exposures relevant to the life extension plans, a series of irradiations of TSX graphite were run in the High Flux Isotope Reactor (HFIR) at Oak Ridge. This report contains experimental results of such testing. 5 refs., 3 figs.

  20. LWR pressure vessel surveillance dosimetry improvement program: LWR power reactor surveillance physics-dosimetry data base compendium

    SciTech Connect

    McElroy, W.N.

    1985-08-01

    This NRC physics-dosimetry compendium is a collation of information and data developed from available research and commercial light water reactor vessel surveillance program (RVSP) documents and related surveillance capsule reports. The data represents the results of the HEDL least-squares FERRET-SAND II Code re-evaluation of exposure units and values for 47 PWR and BWR surveillance capsules for W, B and W, CE, and GE power plants. Using a consistent set of auxiliary data and dosimetry-adjusted reactor physics results, the revised fluence values for E > 1 MeV averaged 25% higher than the originally reported values. The range of fluence values (new/old) was from a low of 0.80 to a high of 2.38. These HEDL-derived FERRET-SAND II exposure parameter values are being used for NRC-supported HEDL and other PWR and BWR trend curve data development and testing studies. These studies are providing results to support Revision 2 of Regulatory Guide 1.99. As stated by Randall (Ra84), the Guide is being updated to reflect recent studies of the physical basis for neutron radiation damage and efforts to correlate damage to chemical composition and fluence.

  1. Physics-based multiscale coupling for full core nuclear reactor simulation

    SciTech Connect

    Gaston, Derek R.; Permann, Cody J.; Peterson, John W.; Slaughter, Andrew E.; Andrš, David; Wang, Yaqi; Short, Michael P.; Perez, Danielle M.; Tonks, Michael R.; Ortensi, Javier; Zou, Ling; Martineau, Richard C.

    2015-10-01

    Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different data exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle. 2014 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-SA license

  2. Physics-based multiscale coupling for full core nuclear reactor simulation

    DOE PAGES

    Gaston, Derek R.; Permann, Cody J.; Peterson, John W.; ...

    2015-10-01

    Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different datamore » exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle. 2014 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-SA license« less

  3. 78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-30

    ..., ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,'' LWR Edition... Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1583 or email... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY...

  4. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    SciTech Connect

    David W. Nigg; Amanda E. Schwint; John K. Hartwell; Elisa M. Heber; Veronica Trivillin; Jorge Castillo; Luis Wentzeis; Patrick Sloan; Charles A. Wemple

    2004-10-01

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  5. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    SciTech Connect

    Nigg, D.W.; Schwint, A.E.; Hartwell, J.K.; Heber, E.M.; Trivillin, V.; Castillo, J.; Wentzeis, L.; Sloan, P.; Wemple, C.A.

    2004-10-04

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  6. Reactor Physics Parametric and Depletion Studies in Support of TRISO Particle Fuel Specification for the Next Generation Nuclear Plant

    SciTech Connect

    James W. Sterbentz; Bren Phillips; Robert L. Sant; Gray S. Chang; Paul D. Bayless

    2003-09-01

    Reactor physics calculations were initiated to answer several major questions related to the proposed TRISO-coated particle fuel that is to be used in the prismatic Very High Temperature Reactor (VHTR) or the Next Generation Nuclear Plant (NGNP). These preliminary design evaluation calculations help ensure that the upcoming fuel irradiation tests will test appropriate size and type of fuel particles for a future NGNP reactor design. Conclusions from these calculations are expected to confirm and suggest possible modifications to the current particle fuel parameters specified in the evolving Fuel Specification. Calculated results dispel the need for a binary fuel particle system, which is proposed in the General Atomics GT-MHR concept. The GT-MHR binary system is composed of both a fissile and fertile particle with 350- and 500- micron kernel diameters, respectively. For the NGNP reactor, a single fissile particle system (single UCO kernel size) can meet the reactivity and power cycle length requirements demanded of the NGNP. At the same time, it will provide substantial programmatic cost savings by eliminating the need for dual particle fabrication process lines and dual fuel particle irradiation tests required of a binary system. Use of a larger 425-micron kernel diameter single fissile particle (proposed here), as opposed to the 350-micron GT-MHR fissile particle size, helps alleviate current compact particle packing fractions fabrication limitations (<35%), improves fuel block loading for higher n-batch reload options, and tracks the historical correlation between particle size and enrichment (10 and 14 wt% U-235 particle enrichments are proposed for the NGNP). Overall, the use of the slightly larger kernel significantly broadens the NGNP reactor core design envelope and provides increased design margin to accommodate the (as yet) unknown final NGNP reactor design. Maximum power-peaking factors are calculated for both the initial and equilibrium NGNP cores

  7. Thermal Properties of G-348 Graphite

    SciTech Connect

    McEligot, Donald; Swank, W. David; Cottle, David L.; Valentin, Francisco I.

    2016-05-01

    Fundamental measurements have been obtained in the INL Graphite Characterization Laboratory to deduce the temperature dependence of thermal conductivity for G-348 isotropic graphite, which has been used by City College of New York in thermal experiments related to gas-cooled nuclear reactors. Measurements of thermal diffusivity, mass, volume and thermal expansion were converted to thermal conductivity in accordance with ASTM Standard Practice C781-08. Data are tabulated and a preliminary correlation for the thermal conductivity is presented as a function of temperature from laboratory temperature to 1000C.

  8. Neutron fluxes in test reactors

    SciTech Connect

    Youinou, Gilles Jean-Michel

    2017-01-01

    Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

  9. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    NASA Astrophysics Data System (ADS)

    Dautray, Robert

    2011-06-01

    The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation

  10. Status of Chronic Oxidation Studies of Graphite

    SciTech Connect

    Contescu, Cristian I.; Mee, Robert W.

    2016-05-01

    Graphite will undergo extremely slow, but continuous oxidation by traces of moisture that will be present, albeit at very low levels, in the helium coolant of HTGR. This chronic oxidation may cause degradation of mechanical strength and thermal properties of graphite components if a porous oxidation layer penetrates deep enough in the bulk of graphite components during the lifetime of the reactor. The current research on graphite chronic oxidation is motivated by the acute need to understand the behavior of each graphite grade during prolonged exposure to high temperature chemical attack by moisture. The goal is to provide the elements needed to develop predictive models for long-time oxidation behavior of graphite components in the cooling helium of HTGR. The tasks derived from this goal are: (1) Oxidation rate measurements in order to determine and validate a comprehensive kinetic model suitable for prediction of intrinsic oxidation rates as a function of temperature and oxidant gas composition; (2) Characterization of effective diffusivity of water vapor in the graphite pore system in order to account for the in-pore transport of moisture; and (3) Development and validation of a predictive model for the penetration depth of the oxidized layer, in order to assess the risk of oxidation caused damage of particular graphite grades after prolonged exposure to the environment of helium coolant in HTGR. The most important and most time consuming of these tasks is the measurement of oxidation rates in accelerated oxidation tests (but still under kinetic control) and the development of a reliable kinetic model. This report summarizes the status of chronic oxidation studies on graphite, and then focuses on model development activities, progress of kinetic measurements, validation of results, and improvement of the kinetic models. Analysis of current and past results obtained with three grades of showed that the classical Langmuir-Hinshelwood model cannot reproduce all

  11. Modeling the Multiaxial Strength of H-451 Nuclear Grade Graphite

    SciTech Connect

    Burchell, Timothy D; Yahr, Terry; Battiste, Rick

    2007-01-01

    The core of a prismatic High Temperature Reactor (HTR) is constructed from an array of nuclear graphite components. Similarly, the core of a Pebble Bed HTR is confined by large graphite blocks which define the (annular) core geometry. In both HTR designs the large graphite components act as neutron moderator and reflector as well as providing mechanical support to the active core. During reactor operation the graphite components of the core are subjected to complex stress states. Consequently, core designers need a suitable theory of failure. Both deterministic (e.g., maximum principal stress theory) and probabilistic (e.g., Weibull failure theory) have been considered. To test candidate failure theories a multiaxial testing program was conducted at Oak Ridge National Laboratory on H-451 graphite. Large specimens ({approx}27 cm length) were subjected to combined axial stress (tension and compression) and internal pressure. A total of 59 specimens were tested at 9 stress ratios in the first and fourth stress quadrants. Here, we report the basis and performance of a microstructurally based graphite fracture model and multiaxial strength data for grade H-451 graphite, along with the application of the model to predict the failure envelope for H-451 graphite in the first and fourth multiaxial quadrants.

  12. Approaches to Deal with Irradiated Graphite in Russia - Proposal for New IAEA CRP on Graphite Waste Management - 12364

    SciTech Connect

    Kascheev, Vladimir; Poluektov, Pavel; Ustinov, Oleg

    2012-07-01

    The problems of spent reactor graphite are being shown, the options of its disposal is considered. Burning method is selected as the most efficient and waste-free. It is made a comparison of amounts of {sup 14}C that entering the environment in a natural way during the operation of nuclear power plants (NPPs) and as a result of the proposed burning of spent reactor graphite. It is shown the possibility of burning graphite with the arrival of {sup 14}C into the atmosphere within the maximum allowable emissions. This paper analyzes the different ways of spent reactor graphite treatment. It is shown the possibility of its reprocessing by burning method in the air flow. It is estimated the effect of this technology to the overall radiation environment and compared its contribution to the general background radiation due to cosmic radiation and NPPs emission. It is estimated the maximum permissible speeds of burning reactor graphite (for example, RBMK graphite) for areas with different conditions of agricultural activities. (authors)

  13. A physical pulverization strategy for preparing a highly active composite of CoOx and crushed graphite for lithium-oxygen batteries.

    PubMed

    Ming, Jun; Kwak, Won-Jin; Park, Jin-Bum; Shin, Chang-Dae; Lu, Jun; Curtiss, Larry; Amine, Khalil; Sun, Yang-Kook

    2014-07-21

    A new physical pulverization strategy has been developed to prepare a highly active composite of CoOx and crushed graphite (CG) for the cathode in lithium-oxygen batteries. The effect of CoOx loading on the charge potential in the oxygen evolution reaction (Li(2)O(2) →2 Li(+) +O(2) +2e(-)) was investigated in coin-cell tests. The CoOx (38.9 wt %)/CG composite showed a low charge potential of 3.92 V with a delivered capacity of 2 mAh cm(-2) under a current density of 0.2 mA cm(-2). The charge potential was 4.10 and 4.15 V at a capacity of 5 and 10 mAh cm(-2), respectively, with a current density of 0.5 mA cm(-2). The stability of the electrolyte and discharge product on the gas-diffusion layer after the cycling were preliminarily characterized by (1)H nuclear magnetic resonance spectroscopy, scanning electron microscopy, X-ray photoelectron spectroscopy, and X-ray diffraction. The high activity of the composite was further analyzed by electrochemical impedance spectroscopy, cyclic voltammetry, and potential-step chronoamperometry. The results indicate that our near-dry milling method is an effective and green approach to preparing a nanocomposite cathode with high surface area and porosity, while using less solvent. Its relative simplicity compared with the traditional solution method could facilitate its widespread application in catalysis, energy storage, and materials science.

  14. The Application of Modern Nodal Methods to Pwr Reactor Physics Analysis.

    NASA Astrophysics Data System (ADS)

    Knight, M. P.

    Available from UMI in association with The British Library. The objective of this research is to develop efficient computational procedures for PWR reactor calculations, based on modern nodal methods. The analytic nodal method, which is characterised by the use of exact exponential expansions in transverse-integrated equations, is implemented within an existing finite-difference code. This shows considerable accuracy and efficiency on standard benchmark problems, very much in line with existing experience with nodal methods. Assembly powers can be calculated to within 2.0% with just one mesh per assembly. The recovery of fine detail from a nodal solution based on such a coarse mesh requires additional effort. Techniques are develolped in this thesis which allow the basic nodal equations to be used in this reconstruction, and therefore provide a consistent approach. Pin powers can be recovered from assembly-averaged values with little further loss of accuracy. A similar investigation is followed with the transverse leakage distribution. An improvement, which uses known local behaviour, is shown to be very effective in some limited applications, but overall provides little advantage over the much simpler quadratic model. For heterogeneous calculations it is essential that the homogenisation techniques are well matched to the nodal method. The asymmetric design of some assemblies provides a severe test. Techniques are devised that allow some overall representation of this asymmetry to be retained in the reactor calculation, even when using one mesh per assembly. Extensions of this procedure provide an almost exact global representation of a heterogeneous assembly. A complete comparison is performed between reactor calculations at one mesh per pin, and at one mesh per assembly using nodal and homogenisation methods. Homogenisation errors and nodal coarse-mesh errors are shown to be very similar, amounting to about 0.1% on reactor eigenvalue, 2.0% on assembly power and

  15. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    SciTech Connect

    Ortensi, Javier; Baker, Benjamin Allen; Schunert, Sebastian; Wang, Yaqi; Gleicher, Frederick Nathan; DeHart, Mark David

    2016-06-01

    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  16. Interface structure between tetraglyme and graphite

    NASA Astrophysics Data System (ADS)

    Minato, Taketoshi; Araki, Yuki; Umeda, Kenichi; Yamanaka, Toshiro; Okazaki, Ken-ichi; Onishi, Hiroshi; Abe, Takeshi; Ogumi, Zempachi

    2017-09-01

    Clarification of the details of the interface structure between liquids and solids is crucial for understanding the fundamental processes of physical functions. Herein, we investigate the structure of the interface between tetraglyme and graphite and propose a model for the interface structure based on the observation of frequency-modulation atomic force microscopy in liquids. The ordering and distorted adsorption of tetraglyme on graphite were observed. It is found that tetraglyme stably adsorbs on graphite. Density functional theory calculations supported the adsorption structure. In the liquid phase, there is a layered structure of the molecular distribution with an average distance of 0.60 nm between layers.

  17. Multi-step Monte Carlo calculations applied to nuclear reactor instrumentation - source definition and renormalization to physical values

    SciTech Connect

    Radulovic, Vladimir; Barbot, Loic; Fourmentel, Damien; Villard, Jean-Francois; Snoj, Luka; Zerovnik, Gasper; Trkov, Andrej

    2015-07-01

    Significant efforts have been made over the last few years in the French Alternative Energies and Atomic Energy Commission (CEA) to adopt multi-step Monte Carlo calculation schemes in the investigation and interpretation of the response of nuclear reactor instrumentation detectors (e.g. miniature ionization chambers - MICs and self-powered neutron or gamma detectors - SPNDs and SPGDs). The first step consists of the calculation of the primary data, i.e. evaluation of the neutron and gamma flux levels and spectra in the environment where the detector is located, using a computational model of the complete nuclear reactor core and its surroundings. These data are subsequently used to define sources for the following calculation steps, in which only a model of the detector under investigation is used. This approach enables calculations with satisfactory statistical uncertainties (of the order of a few %) within regions which are very small in size (the typical volume of which is of the order of 1 mm{sup 3}). The main drawback of a calculation scheme as described above is that perturbation effects on the radiation conditions caused by the detectors themselves are not taken into account. Depending on the detector, the nuclear reactor and the irradiation position, the perturbation in the neutron flux as primary data may reach 10 to 20%. A further issue is whether the model used in the second step calculations yields physically representative results. This is generally not the case, as significant deviations may arise, depending on the source definition. In particular, as presented in the paper, the injudicious use of special options aimed at increasing the computation efficiency (e.g. reflective boundary conditions) may introduce unphysical bias in the calculated flux levels and distortions in the spectral shapes. This paper presents examples of the issues described above related to a case study on the interpretation of the signal from different types of SPNDs, which

  18. Metal burning in graphite-moderated reactors

    SciTech Connect

    Wichner, R.P.; Ball, S.J.; Daw, C.S.; Thomas, J.F.

    1997-05-01

    Pinto beans, sweet corn, and zucchini squash (Cucurbita pepo var. black beauty) were grown in a randomized complete-block field/pot experiment at a site that contained the highest observed levels of surface gross gamma radioactivity within Los Alamos Canyon (LAC) at Los Alamos National Laboratory. Soils as well as washed edible and nonedible crop tissues were analyzed for various radionuclides and heavy metals. Most radionuclides, with the exception of {sup 3}H and {sup tot}U, in soil from LAC were detected in significantly higher concentrations (p <0.01) than in soil collected from regional background (RBG) locations. Similarly, most radionuclides in edible crop portions of beans, squash, and corn were detected in significantly higher (p <0.01 and 0.05) concentrations than RBG. Most soil-to-plant concentration ratios for radionuclides in edible and nonedible crop tissues from LAC were within the default values given by the Nuclear Regulatory Commission and Environmental Protection Agency. All heavy metals in soils, as well as edible and nonedible crop tissues grown in soils from LAC, were within RBG concentrations. Overall, the total maximum net positive committed effective dose equivalent (CEDE)--the CEDE plus two sigma for each radioisotope minus background and then all positive doses summed--to a hypothetical 50-year resident that ingested 160 kg of beans, corn, and squash in equal proportions, was 74 mrem y{sup -1}. This dose was below the International Commission on Radiological Protection permissible dose limit (PDL) of 100 mrem y{sup -1} from all pathways; however, the addition of other internal and external exposure route factors may increase the overall dose over the PDL. Also, the risk of an excess cancer fatality, based on 74 mrem y{sup -1}, was 3.7 x 10{sup -5} (37 in a million), which is above the Environmental Protection Agency`s (acceptable) guideline of one in a million. 25 refs.

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

    SciTech Connect

    Margaret A. Marshall; John D. Bess

    2012-11-01

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

  20. Role of nuclear grade graphite in controlling oxidation in modular HTGRs

    SciTech Connect

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

    2014-11-01

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

  1. After SNO and before KamLAND: present and future of Solar and Reactor Neutrino Physics

    NASA Astrophysics Data System (ADS)

    Aliani, P.; Antonelli, V.; Ferrari, R.; Picariello, M.; Torrente-Lujan, E.

    2003-02-01

    We present a short review of the existing evidence in favor of neutrino mass and neutrino oscillations which come from different kinds of experiments. We focus our attention in particular on solar neutrinos, presenting a review of some recent analysis of all available neutrino oscillation evidence in Solar experiments including the recent SNO CC and NC data. We present in detail the power of the reactor experiment KamLAND for discriminating existing solutions to the SNP and giving accurate information on neutrino masses and mixing angles.

  2. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    SciTech Connect

    1994-05-27

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

  3. Coupled reactor physics and coolant dynamics of heavy-liquid-metal-coolant systems

    SciTech Connect

    Cahalan, J.E.; Dunn, F.E.; Taiwo, T.A.

    1999-07-01

    Cooling of advanced nuclear designs with heavy liquid metals such as lead or lead-bismuth eutectic offers improved safety characteristics and higher operating efficiencies compared with previously considered liquid-metal coolants such as sodium or NaK. Such applications would, however, also introduce additional safety concerns and design challenges, therefore necessitating a verifiable computational tool for transient design-basis analysis of heavy-liquid-metal-coolant (HLMC) systems. This capability would enable analysts to compare operational and safety characteristics of design alternatives and to evaluate relative performance advantages with a consistent, deterministic measure. No existing computer code provides all the computational capabilities needed for system analysis of HLMC designs. However, the SASSYS-1 computer code, long used for analysis of sodium-cooled designs, has been adapted to provide a scoping reactor kinetics/thermal-hydraulics analysis capability and a framework for future development. Revision of the SASSYS-1 reactor kinetics, thermophysical properties, convective heat transfer, and flow momentum transfer models has permitted preliminary design and safety assessments of HLMC systems. Preliminary results are given here.

  4. Physics effects of accidental submersion of space power reactors in water

    NASA Astrophysics Data System (ADS)

    Bhattacharyya, S. K.; Lell, R. M.

    A study has been performed of the generic aspects of the accidental submersion phenomenon in fast spectrum space power reactors. Since the net effect of submersion is a balance of competing effects of neutron spectrum softening and neutron reflection, a systematic study using rigorous analytical methods is necessary to extract the trends. Monte Carlo calculations were used to determine essential flooding characteristics in realistic infinite lattice cells. For practical clad/structural materials, computed flooding reactivity worths in infinite lattices were found to be negative because parasitic absorption rises faster than neutron production over the neutron energy range of importance. The flooding worth depends strongly on cell design, becoming increasingly negative as the coolant volume fraction increases. Negative flooding worth was calculated for all fissile enrichments considered (24 to 93 percent). Transport theory analyses of idealized cores of realistic sizes showed that computed flooding worths are positive or slightly negative for small cores because reflection of thermalized neutrons by the water outside dominates other flooding effects. As core size increases, infinite lattice behavior in the interior dominates, and flooding worth becomes increasingly negative. The loss of control drums was found to increase the core radius for crossover to negative flooding worths. Monte Carlo and Sn calculations for realistic core designs confirmed the dependence of sign and magnitude of the flooding worth on core size and on the presence or absence of control drums. Experimental evidence from terrestrial reactor programs have demonstrated the need for critical experiments to benchmark and calibrate analytical procedures for flooding calculations.

  5. Development of an ASTM Graphite Oxidation Test Method

    SciTech Connect

    Contescu, Cristian I; Baker, Frederick S; Burchell, Timothy D

    2006-01-01

    Oxidation behavior of graphite is of practical interest because of extended use of graphite materials in nuclear reactors. High temperature gas-cooled reactors are expected to become the nuclear reactors of the next generation. The most critical factor in their safe operation is an air-ingress accident, in which case the graphite materials in the moderator and reflector would come in contact with oxygen at a high temperature. Many results on graphite oxidation have been obtained from TGA measurements using commercial instruments, with sample sizes of a few hundred milligrams. They have demonstrated that graphite oxidation is in kinetic control regime at low temperatures, but becomes diffusion-limited at high temperatures. These effects are better understood from measurement results with large size samples, on which the shape and structural factors that control diffusion can be more clearly evidenced. An ASTM test for characterization of oxidation resistance of machined carbon and graphite materials is being developed with ORNL participation. The test recommends the use of large machined samples (~ 20 grams) in a dry air flow system. We will report on recent results and progress in this direction.

  6. Formation mechanism of graphite hexagonal pyramids by argon plasma etching of graphite substrates

    NASA Astrophysics Data System (ADS)

    Glad, X.; de Poucques, L.; Bougdira, J.

    2015-12-01

    A new graphite crystal morphology has been recently reported, namely the graphite hexagonal pyramids (GHPs). They are hexagonally-shaped crystals with diameters ranging from 50 to 800 nm and a constant apex angle of 40°. These nanostructures are formed from graphite substrates (flexible graphite and highly ordered pyrolytic graphite) in low pressure helicon coupling radiofrequency argon plasma at 25 eV ion energy and, purportedly, due to a physical etching process. In this paper, the occurrence of peculiar crystals is shown, presenting two hexagonal orientations obtained on both types of samples, which confirms such a formation mechanism. Moreover, by applying a pretreatment step with different time durations of inductive coupling radiofrequency argon plasma, for which the incident ion energy decreases at 12 eV, uniform coverage of the surface can be achieved with an influence on the density and size of the GHPs.

  7. Synthesis of soluble graphite and graphene.

    PubMed

    Kelly, K F; Billups, W E

    2013-01-15

    Because of graphene's anticipated applications in electronics and its thermal, mechanical, and optical properties, many scientists and engineers are interested in this material. Graphene is an isolated layer of the π-stacked hexagonal allotrope of carbon known as graphite. The interlayer cohesive energy of graphite, or exfoliation energy, that results from van der Waals attractions over the interlayer spacing distance of 3.34 Å (61 meV/C atom) is many times weaker than the intralayer covalent bonding. Since graphene itself does not occur naturally, scientists and engineers are still learning how to isolate and manipulate individual layers of graphene. Some researchers have relied on the physical separation of the sheets, a process that can sometimes be as simple as peeling of sheets from crystalline graphite using Scotch tape. Other researchers have taken an ensemble approach, where they exploit the chemical conversion of graphite to the individual layers. The typical intermediary state is graphite oxide, which is often produced using strong oxidants under acidic conditions. Structurally, researchers hypothesize that acidic functional groups functionalize the oxidized material at the edges and a network of epoxy groups cover the sp(2)-bonded carbon network. The exfoliated material formed under these conditions can be used to form dispersions that are usually unstable. However, more importantly, irreversible defects form in the basal plane during oxidation and remain even after reduction of graphite oxide back to graphene-like material. As part of our interest in the dissolution of carbon nanomaterials, we have explored the derivatization of graphite following the same procedures that preserve the sp(2) bonding and the associated unique physical and electronic properties in the chemical processing of single-walled carbon nanotubes. In this Account, we describe efficient routes to exfoliate graphite either into graphitic nanoparticles or into graphene without

  8. Chernobyl lessons learned review of N Reactor

    SciTech Connect

    Weber, E.T.; McNeece, J.P.; Omberg, R.P.; Stepnewski, D.D.; Lutz, R.J.; Henry, R.E.; Bonser, K.D.; Miller, N.R.

    1987-10-01

    A broad-base review of the N Reactor plant, design characteristics, administrative controls and responses unique to upset conditions has been completed. The review was keyed to Nuclear Regulatory Commission (NRC)-defined issues associated with the Chernobyl accident. Physical features of N Reactor that preclude an accident like Chernobyl include: lack of autocatalytic reactivity insertion (i.e., negative coolant void and power coefficents) and two separate, fast-acting scram systems. Administrative controls in place at N Reactor would effectively protect against the operator errors and safety violations that set up the Chernobyl accident. Several items were identified where further near-term action is appropriate to ensure effectiveness of existing safety features: Resolve a question concerning the exact point at which Emergency Core Cooling System (ECCS) activation by manual actions should be implemented or deferred if automatic ECCS trip fails. Ensure appropriate revision of the Emergency Response Guides and full communication of the correct procedure to all Operations, Safety and cognizant Technology staff. Train reactor operators in the currently recognized significance of the Graphite and Shield Cooling System (GSCS) in severe accident situations and cover this appropriately in the Emergency Response Guides. Complete reviews which establish an independent verification that pressure tube rupture will not propagate to other tubes. 15 refs., 3 tabs.

  9. Producing graphite with desired properties

    NASA Technical Reports Server (NTRS)

    Dickinson, J. M.; Imprescia, R. J.; Reiswig, R. D.; Smith, M. C.

    1971-01-01

    Isotropic or anisotropic graphite is synthesized with precise control of particle size, distribution, and shape. The isotropic graphites are nearly perfectly isotropic, with thermal expansion coefficients two or three times those of ordinary graphites. The anisotropic graphites approach the anisotropy of pyrolytic graphite.

  10. Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)

    SciTech Connect

    Ingersoll, DT

    2005-12-15

    A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to coolant voiding. Three organizations participated in the study: Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Although the intent was to generate a new reference LS-VHTR core design, the emphasis was on performing parametric studies of the many variables that constitute a design. The results of the parametric studies not only provide the basis for choosing the optimum balance of design options, they also provide a valuable understanding of the fundamental behavior of the core, which will be the basis of future design trade-off studies. A new 2400-MW(t) baseline design was established that consists of a cylindrical, nonannular core cooled by liquid {sup 7}Li{sub 2}BeF{sub 4} (Flibe) salt. The inlet and outlet coolant temperatures were decreased by 50 C, and the coolant channel diameter was increased to help lower the maximum fuel and vessel temperatures. An 18-month fuel cycle length with 156 GWD/t burnup was achieved with a two-batch shuffling scheme, while maintaining a core power density of 10 MW/m{sup 3} using graphite-coated uranium oxicarbide particle fuel enriched to 15% {sup 235}U and assuming a 25 vol-% packing of the coated particles in the fuel compacts. The revised design appears to have excellent steady-state and transient performance. The previous concern regarding the core's response to coolant voiding has been resolved for the case of Flibe coolant by increasing the coolant channel diameter and the fuel loading. Also, the LSVHTR has a strong decay heat removal performance and appears capable of surviving a loss of forced circulation

  11. Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

    SciTech Connect

    Schneider, E.; Deinert, M.; Bingham Cady, K.

    2006-07-01

    Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)

  12. Relevance of β-delayed neutron data for reactor, nuclear physics and astrophysics applications

    NASA Astrophysics Data System (ADS)

    Kratz, Karl-Ludwig

    2015-02-01

    Initially, yields (or abundances) and branching ratios of β-delayed neutrons (βdn) from fission products (Pn-values) have had their main importance in nuclear reactor control. At that time, the six-group mathematical approximation of the time-dependence of βdn-data in terms of the so-called "Keepin groups" was generally accepted. Later, with the development of high-resolution neutron spectroscopy, βdn data have provided important information on nuclear-structure properties at intermediate excitation energy in nuclei far from stability, as well as in nuclear astrophysics. In this paper, I will present some examples of the βdn-studies performed by the Kernchemie Mainz group during the past three decades. This work has been recognized as an example of "broad scientific diversity" which has led to my nomination for the 2014 Hans A. Bethe prize.

  13. Modelling the graphite fracture mechanisms

    SciTech Connect

    Jacquemoud, C.; Marie, S.; Nedelec, M.

    2012-07-01

    In order to define a design criterion for graphite components, it is important to identify the physical phenomena responsible for the graphite fracture, to include them in a more effective modelling. In a first step, a large panel of experiments have been realised in order to build up an important database; results of tensile tests, 3 and 4 point bending tests on smooth and notched specimens have been analysed and have demonstrated an important geometry related effects on the behavior up to fracture. Then, first simulations with an elastic or an elastoplastic bilinear constitutive law have not made it possible to simulate the experimental fracture stress variations with the specimen geometry, the fracture mechanisms of the graphite being at the microstructural scale. That is the reason why a specific F.E. model of the graphite structure has been developed in which every graphite grain has been meshed independently, the crack initiation along the basal plane of the particles as well as the crack propagation and coalescence have been modelled too. This specific model has been used to test two different approaches for fracture initiation: a critical stress criterion and two criteria of fracture mechanic type. They are all based on crystallographic considerations as a global critical stress criterion gave unsatisfactory results. The criteria of fracture mechanic type being extremely unstable and unable to represent the graphite global behaviour up to the final collapse, the critical stress criterion has been preferred to predict the results of the large range of available experiments, on both smooth and notched specimens. In so doing, the experimental observations have been correctly simulated: the geometry related effects on the experimental fracture stress dispersion, the specimen volume effects on the macroscopic fracture stress and the crack propagation at a constant stress intensity factor. In addition, the parameters of the criterion have been related to

  14. Minor Actinide Transmutation Physics for Low Conversion Ratio Sodium Fast Reactors

    SciTech Connect

    Mehdi Asgari; Samuel E. Bays; Benoit Forget; Rodolfo Ferrer

    2007-09-01

    The effects of varying the reprocessing strategy used in the closed cycle of a Sodium Fast Reactor (SNF) prototype are presented in this paper. The isotopic vector from the aqueous separation of transuranic (TRU) elements in Light Water Reactor (LWR) spent nuclear fuel (SNF) is assumed to also vary according to the reprocessing strategy of the closed fuel cycle. The decay heat, gamma energy, and neutron emission of the fuel discharge at equilibrium are found to vary depending on the separation strategy. The SFR core used in this study corresponds to a burner configuration with a conversion ratio of ~0.5 based on the Super-PRISM design. The reprocessing strategies stemming from the choice of either metal or oxide fuel for the SFR are found to have a large impact on the equilibrium discharge decay heat, gamma energy, and neutron emission. Specifically, metal fuel SFR with pyroprocessing of the discharge produces the largest amount of TRU consumption (166 kg per Effective Full Power Year or EFPY), but also the highest decay heat, gamma energy, and neutron emission. On the other hand, an oxide fuel SFR with PUREX reprocessing minimizes the decay heat and related parameters of interest to a minimum, even when compared to thermal Mixed Oxide (MOX) or Inert Matrix Fuel (IMF) on a per mass basis. On an assembly basis, however, the metal SFR discharge has a lower decay heat than an equivalent oxide SFR assembly for similar minor actinide consumptions (~160 kg/EFPY.) Another disadvantage in the oxide PUREX reprocessing scenario is that there is no consumption of americium and curium, since PUREX reprocessing separates these minor actinides (MA) and requires them to be disposed of externally.

  15. Graphitic nanocapsules: design, synthesis and bioanalytical applications.

    PubMed

    Ding, Ding; Xu, Yiting; Zou, Yuxiu; Chen, Long; Chen, Zhuo; Tan, Weihong

    2017-08-03

    Graphitic nanocapsules are emerging nanomaterials which are gaining popularity along with the development of carbon nanomaterials. Their unique physical and chemical properties, as well as good biocompatibility, make them desirable agents for biomedical and bioanalytical applications. Through rational design, integrating graphitic nanocapsules with other materials provides them with additional properties which make them versatile nanoplatforms for bioanalysis. In this feature article, we present the use and performance of graphitic nanocapsules in a variety of bioanalytical applications. Based on their chemical properties, the specific merits and limitations of magnetic, hollow, and noble metal encapsulated graphitic nanocapsules are discussed. Detection, multi-modal imaging, and therapeutic applications are included. Future directions and potential solutions for further biomedical applications are also suggested.

  16. Neutronic reactor construction

    DOEpatents

    Huston, Norman E.

    1976-07-06

    1. A neutronic reactor comprising a moderator including horizontal layers formed of horizontal rows of graphite blocks, alternate layers of blocks having the rows extending in one direction, the remaining alternate layers having the rows extending transversely to the said one direction, alternate rows of blocks in one set of alternate layers having longitudinal ducts, the moderator further including slotted graphite tubes positioned in the ducts, the reactor further comprising an aluminum coolant tube positioned within the slotted tube in spaced relation thereto, bodies of thermal-neutron-fissionable material, and jackets enclosing the bodies and being formed of a corrosion-resistant material having a low neutron-capture cross section, the bodies and jackets being positioned within the coolant tube so that the jackets are spaced from the coolant tube.

  17. AIR COOLED NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1958-05-27

    A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

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

    SciTech Connect

    Eapen, Jacob; Murty, Korukonda; Burchell, Timothy

    2014-06-02

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

  19. IDENTIFICATION OF CHLOROMETHANE FROMATION PATHS DURING ELECTROCHEMICAL DECHLORINATION OF TCE USING GRAPHITE ELECTRODES

    EPA Science Inventory

    The purpose of this research is to investigate the formation of chloromethane during TCE dechlorination in a mixed electrochemical reactor using graphite electrodes. Chloromethane was the major chlorinated organic compound detected in previous dechlorination experiments. In order...

  20. IDENTIFICATION OF CHLOROMETHANE FORMATION PATHS DURING ELECTROCHEMICAL DECHLORINATION OF TCE USING GRAPHITE ELECTRODES

    EPA Science Inventory

    The purpose of this research is to investigate the formation of chloromethane during TCE dechlorination in a mixed electrochemical reactor using graphite electrodes. Chloromethane was the major chlorinated organic compound detected in previous dechlorination experiments. In order...

  1. IDENTIFICATION OF CHLOROMETHANE FROMATION PATHS DURING ELECTROCHEMICAL DECHLORINATION OF TCE USING GRAPHITE ELECTRODES

    EPA Science Inventory

    The purpose of this research is to investigate the formation of chloromethane during TCE dechlorination in a mixed electrochemical reactor using graphite electrodes. Chloromethane was the major chlorinated organic compound detected in previous dechlorination experiments. In order...

  2. IDENTIFICATION OF CHLOROMETHANE FORMATION PATHS DURING ELECTROCHEMICAL DECHLORINATION OF TCE USING GRAPHITE ELECTRODES

    EPA Science Inventory

    The purpose of this research is to investigate the formation of chloromethane during TCE dechlorination in a mixed electrochemical reactor using graphite electrodes. Chloromethane was the major chlorinated organic compound detected in previous dechlorination experiments. In order...

  3. THERMAL NUCLEAR REACTOR

    DOEpatents

    Fenning, F.W.; Jackson, R.F.

    1957-09-24

    Nuclear reactors of the graphite moderated air cooled type in which canned slugs or rods of fissile material are employed are discussed. Such a reactor may be provided with a means for detecting dust particles in the exhausted air. The means employed are lengths of dust absorbent cord suspended in vertical holes in the shielding structure above each vertical coolant flow channel to hang in the path of the cooling air issuing from the channels, and associated spindles and drive motors for hauling the cords past detectors, such as Geiger counters, for inspecting the cords periodically. This design also enables detecting the individual channel in which a fault condition may have occurred.

  4. Helium-cooled high temperature reactors

    SciTech Connect

    Trauger, D.B.

    1985-01-01

    Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg.

  5. PALLADIUM-FACILITATED ELECTROLYTIC DECHLORINATION OF 2-CHLOROBIPHENYL USING A GRANULAR-GRAPHITE ELECTRODE.

    EPA Science Inventory

    Palladium-assisted electrocatalytic dechlorination of 2-chlorobiphenyl (2-Cl BP) in aqueous solutions was conducted in a membrane-separated electrochemical reactor with granular-graphite packed electrodes. The dechlorination took place at a granular-graphite cathode while Pd was ...

  6. PALLADIUM-FACILITATED ELECTROLYTIC DECHLORINATION OF 2-CHLOROBIPHENYL USING A GRANULAR-GRAPHITE ELECTRODE.

    EPA Science Inventory

    Palladium-assisted electrocatalytic dechlorination of 2-chlorobiphenyl (2-Cl BP) in aqueous solutions was conducted in a membrane-separated electrochemical reactor with granular-graphite packed electrodes. The dechlorination took place at a granular-graphite cathode while Pd was ...

  7. Development of Subspace-based Hybrid Monte Carlo-Deterministric Algorithms for Reactor Physics Calculations

    SciTech Connect

    Abdel-Khalik, Hany S.; Zhang, Qiong

    2014-05-20

    The development of hybrid Monte-Carlo-Deterministic (MC-DT) approaches, taking place over the past few decades, have primarily focused on shielding and detection applications where the analysis requires a small number of responses, i.e. at the detector locations(s). This work further develops a recently introduced global variance reduction approach, denoted by the SUBSPACE approach is designed to allow the use of MC simulation, currently limited to benchmarking calculations, for routine engineering calculations. By way of demonstration, the SUBSPACE approach is applied to assembly level calculations used to generate the few-group homogenized cross-sections. These models are typically expensive and need to be executed in the order of 103 - 105 times to properly characterize the few-group cross-sections for downstream core-wide calculations. Applicability to k-eigenvalue core-wide models is also demonstrated in this work. Given the favorable results obtained in this work, we believe the applicability of the MC method for reactor analysis calculations could be realized in the near future.

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

    SciTech Connect

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

    2011-09-01

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

  9. Graphite Oxidation Simulation in HTR Accident Conditions

    SciTech Connect

    El-Genk, Mohamed

    2012-10-19

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

  10. Method for producing dustless graphite spheres from waste graphite fines

    DOEpatents

    Pappano, Peter J [Oak Ridge, TN; Rogers, Michael R [Clinton, TN

    2012-05-08

    A method for producing graphite spheres from graphite fines by charging a quantity of spherical media into a rotatable cylindrical overcoater, charging a quantity of graphite fines into the overcoater thereby forming a first mixture of spherical media and graphite fines, rotating the overcoater at a speed such that the first mixture climbs the wall of the overcoater before rolling back down to the bottom thereby forming a second mixture of spherical media, graphite fines, and graphite spheres, removing the second mixture from the overcoater, sieving the second mixture to separate graphite spheres, charging the first mixture back into the overcoater, charging an additional quantity of graphite fines into the overcoater, adjusting processing parameters like overcoater dimensions, graphite fines charge, overcoater rotation speed, overcoater angle of rotation, and overcoater time of rotation, before repeating the steps until graphite fines are converted to graphite spheres.

  11. Identification and Quantification of Carbon Phases in Conversion Fuel for the Transient Reactor Test Facility

    SciTech Connect

    Steele, Robert; Mata, Angelica; Dunzik-Gougar, Mary Lou; van Rooyen, Isabella

    2016-06-01

    As part of an overall effort to convert US research reactors to low-enriched uranium (LEU) fuel use, a LEU conversion fuel is being designed for the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory. TREAT fuel compacts are comprised of UO2 fuel particles in a graphitic matrix material. In order to refine heat transfer modeling, as well as determine other physical and nuclear characteristics of the fuel, the amount and type of graphite and non-graphite phases within the fuel matrix must be known. In this study, we performed a series of complementary analyses, designed to allow detailed characterization of the graphite and phenolic resin based fuel matrix. Methods included Scanning Electron and Transmission Electron Microscopies, Raman spectroscopy, X-ray Diffraction, and Dual-Beam Focused Ion Beam Tomography. Our results indicate that no single characterization technique will yield all of the desired information; however, through the use of statistical and empirical data analysis, such as curve fitting, partial least squares regression, volume extrapolation and spectra peak ratios, a degree of certainty for the quantity of each phase can be obtained.

  12. GUM Analysis for SIMS Isotopic Ratios in BEP0 Graphite Qualification Samples, Round 2

    SciTech Connect

    Gerlach, David C.; Heasler, Patrick G.; Reid, Bruce D.

    2009-01-01

    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.

  13. 10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... vault-type room shall be controlled as a separate material access area. (4) Enriched uranium scrap in... perform their duties. (6) Prior to entry into a material access area, packages shall be searched for... OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.60...

  14. Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems

    SciTech Connect

    Michael J. Driscoll; Pavel Hejzlar; Peter Yarsky; Dan Wachs; Kevan Weaver; Ken Czerwinski; Michael Pope; Cliff Davis; Theron Marshall; James Parry

    2005-12-09

    This project is organized under four major tasks (each of which has two or more subtasks) with contributions among the three collaborating organizations (MIT, INEEL and ANL-West): Task A: Core Physics and Fuel Cycle; Task B: Core Thermal Hydraulics; Task C: Plant Design Task; and D: Fuel Design.

  15. Coating method for graphite

    DOEpatents

    Banker, John G.; Holcombe, Jr., Cressie E.

    1977-01-01

    A method of limiting carbon contamination from graphite ware used in induction melting of uranium alloys is provided comprising coating the graphite surface with a suspension of Y.sub.2 O.sub.3 particles in water containing about 1.5 to 4% by weight sodium carboxymethylcellulose.

  16. Coating method for graphite

    DOEpatents

    Banker, J.G.; Holcombe, C.E. Jr.

    1975-11-06

    A method of limiting carbon contamination from graphite ware used in induction melting of uranium alloys is provided. The graphite surface is coated with a suspension of Y/sub 2/O/sub 3/ particles in water containing about 1.5 to 4 percent by weight sodium carboxymethylcellulose.

  17. Effect of mechanical disruption on the effectiveness of three reactors used for dilute acid pretreatment of corn stover Part 1: chemical and physical substrate analysis.

    PubMed

    Wang, Wei; Chen, Xiaowen; Donohoe, Bryon S; Ciesielski, Peter N; Katahira, Rui; Kuhn, Erik M; Kafle, Kabindra; Lee, Christopher M; Park, Sunkyu; Kim, Seong H; Tucker, Melvin P; Himmel, Michael E; Johnson, David K

    2014-01-01

    There is considerable interest in the conversion of lignocellulosic biomass to liquid fuels to provide substitutes for fossil fuels. Pretreatments, conducted to reduce biomass recalcitrance, usually remove at least some of the hemicellulose and/or lignin in cell walls. The hypothesis that led to this research was that reactor type could have a profound effect on the properties of pretreated materials and impact subsequent cellulose hydrolysis. Corn stover was dilute-acid pretreated using commercially relevant reactor types (ZipperClave(®) (ZC), Steam Gun (SG) and Horizontal Screw (HS)) under the same nominal conditions. Samples produced in the SG and HS achieved much higher cellulose digestibilities (88% and 95%, respectively), compared to the ZC sample (68%). Characterization, by chemical, physical, spectroscopic and electron microscopy methods, was used to gain an understanding of the effects causing the digestibility differences. Chemical differences were small; however, particle size differences appeared significant. Sum-frequency generation vibrational spectra indicated larger inter-fibrillar spacing or randomization of cellulose microfibrils in the HS sample. Simons' staining indicated increased cellulose accessibility for the SG and HS samples. Electron microscopy showed that the SG and HS samples were more porous and fibrillated because of mechanical grinding and explosive depressurization occurring with these two reactors. These structural changes most likely permitted increased cellulose accessibility to enzymes, enhancing saccharification. Dilute-acid pretreatment of corn stover using three different reactors under the same nominal conditions gave samples with very different digestibilities, although chemical differences in the pretreated substrates were small. The results of the physical and chemical analyses of the samples indicate that the explosive depressurization and mechanical grinding with these reactors increased enzyme accessibility

  18. Effect of mechanical disruption on the effectiveness of three reactors used for dilute acid pretreatment of corn stover Part 1: chemical and physical substrate analysis

    PubMed Central

    2014-01-01

    Background There is considerable interest in the conversion of lignocellulosic biomass to liquid fuels to provide substitutes for fossil fuels. Pretreatments, conducted to reduce biomass recalcitrance, usually remove at least some of the hemicellulose and/or lignin in cell walls. The hypothesis that led to this research was that reactor type could have a profound effect on the properties of pretreated materials and impact subsequent cellulose hydrolysis. Results Corn stover was dilute-acid pretreated using commercially relevant reactor types (ZipperClave® (ZC), Steam Gun (SG) and Horizontal Screw (HS)) under the same nominal conditions. Samples produced in the SG and HS achieved much higher cellulose digestibilities (88% and 95%, respectively), compared to the ZC sample (68%). Characterization, by chemical, physical, spectroscopic and electron microscopy methods, was used to gain an understanding of the effects causing the digestibility differences. Chemical differences were small; however, particle size differences appeared significant. Sum-frequency generation vibrational spectra indicated larger inter-fibrillar spacing or randomization of cellulose microfibrils in the HS sample. Simons’ staining indicated increased cellulose accessibility for the SG and HS samples. Electron microscopy showed that the SG and HS samples were more porous and fibrillated because of mechanical grinding and explosive depressurization occurring with these two reactors. These structural changes most likely permitted increased cellulose accessibility to enzymes, enhancing saccharification. Conclusions Dilute-acid pretreatment of corn stover using three different reactors under the same nominal conditions gave samples with very different digestibilities, although chemical differences in the pretreated substrates were small. The results of the physical and chemical analyses of the samples indicate that the explosive depressurization and mechanical grinding with these reactors increased

  19. Dependence of strength on particle size in graphite

    SciTech Connect

    Kennedy, E.P.; Kennedy, C.R.

    1980-06-08

    The strength to particle size relationship for specially fabricated graphites has been demonstrated and rationalized using fracture mechanics. In the past, similar studies have yielded empirical data using only commercially available material. Thus, experimental verification of these relationships has been difficult. However, the graphites of this study were fabricated by controlling the particle size ranges for a series of isotropic graphites. All graphites that were evaluated had a constant 1.85 g/cm/sup 3/ density. Thus, particle size was the only variable. This study also considered the particle size effect on other physical properties; coefficient of thermal expansion (CTE), electrical resistivity, fracture strain, and Young's modulus.

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

    SciTech Connect

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

    2012-07-01

    Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m{sup 3} each and placed in interim storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like {sup 14}C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of {sup 14}C by thermal methods. This method is based on results obtained at the Research Centre Juelich (FZJ) in the Frame of the EC programs 'Raphael' and 'Carbowaste'. The process developed at FZJ is based on a preferential oxidation of {sup 14}C in comparison to the bulk {sup 12}C. Explanations for this effect are the inhomogeneous distribution and a weaker bounding of {sup 14}C which is not incorporated in the graphite lattice. However these investigations have only been performed with graphite from the high temperature reactor Arbeitsgemeinschaft Versuchsreaktor Juelich AVR which has been operated in a non-oxidising condition or research reactor graphite operated at room temperature. The reactor Vandellos I has been operated with CO{sub 2} as coolant and significant amounts of graphite have been already oxidised. The aim of the project is to validate whether a {sup 14}C decontamination can also been achieved with graphite from Vandellos I. A second possibility under investigation is the encapsulation of the

  1. Crack growth resistance in nuclear graphites

    NASA Astrophysics Data System (ADS)

    Ouagne, Pierre; Neighbour, Gareth B.; McEnaney, Brian

    2002-05-01

    Crack growth resistance curves for the non-linear fracture parameters KR, JR and R were measured for unirradiated PGA and IM1-24 graphites that are used as moderators in British Magnox and AGR nuclear reactors respectively. All the curves show an initial rising part, followed by a plateau region where the measured parameter is independent of crack length. JR and R decreased at large crack lengths. The initial rising curves were attributed to development of crack bridges in the wake of the crack front, while, in the plateau region, the crack bridging zone and the frontal process zone, ahead of the crack tip, reached steady state values. The decreases at large crack lengths were attributed to interaction of the frontal zone with the specimen end face. Microscopical evidence for graphite fragments acting as crack bridges showed that they were much smaller than filler particles, indicating that the graphite fragments are broken down during crack propagation. There was also evidence for friction points in the crack wake zone and shear cracking of some larger fragments. Inspection of KR curves showed that crack bridging contributed ~0.4 MPa m0.5 to the fracture toughness of the graphites. An analysis of JR and R curves showed that the development of the crack bridging zone in the rising part of the curves contributed ~20% to the total work of fracture. Energies absorbed during development of crack bridges and steady state crack propagation were greater for PGA than for IM1-24 graphite. These differences reflect the greater extent of irreversible processes occurring during cracking in the coarser microtexture of PGA graphite.

  2. Analysis of granular flow in a pebble-bed nuclear reactor.

    PubMed

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

    2006-08-01

    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 6-cm-diam spheres draining in a cylindrical vessel of diameter 3.5m and height 10 m 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.

  3. Analysis of granular flow in a pebble-bed nuclear reactor

    SciTech Connect

    Rycroft, C H; Grest, Gary S; Landry, James W; Bazant, Martin Z

    2006-04-17

    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 ma jor 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° 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.

  4. Nuclear reactor shield including magnesium oxide

    DOEpatents

    Rouse, Carl A.; Simnad, Massoud T.

    1981-01-01

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

  5. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  6. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    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.

  7. Graphite matrix materials for nuclear waste isolation

    SciTech Connect

    Morgan, W.C.

    1981-06-01

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

  8. Quinoxaline polymers and copolymers derived from 1,4-bis(1'-naphthalenyloxalyl)benzene and their graphite composites. [polymer chemistry and polymer physics

    NASA Technical Reports Server (NTRS)

    Port, W. S.

    1976-01-01

    Experimental studies were performed with new polyquinoxalines and their graphite composites. Four polymers were synthesized, and then were characterized with respect to their inherent viscosity, elemental chemical analysis, mechanical, and thermodynamic properties. Structural formulas of the polymers and their precursors are given; methods of synthesis are described; and specifically examined was the preparation of polymers from 3,3' diamino-benzidine from 1,4- and 1,3- bis ((1'-napthalenyl) oxalyl) benzene respectively. Also considered was the preparation of polyquinoxalines from poly (p-benzil), and 1,2- aryldiamines.

  9. 7. Another picture of workers laying up the graphite core ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    7. Another picture of workers laying up the graphite core of the 105-B pile. This view is towards the rear of the pile. The gun barrels can be seen protruding into the pile. D-3047 - B Reactor, Richland, Benton County, WA

  10. NEW METHOD OF GRAPHITE PREPARATION

    DOEpatents

    Stoddard, S.D.; Harper, W.T.

    1961-08-29

    BS>A method is described for producing graphite objects comprising mixing coal tar pitch, carbon black, and a material selected from the class comprising raw coke, calcined coke, and graphite flour. The mixture is placed in a graphite mold, pressurized to at least 1200 psi, and baked and graphitized by heating to about 2500 deg C while maintaining such pressure. (AEC)

  11. Scalable Methods for Uncertainty Quantification, Data Assimilation and Target Accuracy Assessment for Multi-Physics Advanced Simulation of Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Khuwaileh, Bassam

    High fidelity simulation of nuclear reactors entails large scale applications characterized with high dimensionality and tremendous complexity where various physics models are integrated in the form of coupled models (e.g. neutronic with thermal-hydraulic feedback). Each of the coupled modules represents a high fidelity formulation of the first principles governing the physics of interest. Therefore, new developments in high fidelity multi-physics simulation and the corresponding sensitivity/uncertainty quantification analysis are paramount to the development and competitiveness of reactors achieved through enhanced understanding of the design and safety margins. Accordingly, this dissertation introduces efficient and scalable algorithms for performing efficient Uncertainty Quantification (UQ), Data Assimilation (DA) and Target Accuracy Assessment (TAA) for large scale, multi-physics reactor design and safety problems. This dissertation builds upon previous efforts for adaptive core simulation and reduced order modeling algorithms and extends these efforts towards coupled multi-physics models with feedback. The core idea is to recast the reactor physics analysis in terms of reduced order models. This can be achieved via identifying the important/influential degrees of freedom (DoF) via the subspace analysis, such that the required analysis can be recast by considering the important DoF only. In this dissertation, efficient algorithms for lower dimensional subspace construction have been developed for single physics and multi-physics applications with feedback. Then the reduced subspace is used to solve realistic, large scale forward (UQ) and inverse problems (DA and TAA). Once the elite set of DoF is determined, the uncertainty/sensitivity/target accuracy assessment and data assimilation analysis can be performed accurately and efficiently for large scale, high dimensional multi-physics nuclear engineering applications. Hence, in this work a Karhunen-Loeve (KL

  12. Superlubricity of graphite.

    PubMed

    Dienwiebel, Martin; Verhoeven, Gertjan S; Pradeep, Namboodiri; Frenken, Joost W M; Heimberg, Jennifer A; Zandbergen, Henny W

    2004-03-26

    Using a home-built frictional force microscope that is able to detect forces in three dimensions with a lateral force resolution down to 15 pN, we have studied the energy dissipation between a tungsten tip sliding over a graphite surface in dry contact. By measuring atomic-scale friction as a function of the rotational angle between two contacting bodies, we show that the origin of the ultralow friction of graphite lies in the incommensurability between rotated graphite layers, an effect proposed under the name of "superlubricity" [Phys. Rev. B 41, 11 837 (1990)

  13. Electroless plating of graphite with copper and nickel

    SciTech Connect

    Caturla, F.; Molina, F.; Molina-Sabio, M.; Rodriguez-Reinoso, F.; Esteban, A.

    1995-12-01

    Decommissioning in the European Union of gas-cooled nuclear reactors using graphite as the moderator will generate a large amount of irradiated graphite as waste. Graphite is a radioactive waste of relatively low activity and consequently the options considered for the management of the waste may include: (i) incineration, (ii) ocean bed disposal, (iii) deep geological disposal, and (iv) shallow land burial. In case the last is the selected mode, an appropriate conditioning procedure is necessary before final disposal, by covering the graphite with a material avoiding or reducing the emission of radionuclides to its surrounding. This work analyses the possibility of conditioning graphite pieces (with a large proportion of pores of different sizes up to 100 {micro}m) with a metal coating of copper or nickel produced by electroless plating, with the aim of completely isolating the graphite from its surrounding. Electroless plating with copper results in a very large proportion of pores filled or covered, but a fraction of the pores remain in the graphite, which decreases with increasing thickness of metal deposit. Furthermore, the copper plating is permeable to liquids and consequently the graphite does not become completely isolated from the surrounding. The percentage of porosity filled or covered by nickel deposits is similar to copper, but they are not permeable to liquids, at least when the thickness is relatively high, and consequently the access of the liquids to the graphite is rather limited. However, when electroless plating with copper is followed by nickel deposition the graphite becomes isolated from the exterior.

  14. As-Run Physics Analysis for the UCSB-1 Experiment in the Advanced Test Reactor

    SciTech Connect

    Nielsen, Joseph Wayne

    2015-09-01

    The University of California Santa Barbara (UCSB) -1 experiment was irradiated in the A-10 position of the ATR. The experiment was irradiated during cycles 145A, 145B, 146A, and 146B. Capsule 6A was removed from the test train following Cycle 145A and replaced with Capsule 6B. This report documents the as-run physics analysis in support of Post-Irradiation Examination (PIE) of the test. This report documents the as-run fluence and displacements per atom (DPA) for each capsule of the experiment based on as-run operating history of the ATR. Average as-run heating rates for each capsule are also presented in this report to support the thermal analysis.

  15. Temperature field in Graphite-Silicon-Graphite samples heated in monoellipsoidal mirror furnaces

    NASA Astrophysics Data System (ADS)

    Rivas, Damián; Haya, Rodrigo

    1999-01-01

    The heating of cylindrical compound samples in monoellipsoidal mirror furnaces is analyzed by means of a conduction-radiation model that includes the radiative exchange between the sample and the mirror, and that takes into account the temperature dependence of the physical properties of the materials that form the sample. Graphite-Silicon-Graphite samples are considered. The melting of the Silicon part, and the temperature difference between the two Graphite rods that hold the Silicon melt zone are analyzed. The relative position of the Silicon part in the compound sample turns out to be a very sensitive parameter: it affects (1) the power needed to melt the Silicon zone, and (2) the temperature difference between the solid Graphite rods.

  16. ICP-MS measurement of diffusion coefficients of Cs in NBG-18 graphite

    NASA Astrophysics Data System (ADS)

    Carter, L. M.; Brockman, J. D.; Robertson, J. D.; Loyalka, S. K.

    2015-11-01

    Graphite is used in the HGTR/VHTR as moderator and it also functions as a barrier to fission product release. Therefore, an elucidation of transport of fission products in reactor-grade graphite is required. We have measured diffusion coefficients of Cs in graphite NBG-18 using the release method, wherein we infused spheres of NBG-18 with Cs and measured the release rates in the temperature range of 1090-1395 K. We have obtained: These seem to be the first reported values of Cs diffusion coefficients in NBG-18. The values are lower than those reported for other graphites in the literature.

  17. NUCLEAR REACTORS

    DOEpatents

    Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

    1961-12-01

    An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

  18. Comparison of Oxidation Behaviors of Different Grades of Nuclear Graphite

    SciTech Connect

    Luo Xiaowei; Robin, Jean-Charles; Yu Suyuan

    2005-09-15

    The oxidation behaviors of different grades of nuclear graphite - PAEB, PCEB, PPEA, and IG-11 - were studied thermogravimetrically at 400, 800, and 1200 deg. C as a part of work to select one grade of nuclear graphite for use in a gas turbine-modular helium reactor (GT-MHR). The results showed that all grades of nuclear graphite resist oxidation at 400 deg. C. The difference in oxidation between different grades of nuclear graphite was greater at 800 deg. C than at 400 deg. C and 1200 deg. C. At 800 deg. C, for the same grade of nuclear graphite, when the centerline of the specimen is parallel to the axis of extrusion (with grain), the oxidation rate is greater than that of the graphite specimen with the centerline perpendicular to the axis of extrusion (against grain). The experimental results revealed that PPEA had the best oxidation resistance, and IG-11 had the worst due to high impurities. Moreover, the oxidation experiment exhibited that there were some oxidizable materials in unclear nuclear graphite.

  19. Characterization of nuclear graphite elastic properties using laser ultrasonic methods

    NASA Astrophysics Data System (ADS)

    Zeng, Fan W.; Han, Karen; Olasov, Lauren R.; Gallego, Nidia C.; Contescu, Cristian I.; Spicer, James B.

    2015-05-01

    Laser ultrasonic methods have been used to characterize the elastic behaviors of commercially-available and legacy nuclear graphites. Since ultrasonic techniques are sensitive to various aspects of graphite microstructure including preferred grain orientation, microcrack orientation and porosity, laser ultrasonics is a candidate technique for monitoring graphite degradation and structural integrity in environments expected in high-temperature, gas-cooled nuclear reactors. Aspects of materials texture can be assessed by studying ultrasonic wavespeeds as a function of propagation direction and polarization. Shear wave birefringence measurements, in particular, can be used to evaluate elastic anisotropy. In this work, laser ultrasonic measurements of graphite moduli have been made to provide insight into the relationship between the microstructures and the macroscopic stiffnesses of these materials. In particular, laser ultrasonic measurements have been made using laser line sources to produce shear waves with specific polarizations. By varying the line orientation relative to the sample, shear wave birefringence measurements have been recorded. Results from shear wave birefringence measurements show that an isostatically molded graphite, such as PCIB, behaves isotropically, while an extruded graphite, such as H-451, displays significant ultrasonic texture. Graphites have complicated microstructures that depend on the manufacturing processes used, and ultrasonic texture in these materials could originate from grain orientation and preferred microcrack alignment. Effects on material isotropy due to service related microstructural changes are possible and the ultimate aim of this work is to determine the degree to which these changes can be assessed nondestructively using laser ultrasonics measurements.

  20. Characterization of nuclear graphite elastic properties using laser ultrasonic methods

    SciTech Connect

    Zeng, Fan W; Han, Karen; Olasov, Lauren R; Gallego, Nidia C; Contescu, Cristian I; Spicer, James B

    2015-01-01

    Laser ultrasonic methods have been used to characterize the elastic behaviors of commercially-available and legacy nuclear graphites. Since ultrasonic techniques are sensitive to various aspects of graphite microstructure including preferred grain orientation, microcrack orientation and porosity, laser ultrasonics is a candidate technique for monitoring graphite degradation and structural integrity in environments expected in high-temperature, gas-cooled nuclear reactors. Aspects of materials texture can be assessed by studying ultrasonic wavespeeds as a function of propagation direction and polarization. Shear wave birefringence measurements, in particular, can be used to evaluate elastic anisotropy. In this work, laser ultrasonic measurements of graphite moduli have been made to provide insight into the relationship between the microstructures and the macroscopic stiffnesses of these materials. In particular, laser ultrasonic measurements have been made using laser line sources to produce shear waves with specific polarizations. By varying the line orientation relative to the sample, shear wave birefringence measurements have been recorded. Results from shear wave birefringence measurements show that an isostatically molded graphite, such as PCIB, behaves isotropically, while an extruded graphite, such as H-451, displays significant ultrasonic texture. Graphites have complicated microstructures that depend on the manufacturing processes used, and ultrasonic texture in these materials could originate from grain orientation and preferred microcrack alignment. Effects on material isotropy due to service related microstructural changes are possible and the ultimate aim of this work is to determine the degree to which these changes can be assessed nondestructively using laser ultrasonics measurements

  1. Understanding the reaction of nuclear graphite with molecular oxygen: Kinetics, transport, and structural evolution

    NASA Astrophysics Data System (ADS)

    Kane, Joshua J.; Contescu, Cristian I.; Smith, Rebecca E.; Strydom, Gerhard; Windes, William E.

    2017-09-01

    For the next generation of nuclear reactors, HTGRs specifically, an unlikely air ingress warrants inclusion in the license applications of many international regulators. Much research on oxidation rates of various graphite grades under a number of conditions has been undertaken to address such an event. However, consequences to the reactor result from the microstructural changes to the graphite rather than directly from oxidation. The microstructure is inherent to a graphite's properties and ultimately degradation to the graphite's performance must be determined to establish the safety of reactor design. To understand the oxidation induced microstructural change and its corresponding impact on performance, a thorough understanding of the reaction system is needed. This article provides a thorough review of the graphite-molecular oxygen reaction in terms of kinetics, mass and energy transport, and structural evolution: all three play a significant role in the observed rate of graphite oxidation. These provide the foundations of a microstructurally informed model for the graphite-molecular oxygen reaction system, a model kinetically independent of graphite grade, and capable of describing both the observed and local oxidation rates under a wide range of conditions applicable to air-ingress.

  2. Modelling of activation processes for GR-280 graphite at Ignalina NPP.

    PubMed

    Smaizys, Arturas; Narkunas, Ernestas; Poskas, Povilas

    2005-01-01

    Ignalina Nuclear Power Plant (INPP) operates two RBMK-1500 water-cooled graphite-moderated channel-type power reactors. The total mass of graphite in the cores of both units at INPP is about 3600 tons. Modelling of activation processes in the reactor's structural materials is necessary for decommissioning planning, because large amounts of activated structural materials (graphite, stainless steel, concrete, etc.) should be managed as radioactive waste. Knowledge of radiological characteristics and a radioactive inventory of irradiated materials are essential in planning of the decommissioning processes. The purpose of this work was to perform conservative neutron activation analysis for decommissioning purposes of INPP. ORIGEN computer code was used for the calculations. Activity levels were calculated for different nuclides present in the graphite and estimates were made how these activity levels depend on irradiation time, neutron flux value and other parameters. Obtained results were compared with the data available from other investigations for GR-280 graphite.

  3. APPLICATIONS OF LASERS AND OTHER TOPICS IN LASER PHYSICS AND TECHNOLOGY: Hybrid reactor based on laser thermonuclear fusion

    NASA Astrophysics Data System (ADS)

    Basov, N. G.; Belousov, N. I.; Grishunin, P. A.; Kalmykov, Yu K.; Lebo, I. G.; Rozanov, Vladislav B.; Sklizkov, G. V.; Subbotin, V. I.; Finkel'shteĭn, K. I.; Kharitonov, V. V.; Sherstnev, K. B.

    1987-10-01

    A physicotechnical and parametric analysis is used as the basis for a conceptual design of a thermonuclear inertial-confinement hybrid reactor as a breeder of fuel for fission nuclear power stations. It is proposed to use a laser as a driver in this reactor.

  4. Graphite filter atomizer in atomic absorption spectrometry

    NASA Astrophysics Data System (ADS)

    Katskov, Dmitri A.

    2007-09-01

    Graphite filter atomizers (GFA) for electrothermal atomic absorption spectrometry (ETAAS) show substantial advantages over commonly employed electrothermal vaporizers and atomizers, tube and platform furnaces, for direct determination of high and medium volatility elements in matrices associated with strong spectral and chemical interferences. Two factors provide lower limits of detection and shorter determination cycles with the GFA: the vaporization area in the GFA is separated from the absorption volume by a porous graphite partition; the sample is distributed over a large surface of a collector in the vaporization area. These factors convert the GFA into an efficient chemical reactor. The research concerning the GFA concept, technique and analytical methodology, carried out mainly in the author's laboratory in Russia and South Africa, is reviewed. Examples of analytical applications of the GFA in AAS for analysis of organic liquids and slurries, bio-samples and food products are given. Future prospects for the GFA are discussed in connection with analyses by fast multi-element AAS.

  5. Graphite thermal expansion reference for high temperature

    NASA Technical Reports Server (NTRS)

    Gaal, P. S.

    1974-01-01

    The design requirements of the aerospace and high-temperature nuclear reactor industries necessitate reliable thermal expansion data for graphite and other carbonaceous materials. The feasibility of an acceptable reference for calibration of expansion measuring systems that operate in carbon-rich atmospheres at temperatures ranging to 2500 C is the prime subject of this work. Present-day graphite technology provides acceptable materials for stable, reproducible references, as reflected by some of the candidate materials. The repeatability for a single specimen in a given expansion measuring system was found to be plus or minus 1%, while the combined results of several tests made on a number of samples fell within a plus or minus 2.5% band.

  6. Intercalated graphite electrical conductors

    NASA Technical Reports Server (NTRS)

    Banks, B. A.

    1983-01-01

    For years NASA has wanted to reduce the weight of spacecraft and aircraft. Experiments are conducted to find a lightweight synthetic metal to replace copper. The subject of this paper, intercalated graphite, is such a material. Intercalated graphite is made by heating petroleum or coal to remove the hydrogen and to form more covalent bonds, thus increasing the molecular weight. The coal or petroleum eventually turns to pitch, which can then be drawn into a fiber. With continued heating the pitch-based fiber releases hydrogen and forms a carbon fiber. The carbon fiber, if heated sufficiently, becomes more organized in parallel layers of hexagonally arranged carbon atoms in the form of graphite. A conductor of intercalated graphite is potentially useful for spacecraft or aircraft applications because of its low weight.

  7. Silicon on graphite cloth

    SciTech Connect

    Rand, J.A.; Cotter, J.E.; Thomas, C.J.; Ingram, A.E.; Bai, Y.B.; Ruffins, T.R.; Barnett, A.M.

    1994-12-31

    A new polycrystalline silicon solar cell has been developed that utilizes commercially available graphite cloth as a substrate. This solar cell has achieved an energy conversion efficiency of 13.4% (AM1.5G). It is believed that this is a record efficiency for a silicon solar cell formed on a graphite substrate. The silicon-on-fabric structure is comprised of a thin layer of polycrystalline silicon grown directly on the graphite fabric substrate. The structure is fabricated by a low-cost ribbon process that avoids the expense and waste of wafering. The fabric substrate gives structural support to the thin device. Critical to the achievement of device quality silicon layers is control over impurities in the graphite fabric. The silicon-on-fabric technology has the potential to supply lightweight, low-cost solar cells to weight-sensitive markets at a fraction of the cost of conventionally thinned wafers.

  8. Identification of active sonochemical zones in a triple frequency ultrasonic reactor via physical and chemical characterization techniques.

    PubMed

    Tiong, T Joyce; Liew, Derick K L; Gondipon, Ramona C; Wong, Ryan W; Loo, Yuen Ling; Lok, Matthew S T; Manickam, Sivakumar

    2017-03-01

    Coupling multiple frequencies in ultrasonic systems is one of the highly desired area of research for sonochemists, as it is known for producing synergistic effects on various ultrasonic reactions. In this study, the characteristics of a hexagonal-shaped triple frequency ultrasonic reactor with the combination frequencies of 28, 40 and 70kHz were studied. The results showed that uniform temperature increment was achieved throughout the reactor at all frequency combinations. On the other hand, sonochemiluminescence emission and degradation rate of Rhodamine B varies throughout different areas of the reactor, indicating the presence of acoustic 'hot spots' at certain areas of the reactor. Also, coupling dual and triple frequencies showed a decrease in the hydroxyl radical (OH) production, suggesting probable wave cancelling effect in the system. The results can therefore be served as a guide to optimize the usage of a triple frequency ultrasonic reactor for future applications.

  9. VIEW OF GRAPHITE BLOCK SHIELDING WALL (NOT IN ORIGINAL LOCATION), ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    VIEW OF GRAPHITE BLOCK SHIELDING WALL (NOT IN ORIGINAL LOCATION), LEVEL -15’, LABORATORY/OFFICE WING, SHOWING COOLING WATER PUMPS, LOOKING WEST - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

  10. VIEW OF GRAPHITE BLOCK SHIELDING WALL (NOT IN ORIGINAL LOCATION), ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    VIEW OF GRAPHITE BLOCK SHIELDING WALL (NOT IN ORIGINAL LOCATION), LEVEL -15’, LABORATORY/OFFICE WING, LOOKING SOUTHWEST - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

  11. Kinetics of Chronic Oxidation of NBG-17 Nuclear Graphite by Water Vapor

    SciTech Connect

    Contescu, Cristian I; Burchell, Timothy D; Mee, Robert

    2015-05-01

    This report presents the results of kinetic measurements during accelerated oxidation tests of NBG-17 nuclear graphite by low concentration of water vapor and hydrogen in ultra-high purity helium. The objective is to determine the parameters in the Langmuir-Hinshelwood (L-H) equation describing the oxidation kinetics of nuclear graphite in the helium coolant of high temperature gas-cooled reactors (HTGR). Although the helium coolant chemistry is strictly controlled during normal operating conditions, trace amounts of moisture (predictably < 0.2 ppm) cannot be avoided. Prolonged exposure of graphite components to water vapor at high temperature will cause very slow (chronic) oxidation over the lifetime of graphite components. This behavior must be understood and predicted for the design and safe operation of gas-cooled nuclear reactors. The results reported here show that, in general, oxidation by water of graphite NBG-17 obeys the L-H mechanism, previously documented for other graphite grades. However, the characteristic kinetic parameters that best describe oxidation rates measured for graphite NBG-17 are different than those reported previously for grades H-451 (General Atomics, 1978) and PCEA (ORNL, 2013). In some specific conditions, certain deviations from the generally accepted L-H model were observed for graphite NBG-17. This graphite is manufactured in Germany by SGL Carbon Group and is a possible candidate for the fuel elements and reflector blocks of HTGR.

  12. Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

    SciTech Connect

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1984-06-01

    ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.

  13. ELECTROCHEMICAL DEGRADATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODES: IDENTIFICATION AND QUALIFICATION OF DECHLORINATION PRODUCTS

    EPA Science Inventory

    TCE was successfully dechlorinated in aqueous solution using granular graphite as the cathode in a mixed electrochemical reactor. In experiments with an initial TCE concentration of less than 100 mg/l, TCE was reduced approximately by 75% in the reactor under an applied cell volt...

  14. ELECTROCHEMICAL DEGRADATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODES: IDENTIFICATION AND QUALIFICATION OF DECHLORINATION PRODUCTS

    EPA Science Inventory

    TCE was successfully dechlorinated in aqueous solution using granular graphite as the cathode in a mixed electrochemical reactor. In experiments with an initial TCE concentration of less than 100 mg/l, TCE was reduced approximately by 75% in the reactor under an applied cell volt...

  15. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  16. Neutrino mass hierarchy and precision physics with medium-baseline reactors: Impact of energy-scale and flux-shape uncertainties

    NASA Astrophysics Data System (ADS)

    Capozzi, F.; Lisi, E.; Marrone, A.

    2015-11-01

    Nuclear reactors provide intense sources of electron antineutrinos, characterized by few-MeV energy E and unoscillated spectral shape Φ (E ). High-statistics observations of reactor neutrino oscillations over medium-baseline distances L ˜O (50 ) km would provide unprecedented opportunities to probe both the long-wavelength mass-mixing parameters (δ m2 and θ12) and the short-wavelength ones (Δ mee 2 and θ13), together with the subtle interference effects associated with the neutrino mass hierarchy (either normal or inverted). In a given experimental setting—here taken as in the JUNO project for definiteness—the achievable hierarchy sensitivity and parameter accuracy depend not only on the accumulated statistics but also on systematic uncertainties, which include (but are not limited to) the mass-mixing priors and the normalizations of signals and backgrounds. We examine, in addition, the effect of introducing smooth deformations of the detector energy scale, E →E'(E ), and of the reactor flux shape, Φ (E )→Φ'(E ), within reasonable error bands inspired by state-of-the-art estimates. It turns out that energy-scale and flux-shape systematics can noticeably affect the performance of a JUNO-like experiment, both on the hierarchy discrimination and on precision oscillation physics. It is shown that a significant reduction of the assumed energy-scale and flux-shape uncertainties (by, say, a factor of 2) would be highly beneficial to the physics program of medium-baseline reactor projects. Our results also shed some light on the role of the inverse-beta decay threshold, of geoneutrino backgrounds, and of matter effects in the analysis of future reactor oscillation data.

  17. Solid State Physics. Nitrogen Adsorption by Thermoexfoliated Graphite / Slāpekļa Adsorbcija Uz Termoeksfoliētā Grafīta

    NASA Astrophysics Data System (ADS)

    Grehov, V.; Kalnacs, J.; Matzui, L.; Knite, M.; Murashov, A.; Vilken, A.

    2013-02-01

    Adsorption by thermochemically exfoliated graphite (TEG) is studied and compared with that by other carbon structures under the same conditions. In BET determination of the specific surface area (SBET) for the TEG samples it was found that good approximation could be observed in two different pressure ranges. Such ranges of BET approximation are also visible in the isotherms of aquadag and milled graphite. The experimental results are discussed and their interpretation proposed Ar sorbcijas iekārtu Autosorb-1 (Quantochrome Instruments Co, Florida, USA) pētīta termiski eksfoliēta grafīta slāpekļa sorbcijas spēja salīdzinājumā ar citu oglekļa struktūru sorbciju tādos pašos apstākļos. Atrasti divi izotermu rajoni, kas raksturojas ar mazu (SBET1) un palielinātu (SBET2) īpatnējo virsmu.. Šāda veida izotermas raksturīgas slikti adsorbējošiem materiāliem, pie kādiem pieder arī akvadags. Labi adsorbējošās oglekļa struktūrām, tādām kā aktīvā ogle un oglekļa nanocaurules, raksturīgas cita veida izotermas. Apspriesta eksperimentālā rezultāta interpretācija

  18. Temperature effects on the behavior of carbon 14 in nuclear graphite

    NASA Astrophysics Data System (ADS)

    Silbermann, G.; Moncoffre, N.; Toulhoat, N.; Bérerd, N.; Perrat-Mabilon, A.; Laurent, G.; Raimbault, L.; Sainsot, P.; Rouzaud, J.-N.; Deldicque, D.

    2014-08-01

    The dismantling of the 1st French generation UNGG (Uranium Naturel Graphite Gas) nuclear reactors operated by the French utility, EDF (Electricité de France) will generate around 17,000 tons of irradiated graphite wastes that have to be disposed of. 14C is one of the main radioactive dose contributors over 10,000 years. For the management of this waste, it is mandatory to get an accurate estimation of 14C. The general aim of our work is therefore to simulate the behavior of 14C in nuclear graphite and to elucidate the coupled and decoupled effects of temperature, irradiation and radiolytic corrosion that mainly influence 14C behavior in graphite during reactor operation. This paper focuses on the behavior of 13C implanted into nuclear graphite and used to simulate the presence of 14C displaced from its original structural site through recoil during neutron irradiation. It aims at evaluating both the temperature and the disorder level of the implanted graphite structure effects on 13C migration using two complementary techniques, NRA and SIMS, to evaluate the 13C distribution at the millimeter and micrometer lateral scales respectively. Raman micro-spectroscopy is used to check the graphite structure evolution. The results show that 13C is not released up to 1600 °C whatever the initial structural disorder level of the implanted graphite. This might be due to the fact that 13C might be trapped into interstitial clusters. The extrapolation of our results to the behavior of 14C shows that reactor temperatures (200-500 °C) did not induce any 14C release. Moreover, as long as there is no gasification of the graphite matrix, high temperatures tend to stabilize 14C into the remaining graphite structure. This fact has to be considered in case of high temperature purification of 14C from irradiated graphite.

  19. Analysis on Thermal Conductivity of Graphite/Al Composite by Experimental and Modeling Study

    NASA Astrophysics Data System (ADS)

    Xue, C.; Bai, H.; Tao, P. F.; Jiang, N.; Wang, S. L.

    2017-01-01

    Graphite/Al composites were fabricated by vacuum hot pressing technology in this study. The main factors affecting the thermal conductivity (TC) of graphite/Al composites were deeply investigated by experimental and modeling study. The results showed that the TC of graphite/Al composite can be improved via designing the preferred orientation of graphite flakes, selecting graphite flakes with large diameter, increasing the content of graphite flakes in graphite/Al composite and solving the poor wettability between Al and graphite. The modified model can well predict the heat transfer behavior of graphite/Al composite.

  20. Improved performance of microbial fuel cell using combination biocathode of graphite fiber brush and graphite granules

    NASA Astrophysics Data System (ADS)

    Zhang, Guo-dong; Zhao, Qing-liang; Jiao, Yan; Zhang, Jin-na; Jiang, Jun-qiu; Ren, Nanqi; Kim, Byung Hong

    2011-08-01

    The efficiency and sustainability of microbial fuel cell (MFC) are heavily dependent on the cathode performance. We show here that the use of graphite fiber brush (GBF) together with graphite granules (GGs) as a basal material for biocathode (MFC reactor type R1) significantly improve the performance of a MFC compared with MFCs using GGs (MFC reactor type R2) or GFB (MFC reactor type R3) individually. Compared with R3, the use of the combination biocathode (R1) can shorten the start-up time by 53.75%, improve coulombic efficiencies (CEs) by 21.0 ± 2.7% at external resistance (REX) of 500 Ω, and increase maximum power densities by 38.2 ± 12.6%. Though the start-up time and open circuit voltage (OCV) of the reactor R2 are similar to R1, the CE (REX = 500 Ω) and maximum power density of R2 are 21.4 ± 1.7% and 38.2 ± 15.6% lower than that of R1. Fluorescence in situ hybridization (FISH) analyses indicate the bacteria on cathodes of R1 and R2 are richer than that of R3. Molecular taxonomic analyses reveal that the biofilm formed on the biocathode surface is dominated by strains belonging to Nitrobacter, Achromobacter, Acinetobacter, and Bacteroidetes. Combination of GFB and GGs as biocathode material in MFC is more efficient and can achieve sustainable electricity recovery from organic substances, which substantially increases the viability and sustainability of MFCs.

  1. Understanding the nature of "superhard graphite".

    PubMed

    Boulfelfel, Salah Eddine; Oganov, Artem R; Leoni, Stefano

    2012-01-01

    Numerous experiments showed that on cold compression graphite transforms into a new superhard and transparent allotrope. Several structures with different topologies have been proposed for this phase. While experimental data are compatible with most of these models, the only way to solve this puzzle is to find which structure is kinetically easiest to form. Using state-of-the-art molecular-dynamics transition path sampling simulations, we investigate kinetic pathways of the pressure-induced transformation of graphite to various superhard candidate structures. Unlike hitherto applied methods for elucidating nature of superhard graphite, transition path sampling realistically models nucleation events necessary for physically meaningful transformation kinetics. We demonstrate that nucleation mechanism and kinetics lead to M-carbon as the final product. W-carbon, initially competitor to M-carbon, is ruled out by phase growth. Bct-C₄ structure is not expected to be produced by cold compression due to less probable nucleation and higher barrier of formation.

  2. EMI Shields made from intercalated graphite composites

    NASA Technical Reports Server (NTRS)

    Gaier, James R.; Terry, Jennifer

    1995-01-01

    Electromagnetic interference (EMI) shielding typically makes up about twenty percent of the mass of a spacecraft power system. Graphite fiber/polymer composites have significantly lower densities and higher strengths than aluminum, the present material of choice for EMI shields, but they lack the electrical conductivity that enables acceptable shielding effectiveness. Bromine intercalated pitch-based graphite/epoxy composites have conductivities fifty times higher than conventional structural graphite fibers. Calculations are presented which indicate that EMI shields made from such composites can have sufficient shielding at less than 20% of the mass of conventional aluminum shields. EMI shields provide many functions other than EMI shielding including physical protection, thermal management, and shielding from ionizing radiation. Intercalated graphite composites perform well in these areas also. Mechanically, they have much higher specific strength and modulus than aluminum. They also have shorter half thicknesses for x-rays and gamma radiation than aluminum. Thermally, they distribute infra-red radiation by absorbing and re-radiating it rather than concentrating it by reflection as aluminum does. The prospects for intercalated graphite fiber/polymer composites for EMI shielding are encouraging.

  3. Electronic speckle pattern interferometry for fracture expansion in nuclear graphite based on PDE image processing methods

    NASA Astrophysics Data System (ADS)

    Tang, Chen; Zhang, Junjiang; Sun, Chen; Su, Yonggang; Su, Kai Leung

    2015-05-01

    Nuclear graphite has been widely used as moderating and reflecting materials. However, due to severe neutron irradiation under high temperature, nuclear graphite is prone to deteriorate, resulting in massive microscopic flaws and even cracks under large stress in the later period of its service life. It is indispensable, therefore, to understand the fracture behavior of nuclear graphite to provide reference to structural integrity and safety analysis of nuclear graphite members in reactors. In this paper, we investigated the fracture expansion in nuclear graphite based on PDE image processing methods. We used the second-order oriented partial differential equations filtering model (SOOPDE) to denoise speckle noise, then used the oriented gradient vector fields for to obtain skeletons. The full-field displacement of fractured nuclear graphite and the location of the crack tip were lastly measured under various loading conditions.

  4. The microstructural modelling of nuclear grade graphite

    NASA Astrophysics Data System (ADS)

    Hall, G.; Marsden, B. J.; Fok, S. L.

    2006-07-01

    By using the finite element method it has been possible to simulate irradiation-induced property changes, namely dimensional and Young's modulus changes, from which the probable microstructural mechanisms have been identified. In the finite element models, both property changes were shown to be dependent upon the filler particle dimensional changes and the accommodation porosity. However, these need to be corroborated through the examination of actual specimens. Further work is also required to adapt the procedure to other graphites, reactor conditions, and material properties.

  5. Effects of Oxidation on Oxidation-Resistant Graphite

    SciTech Connect

    Windes, William; Smith, Rebecca; Carroll, Mark

    2015-05-01

    The Advanced Reactor Technology (ART) Graphite Research and Development Program is investigating doped nuclear graphite grades that exhibit oxidation resistance through the formation of protective oxides on the surface of the graphite material. In the unlikely event of an oxygen ingress accident, graphite components within the VHTR core region are anticipated to oxidize so long as the oxygen continues to enter the hot core region and the core temperatures remain above 400°C. For the most serious air-ingress accident which persists over several hours or days the continued oxidation can result in significant structural damage to the core. Reducing the oxidation rate of the graphite core material during any air-ingress accident would mitigate the structural effects and keep the core intact. Previous air oxidation testing of nuclear-grade graphite doped with varying levels of boron-carbide (B4C) at a nominal 739°C was conducted for a limited number of doped specimens demonstrating a dramatic reduction in oxidation rate for the boronated graphite grade. This report summarizes the conclusions from this small scoping study by determining the effects of oxidation on the mechanical strength resulting from oxidation of boronated and unboronated graphite to a 10% mass loss level. While the B4C additive did reduce mechanical strength loss during oxidation, adding B4C dopants to a level of 3.5% or more reduced the as-fabricated compressive strength nearly 50%. This effectively minimized any benefits realized from the protective film formed on the boronated grades. Future work to infuse different graphite grades with silicon- and boron-doped material as a post-machining conditioning step for nuclear components is discussed as a potential solution for these challenges in this report.

  6. EXPERIMENTAL LIQUID METAL FUEL REACTOR

    DOEpatents

    Happell, J.J.; Thomas, G.R.; Denise, R.P.; Bunts, J.L. Jr.

    1962-01-23

    A liquid metal fuel nuclear fission reactor is designed in which the fissionable material is dissolved or suspended in a liquid metal moderator and coolant. The liquid suspension flows into a chamber in which a critical amount of fissionable material is obtained. The fluid leaves the chamber and the heat of fission is extracted for power or other utilization. The improvement is in the support arrangement for a segrnented graphite core to permit dif ferential thermal expansion, effective sealing between main and blanket liquid metal flows, and avoidance of excessive stress development in the graphite segments. (AEC)

  7. Effect of Reacting Surface Density on the Overall Graphite Oxidation Rate

    SciTech Connect

    Chang H. Oh; Eung Kim; Jong Lim; Richard Schultz; David Petti

    2009-05-01

    Graphite oxidation in an air-ingress accident is presently a very important issue for the reactor safety of the very high temperature gas cooled-reactor (VHTR), the concept of the next generation nuclear plant (NGNP) because of its potential problems such as mechanical degradation of the supporting graphite in the lower plenum of the VHTR might lead to core collapse if the countermeasure is taken carefully. The oxidation process of graphite has known to be affected by various factors, including temperature, pressure, oxygen concentration, types of graphite, graphite shape and size, flow distribution, etc. However, our recent study reveals that the internal pore characteristics play very important roles in the overall graphite oxidation rate. One of the main issues regarding graphite oxidation is the potential core collapse problem that may occur following the degradation of graphite mechanical strength. In analyzing this phenomenon, it is very important to understand the relationship between the degree of oxidization and strength degradation. In addition, the change of oxidation rate by graphite oxidation degree characterization by burn-off (ratio of the oxidized graphite density to the original density) should be quantified because graphite strength degradation is followed by graphite density decrease, which highly affects oxidation rates and patterns. Because the density change is proportional to the internal pore surface area, they should be quantified in advance. In order to understand the above issues, the following experiments were performed: (1)Experiment on the fracture of the oxidized graphite and validation of the previous correlations, (2) Experiment on the change of oxidation rate using graphite density and data collection, (3) Measure the BET surface area of the graphite. The experiments were performed using H451 (Great Lakes Carbon Corporation) and IG-110 (Toyo Tanso Co., Ltd) graphite. The reason for the use of those graphite materials is because

  8. Comparison of irradiation behaviour of HTR graphite grades

    NASA Astrophysics Data System (ADS)

    Heijna, M. C. R.; de Groot, S.; Vreeling, J. A.

    2017-08-01

    The INNOGRAPH irradiations were executed in the High Flux Reactor (HFR) in Petten by NRG supported by the European Framework programs HTR-M, RAPHAEL, and ARCHER to generate data on the irradiation behaviour of graphite grades for High Temperature Reactor (HTR) application available at that time. Samples of the graphite grades NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, and IG-110 have been irradiated at 750 °C and 950 °C. The inherent scatter induced by the probabilistic material behaviour of graphite requires uncertainty and scatter induced by test conditions and post-irradiation examination to be minimized. The INNOGRAPH irradiations supplied an adequate number of irradiated samples to enable accurate determination of material properties and their evolution under irradiation. This allows comparison of different graphite grades and a qualitative assessment of their appropriateness for HTR applications, as a basis of selection, design and core component lifetime. The results indicate that coarse grained graphite grades exhibit more favourable behaviour for application in HTRs due to their low dimensional anisotropy and fracture propagation resilience.

  9. GAS COOLED NUCLEAR REACTORS

    DOEpatents

    Long, E.; Rodwell, W.

    1958-06-10

    A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

  10. Physical and Biological Release of Poly- and Perfluoroalkyl Substances (PFASs) from Municipal Solid Waste in Anaerobic Model Landfill Reactors.

    PubMed

    Allred, B McKay; Lang, Johnsie R; Barlaz, Morton A; Field, Jennifer A

    2015-07-07

    A wide variety of consumer products that are treated with poly- and perfluoroalkyl substances (PFASs) and related formulations are disposed of in landfills. Landfill leachate has significant concentrations of PFASs and acts as secondary point sources to surface water. This study models how PFASs enter leachate using four laboratory-scale anaerobic bioreactors filled with municipal solid waste (MSW) and operated over 273 days. Duplicate reactors were monitored under live and abiotic conditions to evaluate influences attributable to biological activity. The biologically active reactors simulated the methanogenic conditions that develop in all landfills, producing ∼140 mL CH4/dry g refuse. The average total PFAS leaching measured in live reactors (16.7 nmol/kg dry refuse) was greater than the average for abiotic reactors (2.83 nmol/kg dry refuse), indicating biological processes were primarily responsible for leaching. The low-level leaching in the abiotic reactors was primarily due to PFCAs ≤C8 (2.48 nmol/kg dry refuse). Concentrations of known biodegradation intermediates, including methylperfluorobutane sulfonamide acetic acid and the n:2 and n:3 fluorotelomer carboxylates, increased steadily after the onset of methanogenesis, with the 5:3 fluorotelomer carboxylate becoming the single most concentrated PFAS observed in live reactors (9.53 nmol/kg dry refuse).

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

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-09-01

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

  12. Recompressed exfoliated graphite articles

    DOEpatents

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

    2013-08-06

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

  13. Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

    SciTech Connect

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

    2007-05-01

    Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm / shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm / shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the “International Handbook of Evaluated Criticality Safety Benchmark Experiments” have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement / shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy

  14. Validation of SCALE and the TRITON Depletion Sequence for Gas-Cooled Reactor Analysis

    SciTech Connect

    DeHart, Mark D; Pritchard, Megan L

    2008-01-01

    The very-high-temperature reactor (VHTR) is an advanced reactor concept that uses graphite-moderated fuel and helium gas as a coolant. At present there are two primary VHTR reactor designs under consideration for development: in the pebble-bed reactor, a core is loaded with 'pebbles' consisting of 6 cm diameter spheres, while in a high-temperature gas-cooled reactor, fuel rods are placed within prismatic graphite blocks. In both systems, fuel elements (spheres or rods) are comprised of tristructural-isotropic (TRISO) fuel particles. The TRISO particles are either dispersed in the matrix of a graphite pebble for the pebble-bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks for the prismatic concept. Two levels of heterogeneity exist in such fuel designs: (1) microspheres of TRISO particles dispersed in a graphite matrix of a cylindrical or spherical shape, and (2) neutron interactions at the rod-to-rod or sphere-to-sphere level. Such double heterogeneity (DH) provides a challenge to multigroup cross-section processing methods, which must treat each level of heterogeneity separately. A new capability to model doubly heterogeneous systems was added to the SCALE system in the release of Version 5.1. It was included in the control sequences CSAS and CSAS6, which use the Monte Carlo codes KENO V.a and KENO-VI, respectively, for three-dimensional neutron transport analyses and in the TRITON sequence, which uses the two-dimensional lattice physics code NEWT along with both versions of KENO for transport and depletion analyses. However, the SCALE 5.1 version of TRITON did not support the use of the DH approach for depletion. This deficiency has been addressed, and DH depletion will be available as an option in the upcoming release of SCALE 6. At present Oak Ridge National Laboratory (ORNL) staff are developing a set of calculations that may be used to validate SCALE for DH calculations. This paper discusses the results of

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

    SciTech Connect

    Burchell, Timothy D; Bratton, Rob; Marsden, Barry; Srinivasan, Makuteswara; Penfield, Scott; Mitchell, Mark; Windes, Will

    2008-03-01

    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

  16. Graphite to Inconel brazing using active filler metal

    SciTech Connect

    King, J.F.; Baity, F.W.; Walls, J.C.; Hoffman, D.J.

    1989-01-01

    Ion cyclotron resonant frequency (ICRF) antennas are designed to supply large amounts of auxiliary heating power to fusion-grade plasmas in the Toroidal Fusion Test Reactor (TFTR) and Tore Supra fusion energy experiments. A single Faraday shield structure protects a pair of resonant double loops which are designed to launch up to 2 MW of power per loop. The shield consists of two tiers of actively cooled Inconel alloy tubes with the front tier being covered with semicircular graphite tiles. Successful operation of the antenna requires the making of high integrity bonds between the Inconel tubes and graphite tiles by brazing. This paper discusses this process.

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

    SciTech Connect

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

    2008-12-01

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

  18. Coatings for graphite fibers

    NASA Technical Reports Server (NTRS)

    Galasso, F. S.; Scola, D. A.; Veltri, R. D.

    1980-01-01

    Graphite fibers released from composites during burning or an explosion caused shorting of electrical and electronic equipment. Silicon carbide, silica, silicon nitride and boron nitride were coated on graphite fibers to increase their electrical resistances. Resistances as high as three orders of magnitude higher than uncoated fiber were attained without any significant degradation of the substrate fiber. An organo-silicone approach to produce coated fibers with high electrical resistance was also used. Celion 6000 graphite fibers were coated with an organo-silicone compound, followed by hydrolysis and pyrolysis of the coating to a silica-like material. The shear and flexural strengths of composites made from high electrically resistant fibers were considerably lower than the shear and flexural strengths of composites made from the lower electrically resistant fibers. The lower shear strengths of the composites indicated that the coatings on these fibers were weaker than the coating on the fibers which were pyrolyzed at higher temperature.

  19. Cesium diffusion in graphite

    SciTech Connect

    Evans, R.B. III; Davis, W. Jr.; Sutton, A.L. Jr.

    1980-05-01

    Experiments on diffusion of /sup 137/Cs in five types of graphite were performed. The document provides a completion of the report that was started and includes a presentation of all of the diffusion data, previously unpublished. Except for data on mass transfer of /sup 137/Cs in the Hawker-Siddeley graphite, analyses of experimental results were initiated but not completed. The mass transfer process of cesium in HS-1-1 graphite at 600 to 1000/sup 0/C in a helium atmosphere is essentially pure diffusion wherein values of (E/epsilon) and ..delta..E of the equation D/epsilon = (D/epsilon)/sub 0/ exp (-..delta..E/RT) are about 4 x 10/sup -2/ cm/sup 2//s and 30 kcal/mole, respectively.

  20. Superlubricity of Graphite

    NASA Astrophysics Data System (ADS)

    Dienwiebel, Martin; Verhoeven, Gertjan S.; Pradeep, Namboodiri; Frenken, Joost W.; Heimberg, Jennifer A.; Zandbergen, Henny W.

    2004-03-01

    Using a home-built frictional force microscope that is able to detect forces in three dimensions with a lateral force resolution down to 15 pN, we have studied the energy dissipation between a tungsten tip sliding over a graphite surface in dry contact. By measuring atomic-scale friction as a function of the rotational angle between two contacting bodies, we show that the origin of the ultralow friction of graphite lies in the incommensurability between rotated graphite layers, an effect proposed under the name of “superlubricity” [

    M. Hirano and K. Shinjo, Phys. Rev. BPRBMDO0163-1829 41, 11 837 (1990)10.1103/PhysRevB.41.11837
    ].

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

    NASA Astrophysics Data System (ADS)

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

    2014-12-01

    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 5 K, 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.

  2. Characterization of 14C in Neutron-Irradiated Graphite

    NASA Astrophysics Data System (ADS)

    LaBrier, Daniel Patrick

    A long-term radiological concern regarding irradiated graphite waste is the presence of the radionuclide 14C. Recent studies suggest that a significant portion of 14C contamination present in reactor-irradiated graphite is concentrated on the surface and within near-surface layers. Methods for treating irradiated graphite waste (e.g. pyrolysis, oxidation) in order to remove 14C-bearing species from the bulk graphite are being investigated to lend guidance in optimizing long-term disposal strategies. Characterization studies were performed in order to determine the chemical nature of 14C on irradiated graphite surfaces. Samples of the nuclear-grade graphite NBG-25 were irradiated in a neutron flux of 10 14 n/cm2-s for 360 days at the Advanced Test Reactor (at the Idaho National Laboratory). Surface-sensitive analysis techniques (XPS, ToF-SIMS, SEM/EDS and Raman) were employed to determine the type, location and quantity of specific chemical species and bonds that were present on the surfaces of irradiated graphite samples. Several 14C precursor species were identified on the surfaces of irradiated NBG-25; the quantities of these species decrease at sub-surface depths, which, is consistent with the observation of high concentrations of 14C on the surfaces of graphite reactor components. The elevated presence of surface oxide complexes on irradiated NBG-25 surfaces was attributed directly to neutron irradiation. Pathways for the release of 14C were identified for irradiated NBG-25: carboxyls and lactones (14CO 2), and carbonyls, ethers and quinones (14CO). Increased amounts of C-O and C=O bonding were observed on irradiated NBG-25 surfaces (when compared to unirradiated samples) in the form of interlattice (e.g. ether) and dangling (e.g. carboxyl or quinone) bonds; the quantities of these bond types also decrease at sub-surface depths. The results of this study are consistent with thermal treatment studies that indicate that the primary candidates for the release of

  3. Lithium-Graphite Secondary Battery.

    DTIC Science & Technology

    1976-12-01

    Used in the experiment that studied the effect of operating current. 6. Li/LiClO 4, PC (0.9M)/Graphite + Graphite glue on carbon cloth. 7. Li/ LiBF4 ...DMSU (1.0M)/Graphite + Graphite glue on carbon cloth. 8. Li/ LiBF4 , PC (1.5M)/Graphite + Graphite glue on carbon cloth. 9. Li/LiClO4, DMSU (2.1M)/Pt. 10... LiBF4 , PC(1.5 M)/Graphite + Graphite glue on carbon cloth. Cycles 1 and 2 51 24. Same as 23. Cycle no. 3, 1-6.3 mA, Q n=2.17 mEq 52 25. Typical

  4. Functionalization of Natural Graphite for Use as Reinforcement in Polymer Nanocomposites.

    PubMed

    Araujo, Rafael; Marques, Maria F V; Jonas, Renato; Grafova, Iryna; Grafov, Andriy

    2015-08-01

    Graphite is a naturally abundant material that has been used as reinforcing filler to produce polymeric nanocomposites for various applications including automotive, aerospace and electric-electronic. The objective of this study was to develop methodologies of graphite nanosheets preparation and for incorporation into polymer matrices. By means of different chemical and physical treatments, natural graphite was modified and subsequently characterized by X-ray diffraction (XRD), infrared spectroscopy (FTIR), scanning electron microscopy (SEM), thermogravimetry (TGA) and the particle size determination. The results obtained clearly show that after the treatments employed, polar chemical groups were inserted on the natural graphite surface. Nanosized graphite particles of high aspect ratio were obtained.

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

    DOEpatents

    Makowiecki, Daniel M.; Ramsey, Philip B.; Juntz, Robert S.

    1995-01-01

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

  6. Improved graphite furnace atomizer

    DOEpatents

    Siemer, D.D.

    1983-05-18

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

  7. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect

    P. Delmolino

    2005-05-06

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

  8. Tungsten Deposition on Graphite using Plasma Enhanced Chemical Vapour Deposition.

    NASA Astrophysics Data System (ADS)

    Sharma, Uttam; Chauhan, Sachin S.; Sharma, Jayshree; Sanyasi, A. K.; Ghosh, J.; Choudhary, K. K.; Ghosh, S. K.

    2016-10-01

    The tokamak concept is the frontrunner for achieving controlled thermonuclear reaction on earth, an environment friendly way to solve future energy crisis. Although much progress has been made in controlling the heated fusion plasmas (temperature ∼ 150 million degrees) in tokamaks, technological issues related to plasma wall interaction topic still need focused attention. In future, reactor grade tokamak operational scenarios, the reactor wall and target plates are expected to experience a heat load of 10 MW/m2 and even more during the unfortunate events of ELM's and disruptions. Tungsten remains a suitable choice for the wall and target plates. It can withstand high temperatures, its ductile to brittle temperature is fairly low and it has low sputtering yield and low fuel retention capabilities. However, it is difficult to machine tungsten and hence usages of tungsten coated surfaces are mostly desirable. To produce tungsten coated graphite tiles for the above-mentioned purpose, a coating reactor has been designed, developed and made operational at the SVITS, Indore. Tungsten coating on graphite has been attempted and successfully carried out by using radio frequency induced plasma enhanced chemical vapour deposition (rf -PECVD) for the first time in India. Tungsten hexa-fluoride has been used as a pre-cursor gas. Energy Dispersive X-ray spectroscopy (EDS) clearly showed the presence of tungsten coating on the graphite samples. This paper presents the details of successful operation and achievement of tungsten coating in the reactor at SVITS.

  9. Understanding the reaction of nuclear graphite with molecular oxygen: Kinetics, transport, and structural evolution

    DOE PAGES

    Kane, Joshua J.; Contescu, Cristian I.; Smith, Rebecca E.; ...

    2017-06-08

    A thorough understanding of oxidation is important when considering the health and integrity of graphite components in graphite reactors. For the next generation of graphite reactors, HTGRs specifically, an unlikely air ingress has been deemed significant enough to have made its way into the licensing applications of many international licensing bodies. While a substantial body of literature exists on nuclear graphite oxidation in the presence of molecular oxygen and significant efforts have been made to characterize oxidation kinetics of various grades, the value of existing information is somewhat limited. Often, multiple competing processes, including reaction kinetics, mass transfer, and microstructuralmore » evolution, are lumped together into a single rate expression that limits the ability to translate this information to different conditions. This article reviews the reaction of graphite with molecular oxygen in terms of the reaction kinetics, gas transport, and microstructural evolution of graphite. It also presents the foundations of a model for the graphite-molecular oxygen reaction system that is kinetically independent of graphite grade, and is capable of describing both the bulk and local oxidation rates under a wide range of conditions applicable to air-ingress.« less

  10. Graphite-based photovoltaic cells

    DOEpatents

    Lagally, Max; Liu, Feng

    2010-12-28

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

  11. A SCALE 5.0 Reactor Physics Assessment using the Module TRITON against Mixed Oxide (MOX) OECD/NEA Benchmarks

    SciTech Connect

    Saccheri, J.G.B.; Diamond, D.J.

    2006-07-01

    Reactor physics numerical benchmarks have been performed at the Brookhaven National Laboratory (BNL) with the software package SCALE 5.0 and its TRITON module to assess their capability to predict neutronics parameters for mixed oxide (MOX) fuels. The results of such calculations are herein presented. Specifically, BNL results for neutron multiplication factors (kINF), neutron fluxes and fuel burnup have been added to published OECD/NEA benchmarks for MOX fuels and particular emphasis has been given to the impact of cross-section libraries and their energy structure on the results. Among the OECD/NEA published benchmarks two have been considered here: the first one models a fuel pin surrounded by moderator, in which two different MOX fuels can be introduced, and for each one of them kINF and neutron fluxes as a function of burnup are calculated. The second one includes both a fuel pin case and a macro-cell case (a heterogeneous 30 by 30 configuration of fuel pins), for which the void coefficient is determined by calculating kINF at zero burnup as a function of moderation. The calculations are repeated for several combinations of MOX and uranium oxide fuels using several different cross-section libraries. The final results have been compared with each other. This study shows that SCALE 5.0 (with TRITON) overall performs in line with the other codes in the benchmark, but the results are dependent on the energy group structure of the cross section libraries used. For instance, when fissile plutonium is increased in the fuel, TRITON results become slightly divergent with burnup (with respect to the other codes in the benchmark) and if the standard 44-group library provided with SCALE 5.0 is used void coefficient calculations become inadequate for very low void (below 10% of the operating value of moderator density). Moreover, the prediction capabilities of the code are shown to be dependent on the MOX fuel enrichment and the MOX isotopic composition. (authors)

  12. Physics concept on the constellation type fissile fuels and its application to the prospective Th-{sup 233}U reactor

    SciTech Connect

    Jiahua Zhange

    1994-12-31

    In contrast with the conventional nuclear reactor which usually fuelled with one single fissile nuclide, a constellation type fissile fuels reactor consists of a parent nuclide such as {sup 232}Th or {sup 238}U and its whole family of neutron generated daughter nuclides. All of them are regarded as fissile fuels but of quite different fission ability. The concentration of each daughter nuclide is determined by its saturate concentration ratio with the parent nuclide. In such fuel system, the whole fuel consumed by neutron reaction almost completely results in fission production. In this article, some interesting properties of such fuel system, determination of the saturate concentration of each daughter nuclide and applicability to Th-{sup 233}U reactor will be discussed.

  13. GRAPHITE BONDING METHOD

    DOEpatents

    King, L.D.P.

    1964-02-25

    A process for bonding or joining graphite members together in which a thin platinum foil is placed between the members, heated in an inert atmosphere to a temperature of 1800 deg C, and then cooled to room temperature is described. (AEC)

  14. Coatings for Graphite Fibers

    NASA Technical Reports Server (NTRS)

    Galasso, F. S.; Scola, D. A.; Veltri, R. D.

    1980-01-01

    Several approaches for applying high resistance coatings continuously to graphite yarn were investigated. Two of the most promising approaches involved (1) chemically vapor depositing (CVD) SiC coatings on the surface of the fiber followed by oxidation, and (2) drawing the graphite yarn through an organo-silicone solution followed by heat treatments. In both methods, coated fibers were obtained which exhibited increased electrical resistances over untreated fibers and which were not degraded. This work was conducted in a previous program. In this program, the continuous CVD SiC coating process used on HTS fiber was extended to the coating of HMS, Celion 6000, Celion 12000 and T-300 graphite fiber. Electrical resistances three order of magnitude greater than the uncoated fiber were measured with no significant degradation of the fiber strength. Graphite fibers coated with CVD Si3N4 and BN had resistances greater than 10(exp 6) ohm/cm. Lower pyrolysis temperatures were used in preparing the silica-like coatings also resulting in resistances as high as three orders of magnitude higher than the uncoated fiber. The epoxy matrix composites prepared using these coated fibers had low shear strengths indicating that the coatings were weak.

  15. Effect of processing pressure on the properties of graphite foam

    NASA Astrophysics Data System (ADS)

    Shah, Raviraj; Soni, Neha; Shrivastava, R.

    2013-06-01

    Graphite foam samples were prepared by heating mesophase pitch at different processing pressures followed by carbonization and graphitization under inert atmosphere. These samples were characterized for density, surface morphology and thermal conductivity. Microstructure of the samples indicate that processing pressure controls the evolution of volatiles from mesophase pitch to create a structure having optimum pore size, ligament thickness and junctions, which all are responsible for physical and thermal properties of resultant graphite foam. The study reveals that there is a relationship between processing pressure and the final density & thermal conductivity.

  16. Deconstructing graphite: graphenide solutions.

    PubMed

    Pénicaud, Alain; Drummond, Carlos

    2013-01-15

    Growing interest in graphene over past few years has prompted researchers to find new routes for producing this material other than mechanical exfoliation or growth from silicon carbide. Chemical vapor deposition on metallic substrates now allows researchers to produce continuous graphene films over large areas. In parallel, researchers will need liquid, large scale, formulations of graphene to produce functional graphene materials that take advantage of graphene's mechanical, electrical, and barrier properties. In this Account, we describe methods for creating graphene solutions from graphite. Graphite provides a cheap source of carbon, but graphite is insoluble. With extensive sonication, it can be dispersed in organic solvents or water with adequate additives. Nevertheless, this process usually creates cracks and defects in the graphite. On the other hand, graphite intercalation compounds (GICs) provide a means to dissolve rather than disperse graphite. GICS can be obtained through the reaction of alkali metals with graphite. These compounds are a source of graphenide salts and also serve as an excellent electronic model of graphene due to the decoupling between graphene layers. The graphenide macroions, negatively charged graphene sheets, form supple two-dimensional polyelectrolytes that spontaneously dissolve in some organic solvents. The entropic gain from the dissolution of counterions and the increased degrees of freedom of graphene in solution drives this process. Notably, we can obtain graphenide solutions in easily processable solvents with low boiling points such as tetrahydrofuran or cyclopentylmethylether. We performed a statistical analysis of high resolution transmission electronic micrographs of graphene sheets deposited on grids from GICs solution to show that the dissolved material has been fully exfoliated. The thickness distribution peaks with single layers and includes a few double- or triple-layer objects. Light scattering analysis of the

  17. Magnetic frustration of graphite oxide

    NASA Astrophysics Data System (ADS)

    Lee, Dongwook; Seo, Jiwon

    2017-03-01

    Delocalized π electrons in aromatic ring structures generally induce diamagnetism. In graphite oxide, however, π electrons develop ferromagnetism due to the unique structure of the material. The π electrons are only mobile in the graphitic regions of graphite oxide, which are dispersed and surrounded by sp3-hybridized carbon atoms. The spin-glass behavior of graphite oxide is corroborated by the frequency dependence of its AC susceptibility. The magnetic susceptibility data exhibit a negative Curie temperature, field irreversibility, and slow relaxation. The overall results indicate that magnetic moments in graphite oxide slowly interact and develop magnetic frustration.

  18. Magnetic frustration of graphite oxide

    PubMed Central

    Lee, Dongwook; Seo, Jiwon

    2017-01-01

    Delocalized π electrons in aromatic ring structures generally induce diamagnetism. In graphite oxide, however, π electrons develop ferromagnetism due to the unique structure of the material. The π electrons are only mobile in the graphitic regions of graphite oxide, which are dispersed and surrounded by sp3-hybridized carbon atoms. The spin-glass behavior of graphite oxide is corroborated by the frequency dependence of its AC susceptibility. The magnetic susceptibility data exhibit a negative Curie temperature, field irreversibility, and slow relaxation. The overall results indicate that magnetic moments in graphite oxide slowly interact and develop magnetic frustration. PMID:28327606

  19. Fast Reactors

    NASA Astrophysics Data System (ADS)

    Esposito, S.; Pisanti, O.

    The following sections are included: * Elementary Considerations * The Integral Equation to the Neutron Distribution * The Critical Size for a Fast Reactor * Supercritical Reactors * Problems and Exercises

  20. MEANS FOR COOLING REACTORS

    DOEpatents

    Wheeler, J.A.

    1957-11-01

    A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.