Science.gov

Sample records for graphite reactor physics

  1. Change in physical properties of high density isotropic graphites irradiated in the ?JOYO? fast reactor

    NASA Astrophysics Data System (ADS)

    Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

    1995-08-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor "JOYO" to fluences from 2.11 to 2.86 × 10 26 n/m 2 ( E > 0.1 MeV) at temperatures from 549 to 597°C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens.

  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. FSV experience in support of the GT-MHR reactor physics, fuel performance, and graphite

    SciTech Connect

    Baxter, A.M.; McEachern, D.; Hanson, D.L.; Vollman, R.E.

    1994-11-01

    The Fort St. Vrain (FSV) power plant was the most recent operating graphite-moderated, helium-cooled nuclear power plant in the United States. Many similarities exist between the FSV design and the current design of the GT-MHR. Both designs use graphite as the basic building blocks of the core, as structural material, in the reflectors, and as a neutron moderator. Both designs use hexagonal fuel elements containing cylindrical fuel rods with coated fuel particles. Helium is the coolant and the power densities vary by less than 5%. Since material and geometric properties of the GT-MHR core am very similar to the FSV core, it is logical to draw upon the FSV experience in support of the GT-MHR design. In the Physics area, testing at FSV during the first three cycles of operation has confirmed that the calculational models used for the core design were very successful in predicting the core nuclear performance from initial cold criticality through power operation and refueling. There was excellent agreement between predicted and measured initial core criticality and control rod positions during startup. Measured axial flux distributions were within 5% of the predicted value at the peak. The isothermal temperature coefficient at zero power was in agreement within 3%, and even the calculated temperature defect over the whole operating range for cycle 3 was within 8% of the measured defect. In the Fuel Performance area, fuel particle coating performance, and fission gas release predictions and an overall plateout analysis were performed for decommissioning purposes. A comparison between predicted and measured fission gas release histories of Kr-85m and Xe-138 and a similar comparison with specific circulator plateout data indicated good agreement between prediction and measured data. Only I-131 plateout data was overpredicted, while Cs-137 data was underpredicted.

  5. 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)

  6. 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.

  7. 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).

  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. Graphite stored energy in the UCLA research reactor

    SciTech Connect

    Ashbaugh, C.E.; Ostrander, N.C.; Pearlman, H.

    1986-01-01

    One of several reactors of similar basic design built in the same time period, the UCLA reactor entered the relicensing process in 1980. A US Nuclear Regulatory Commission-sponsored generic safety analysis of such reactors included a brief evaluation of the energy stored in the graphite (Wigner Energy), and concluded that this was negligibly small. Shutdown of the UCLA reactor facility in 1984 provided an opportunity to measure the stored energy. Samples of graphite were taken at the following locations: immediately adjacent to the boxes; at the center of the graphite island (where the thermal flux peaks); and also from a stringer approx. 2 in. from the reactor core center. All samples were from nearly the same horizontal plane, at about mid-height of the core. Stored energy was measured by differential thermal calorimetry, on a Du Pont Thermal Analyzer Model 1090, with scanning temperatures up to 550/sup 0/C. The highest value found was 33.2 cal/g, next to the fuel boxes. At the island center, it was 19.2 cal/g. The stored energy is small, and further is confined to the graphite volume adjacent to the fuel boxes, which is a small fraction of the total volume of graphite in the reactor. The potential hazard from release of graphite stored energy is negligible.

  10. Nuclear Reactor Physics

    NASA Astrophysics Data System (ADS)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  11. 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.

  12. 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.

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

    SciTech Connect

    Grover, Stanley Blaine; Metcalfe, M. P.

    2002-04-01

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

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

    SciTech Connect

    Grover, S.B.; Metcalfe, M.P.

    2002-03-14

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

  15. 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.

  16. 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.

  17. 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

  18. 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...

  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. 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.

  1. 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.

  2. 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...

  3. 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...

  4. 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)

  5. 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.

  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. 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.

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

    NASA Astrophysics Data System (ADS)

    Rigali, Mark J.; Nagy, Bartholomew

    1997-01-01

    The presence, concentration, and distribution of organic free radicals as well as their association with specific surface areas and microporosities help characterize the evolution and behavior of the Oklo carbonaceous matter. Such information is necessary in order to evaluate uranium mineralization, liquid bitumen solidification, and radionuclide containment at Oklo. In the Oklo ore deposits and natural fission reactors carbonaceous matter is often referred to as solid graphitic bitumen. The carbonaceous parts of the natural reactors may contain as much as 65.9% organic C by weight in heterogeneous distribution within the clay-rich matrix. The solid carbonaceous matter immobilized small uraninite crystals and some fission products enclosed in this uraninite and thereby facilitated radionuclide containment in the reactors. Hence, the Oklo natural fission reactors are currently the subjects of detailed studies because they may be useful analogues to support performance assessment of radionuclide containment at anthropogenic radioactive waste repository sites. Seven carbonaceous matter rich samples from the 1968 ± 50 Ma old natural fission reactors and the associated Oklo uranium ore deposit were studied by electron spin resonance (ESR) spectroscopy and by measurements of specific surface areas (BET method). Humic acid, fulvic acid, and fully crystalline graphite standards were also examined by ESR spectroscopy for comparison with the Oklo solid graphitic bitumens. With one exception, the ancient Oklo bitumens have higher organic free radical concentrations than the modern humic and fulvic acid samples. The presence of carbon free radicals in the graphite standard could not be determined due to the conductivity of this material. Oklo solid bitumen samples were subjected to various pressures of O 2, a paramagnetic gas. O 2-organic free radical interactions, as revealed by ESR spectral line broadening, indicate that the organic free radicals of the Oklo solid bitumens

  9. 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

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. 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. PMID:16381692

  15. 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.

  16. Evaluation of surface deposits on the channel wall of trepanned reactor core graphite samples

    NASA Astrophysics Data System (ADS)

    Heard, P. J.; Payne, L.; Wootton, M. R.; Flewitt, P. E. J.

    2014-02-01

    Samples have been trepanned from the fuel and interstitial channel walls of PGA graphite reactor cores of two Magnox gas cooled power stations after a period of service. These samples have been considered explicitly for the presence of deposits on the channel facing surfaces. A combination of focused ion beam milling and imaging has been used to determine the presence of such deposits and where present to make measurements of the thickness. These thicknesses vary from a few nanometres to tens of micrometres. In addition, both the chemical composition and chemical state have been investigated using energy dispersive X-ray microanalysis in a scanning electron microscope and Raman spectroscopy respectively. EDX measurements showed that surface deposits found on the channel walls of one of the reactors contained increased concentrations of oxygen, iron, chromium and sulphur compared with the underlying material. Raman spectroscopy also suggested that the deposit had a smaller crystallite size than PGA graphite.

  17. Experimental plan and design of two experiments for graphite irradiation at temperatures up to 1500 °C in the target region of the high flux isotope reactor

    NASA Astrophysics Data System (ADS)

    McDuffee, J. L.; Burchell, T. D.; Heatherly, D. W.; Thoms, K. R.

    2008-10-01

    Two irradiation capsules have been designed for the target region of the high flux isotope reactor (HFIR). The objective is to provide dimensional change and physical property data for four candidate next generation nuclear plant (NGNP) graphites. The capsules will reach peak doses of ˜1.59 and ˜4.76 dpa, respectively, at temperatures of 900, 1200, and 1500 °C.

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. Ultra-thin Graphitic Film: Synthesis and Physical Properties

    NASA Astrophysics Data System (ADS)

    Kaplas, Tommi; Kuzhir, Polina

    2016-02-01

    A scalable technique of chemical vapor deposition (CVD) growth of ultra-thin graphitic film is proposed. Ultra-thin graphitic films grown by a one-step CVD process on catalytic copper substrate have higher crystallinity than pyrolytic carbon grown on a non-catalytic surface and appear to be more robust than a graphene monolayer. The obtained graphitic material, not thicker than 8 nm, survives during the transfer process from a Cu substrate without a template polymer layer, typically used in the graphene transfer process to protect graphene. This makes the transfer process much more simple and cost-effective. Having electrical and optical properties compatible with what was observed for a few layers of CVD graphene, the proposed ultra-thin graphitic film offers new avenues for implementing 2D materials in real-world devices.

  4. 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

  5. Oxidation Damage Evaluation by Non-Destructive Method for Graphite Components in High Temperature Gas-Cooled Reactor

    NASA Astrophysics Data System (ADS)

    Shibata, Taiju; Tada, Tatsuya; Sumita, Junya; Sawa, Kazuhiro

    To develop non-destructive evaluation methods for oxidation damage on graphite components in High Temperature Gas-cooled Reactors (HTGRs), the applicability of ultrasonic wave and micro-indentation methods were investigated. Candidate graphites, IG-110 and IG-430, for core components of Very High Temperature Reactor (VHTR) were used in this study. These graphites were oxidized uniformly by air at 500 °C. The following results were obtained from this study. (1) Ultrasonic wave velocities with 1 MHz can be expressed empirically by exponential formulas to burn-off, oxidation weight loss. (2) The porous condition of the oxidized graphite could be evaluated with wave propagation analysis with a wave-pore interaction model. It is important to consider the non-uniformity of oxidized porous condition. (3) Micro-indentation method is expected to determine the local oxidation damage. It is necessary to assess the variation of the test data.

  6. 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

  7. 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.

  8. 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

  9. 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.

  10. 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.

  11. 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.

  12. 2012 CHEMISTRY & PHYSICS OF GRAPHITIC CARBON MATERIALS GORDON RESEARCH CONFERENCE, JUNE 17-22, 2012

    SciTech Connect

    Fertig, Herbert

    2012-06-22

    This conference will highlight the urgency for research on graphitic carbon materials and gather scientists in physics, chemistry, and engineering to tackle the challenges in this field. The conference will focus on scalable synthesis, characterization, novel physical and electronic properties, structure-properties relationship studies, and new applications of the carbon materials. Contributors

  13. 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.

  14. 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.

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

    SciTech Connect

    Rigali, M.J.; Nagy, B.

    1997-01-01

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

  16. Physical sputtering and chemical erosion studies on plain and lithiated graphite samples

    NASA Astrophysics Data System (ADS)

    Raju, Ramasamy; Racic, Marin; Lee, J.; Ruzic, David

    2007-11-01

    PFC candidate materials must have characteristics allowing for high temperature resilience while limiting deuterium recycling and core contamination from erosion. Graphite is a good choice of material for its high temperature tolerance. However, to reduce deuterium recycling issues of the graphite surface, lithium has been used extensively as a coating on PFC surfaces, though many issues on physical and chemical sputtering still remain. The Ion-surface InterAction Experiment (IIAX) measures the absolute, angular-resolved and self-sputtering yields of many particle/target combinations. Baseline sputtering yield of an untreated ATJ graphite sample is very close to the predicted TRIM estimates with an average of 0.06 +/- 0.02 atoms / ion. Preliminary experiments show that Li was evaporated and deposited with thickness of 320 nm on a Si wafer. Li deposition on a ATJ graphite sample was verified using scanning electron microscopy. Chemical sputtering analysis on a ATJ graphite sample is done, and results confirm the operation of the RGA. Trail experiments on relative levels of Li to C collected during sputtering are analyzed using TOF-SIMS. A deposition rate of 10 nm/min is the most relevant to NSTX. Additional experiments using varying thicknesses and deposition rates of Li are described.

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. Preparation of pyrolytic carbon coating on graphite for inhibiting liquid fluoride salt and Xe135 penetration for molten salt breeder reactor

    NASA Astrophysics Data System (ADS)

    Song, Jinliang; Zhao, Yanling; He, Xiujie; Zhang, Baoliang; Xu, Li; He, Zhoutong; Zhang, DongSheng; Gao, Lina; Xia, Huihao; Zhou, Xingtai; Huai, Ping; Bai, Shuo

    2015-01-01

    A fixed-bed deposition method was used to prepare rough laminar pyrolytic carbon coating (RLPyC) on graphite for inhibiting liquid fluoride salt and Xe135 penetration during use in molten salt breeder reactor. The RLPyC coating possessed a graphitization degree of 44% and had good contact with graphite substrate. A high-pressure reactor was constructed to evaluate the molten salt infiltration in the isostatic graphite (IG-110, TOYO TANSO CO., LTD.) and RLPyC coated graphite under 1.01, 1.52, 3.04, 5.07 and 10.13 × 105 Pa for 12 h. Mercury injection and molten-salt infiltration experiments indicated the porosity and the salt-infiltration amount of 18.4% and 13.5 wt% under 1.52 × 105 Pa of IG-110, which was much less than 1.2% and 0.06 wt% under 10.13 × 105 Pa of the RLPyC, respectively. A vacuum device was constructed to evaluate the Xe135 penetration in the graphite. The helium diffusion coefficient of RLPyC coated graphite was 2.16 × 10-12 m2/s, much less than 1.21 × 10-6 m2/s of the graphite. Thermal cycle experiment indicated the coatings possessed excellent thermal stability. The coated graphite could effectively inhibit the liquid fluoride salt and Xe135 penetration.

  4. 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.

  5. Evaluated Neutron Nuclear Data for Reactor Physics Calculations.

    1988-09-15

    Version 00 The data file KEDAK contains the evaluated neutron nuclear data for a number of materials important for the reactor physics, specific physical experiments, burn up calculations, shielding and other applications.

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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

  11. 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.

  12. 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.

  13. 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

  14. Neutrino physics with accelerator driven subcritical reactors

    NASA Astrophysics Data System (ADS)

    Ciuffoli, Emilio; Evslin, Jarah; Zhao, Fengyi

    2016-01-01

    Accelerator driven system (ADS) subcritical nuclear reactors are under development around the world. They will be intense sources of free, 30-55 MeV μ + decay at rest {overline{ν}}_{μ } . These ADS reactor neutrinos can provide a robust test of the LSND anomaly and a precise measurement of the leptonic CP-violating phase δ, including sign(cos(δ)). The first phase of many ADS programs includes the construction of a low energy, high intensity proton or deuteron accelerator, which can yield competitive bounds on sterile neutrinos.

  15. Hydrogen isotope trapping on graphite collectors during an isotope exchange experiment in the tokomak fusion test reactor

    SciTech Connect

    Kilpatrick, S.J.; Nygren, R.; Wampler, W.R.; Ulrickson, M.; Dylla, H.F.; Manos, D.M.; Ramsey, A.T.; Hirooka, Y.

    1988-01-01

    A rotatable collector probe was used to expose several graphite samples to a deuterium-to-hydrogen-to-deuterium exchange experiment in the Tokamak Fusion Test Reactor (TFTR) at the start of the 1988 operations period. This experiment proved the utility of helium conditioning discharges in accelerating the changeover process. Samples included portions of a tile taken from the inner bumper limiter (POCO AXF-5Q graphite) of TFTR during the recent machine opening, and coupons which had been conditioned in the Plasma Surface Interaction Experimental Facility (PISCES) by exposure to a helium plasma. The samples were exposed to different groups of the /approximately/100 1.4MA discharges that comprised the experiment. They were removed and analyzed for retained deuterium and impurities by nuclear reaction analysis and Rutherford backscattering spectrometry. Codeposited carbon layers had been formed with thicknesses up to several tenths of a micron. The inferred percentages of trapped hydrogenic species were in general agreement with spectroscopic data. The computed carbon fluence per D discharge, 1.2 /times/ 10 X C/cmS, is compared to recent measurements on limiter tiles removed from TFTR. 21 refs., 3 figs., 1 tab.

  16. 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.

  17. 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

  18. METHODS AND RESULTS OF RECONSTRUCTION OF NOBLE GAS RELEASES FROM THE STACKS OF THE MAYAK PA GRAPHITE REACTORS OVER THE WHOLE PERIOD OF THEIR OPERATION

    SciTech Connect

    Glagolenko, Y. V.; Drozhko, Evgeniy G.; Mokrov, Y.; Pyatin, N. P.; Rovny, Sergey I.; Anspaugh, L. R.; Napier, Bruce A.

    2008-06-01

    Brief analysis of design features and operational modes of Mayak PA industrial graphite-uranium reactors (PUGRs) is given. The above mentioned Mayak PA PUGRs determined the rates of releases of radioactive noble gases (RNG) from activation (41Ar) and fission (isotopes of Krypton and Xenon) through the vent stack of the reactor. Information is given on methods and results of experimental determination of RNG atmospheric releases for the period starting from 1965 till PUGRs decommissioning in 1987-1990. A calculation method for reconstruction of radioactive noble gas releases is proposed and justified. The results of reconstruction are given. It is shown that maximum rates of RNG releases from PUGRs high stacks were observed in the 1950s, when ordinary atmospheric air was used as a cover gas for the reactor graphite stacks and gas purification systems (flow-type gas holders) had not been installed yet.

  19. Internal and external atomic steps in graphite exhibit dramatically different physical and chemical properties.

    PubMed

    Lee, Hyunsoo; Lee, Han-Bo-Ram; Kwon, Sangku; Salmeron, Miquel; Park, Jeong Young

    2015-04-28

    We report on the physical and chemical properties of atomic steps on the surface of highly oriented pyrolytic graphite (HOPG) investigated using atomic force microscopy. Two types of step edges are identified: internal (formed during crystal growth) and external (formed by mechanical cleavage of bulk HOPG). The external steps exhibit higher friction than the internal steps due to the broken bonds of the exposed edge C atoms, while carbon atoms in the internal steps are not exposed. The reactivity of the atomic steps is manifested in a variety of ways, including the preferential attachment of Pt nanoparticles deposited on HOPG when using atomic layer deposition and KOH clusters formed during drop casting from aqueous solutions. These phenomena imply that only external atomic steps can be used for selective electrodeposition for nanoscale electronic devices.

  20. Thermodynamic study of krypton adsorbed on graphite using a statistical physics treatment

    NASA Astrophysics Data System (ADS)

    Ben Yahia, M.; Aouaini, F.; Hachicha, M. A.; Knani, S.; Khalfaoui, M.; Ben Lamine, A.

    2013-06-01

    New theoretical expressions for the adsorption isotherm of krypton on exfoliated graphite at 77.3 K were established. The establishment of these new expressions is based on the statistical physics formalism and some working hypotheses. This method allowed the estimation of the physico-chemical parameters in the theoretical models. The parameters intervening in the adsorption process were deduced directly from the experimental adsorption isotherm by numerical simulation. Our objective in this modeling approach is to select an adequate model that presents a high correlation with the experimental curve. We introduce three parameters affecting the adsorption process, namely, the density of krypton receptor sites NM, the number of molecules per site and the krypton adsorption energy. Then, we apply the appropriate model to calculate the thermodynamics functions, such as the entropy, the free enthalpy and the internal energy, that govern the adsorption mechanism.

  1. High thermal conductivity of graphite fiber silicon carbide composites for fusion reactor application

    NASA Astrophysics Data System (ADS)

    Snead, L. L.; Balden, M.; Causey, R. A.; Atsumi, H.

    2002-12-01

    The benefits of using CVI SiC/graphite fiber composites as low tritium retaining, high thermal conductivity composites for fusion applications are presented. Three-dimensional woven composites have been chemically vapor infiltrated with SiC and their thermophysical properties measured. One material used an intermediate grade graphite fiber in all directions (Amoco P55) while a second material used very high thermal conductive fiber (Amoco K-1100) in the high fiber density direction. The overall void was less than 20%. Strength as measured by four-point bending was comparable to those of SiC/SiC composite. The room temperature thermal conductivity in the high conductivity direction was impressive for both materials, with values >70 W/m K for the P-55 and >420 W/m K for the K-1100 variant. The thermal conductivity was measured as a function of temperature and exceeds the highest thermal conductivity of CVD SiC currently available at fusion relevant temperatures (>600 °C). Limited data on the irradiation-induced degradation in thermal conductivity is consistent with carbon fiber composite literature.

  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. 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

  4. 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.

  5. Scalable parallel solution coupling for multi-physics reactor simulation.

    SciTech Connect

    Tautges, T. J.; Caceres, A.; Mathematics and Computer Science

    2009-01-01

    Reactor simulation depends on the coupled solution of various physics types, including neutronics, thermal/hydraulics, and structural mechanics. This paper describes the formulation and implementation of a parallel solution coupling capability being developed for reactor simulation. The coupling process consists of mesh and coupler initialization, point location, field interpolation, and field normalization. We report here our test of this capability on an example problem, namely, a reflector assembly from an advanced burner test reactor. Performance of this coupler in parallel is reasonable for the chosen problem size and range of processor counts. The runtime is dominated by startup costs, which amortize over the entire coupled simulation. Future efforts will include adding more sophisticated interpolation and normalization methods, to accommodate different numerical solvers used in various physics modules and to obtain better conservation properties for certain field types.

  6. 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.

  7. 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.

  8. 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

  9. The International Reactor Physics Experiment Evaluation Project (IRPHEP)

    SciTech Connect

    J. Blair Briggs; Enrico Sartori; Lori Scott

    2006-09-01

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

  10. 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.

  11. 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.

  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. 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

  14. 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.

  15. TAS measurements for reactor physics and nuclear structure

    NASA Astrophysics Data System (ADS)

    Algora, A.; Jordan, D.; Taín, J. L.; Rubio, B.; Agramunt, J.; Caballero, L.; Nácher, E.; Perez-Cerdan, A. B.; Molina, F.; Estevez, E.; Krasznahorkay, A.; Hunyadi, M. D.; Gulyás, J.; Vitéz, A.; Csatlós, M.; Csige, L.; ńysto, J.; Penttilä, H.; Rinta-Antila, S.; Moore, I.; Eronen, T.; Jokinen, A.; Nieminen, A.; Hakala, J.; Karvonen, P.; Kankainen, A.; Hager, U.; Sonoda, T.; Saastamoinen, A.; Rissanen, J.; Kessler, T.; Weber, C.; Ronkainen, J.; Rahaman, S.; Elomaa, V.; Burkard, K.; Hüller, W.; Batist, L.; Gelletly, W.; Nichols, A. L.; Yoshida, T.; Sonzogni, A. A.; Peräjärvi, K.

    2011-10-01

    In this contribution we will present recent total absorption measurements of the beta decay of neutron-rich nuclei performed at the IGISOL facility of the Univ. of Jyväskyla. In the measurements the JYFL Penning Trap was used as a high resolution isobaric separator. The total absorption technique will be described and the impact of recent results in the fields of reactor physics (decay heat calculations) and nuclear structure will be discussed.

  16. 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.

  17. Thermal neutron scattering in graphite

    NASA Astrophysics Data System (ADS)

    Al-Qasir, Iyad Ibrahim

    Generation IV Very High Temperature Reactor (VHTR) concepts, are graphite moderated and gas cooled thermal spectrum reactors. The characteristics of the low energy (E < 1 eV) neutron spectrum in these reactors will be dictated by the process of neutron slowing-down and thermalization in the graphite moderator. The ability to accurately predict this process in these reactors can have significant neutronic and safety implications. In reactor design calculations, thermal neutron scattering cross section libraries are needed for the prediction of the thermal neutron environment in the core. Currently used libraries (ENDF/B-VII) are a product of the 1960s and remain based on many physical approximations. In addition, these libraries show noticeable discrepancies with experimental data. In this work, investigation of thermal neutron scattering in graphite as a function of temperature was performed. The fundamental input for the calculation of thermal neutron scattering cross sections, i.e., the phonon frequency distribution and/or the dispersion relations, was generated using a modern approach that is based on quantum mechanical electronic structure (ab initio) simulations combined with a lattice dynamics direct method supercell approach. The calculations were performed using the VASP and PHONON codes. The VASP calculations used the local density approximation, and the projector augmented-wave pseudopotential. A supercell of 144 atoms was used; and the integration over the Brillouin zone was confined to a 3x3x4 k-mesh generated by the Monkhorst-Pack scheme. A plane-wave basis set with an energy cutoff of 500 eV was applied. The corresponding dispersion relations, heat capacity, and phonon frequency distribution show excellent agreement with experimental data. Despite the use of the above techniques to produce more accurate input data, the examination of the results indicated persistence of the inconsistencies between calculations and measurements at neutron energies

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

    SciTech Connect

    Hung, C.C.; Long, M.; Stahl, M.

    1987-08-01

    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.

  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. 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.

  1. 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.

  2. 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).

  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. 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.

  5. 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)

  6. 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.

  7. 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.

  8. 75 FR 62695 - Physical Protection of Irradiated Reactor Fuel in Transit

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-13

    ... COMMISSION 10 CFR Part 73 RIN 3150-AI64 Physical Protection of Irradiated Reactor Fuel in Transit AGENCY...) is proposing to amend its security regulations pertaining to the transport of irradiated reactor fuel (for purposes of this rulemaking, the terms ``irradiated reactor fuel'' and ``spent nuclear fuel''...

  9. 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.

  10. Experimental investigation on the volume reduction of irradiated graphite arising from the decommissioning of KRR-2

    SciTech Connect

    Lee, D.G.; Cho, Y.J.; Yang, H.C.; Lee, K.W.; Jung, C.H.

    2007-07-01

    Graphite has been used as a moderator and reflector of neutrons in more than 100 nuclear power plants as well as many experimental reactors and plutonium production reactors in various countries. Most of the older graphite moderated reactors are already shut down and arc awaiting decommissioning planning and preparation. The graphite waste has different characteristics than other decommissioning waste due to its physical and chemical properties and also because of the presence of tritium and carbon-14. Therefore radioactive graphite dismantling, handling, conditioning and disposal are a common part of the decommissioning activities. A volume reduction of the waste is needed to reduce disposal cost of radioactive waste. However the existing processing technologies are based mostly on the isolation of radioactive graphite from the environment, they are not able to provide for a significant volume reduction. For this reason, the high-temperature thermal treatment process such as an incineration or a pyrolysis is considered as promising technologies, since it provides a substantial volume reduction. Currently, the fluidized bed incineration is considered as efficient technology for the treatment of radioactive graphite waste. In this study, the fluidized bed incineration condition and the radioisotopes behavior were experimentally investigated by using irradiated graphite waste which has arisen from the decommissioning of Korean Research Reactor 2 (KRR-2). (authors)

  11. Conceptual study of fusion-driven transmutation reactor with ITER physics and engineering constraints

    NASA Astrophysics Data System (ADS)

    Hong, Bong

    2011-10-01

    A conceptual study of fusion-driven transmutation reactor was performed based on ITER physics and engineering constraints. A compact reactor concept is desirable from an economic viewpoint. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burnup calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor.

  12. 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

  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. Recent improvements of reactor physics codes in MHI

    NASA Astrophysics Data System (ADS)

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

    2015-12-01

    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.

  15. 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.

  16. 78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-30

    ... COMMISSION Proposed Revision to Physical Security--Standard Design Certification and Operating Reactors...: Section 13.6.2 ``Physical Security--Standard Design Certification and Operating Reactors.'' The NRC seeks... security reviews of design certification applications. DATES: Comments must be filed no later than...

  17. 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

  18. 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.

  19. Hybrid Reduced Order Modeling Algorithms for Reactor Physics Calculations

    NASA Astrophysics Data System (ADS)

    Bang, Youngsuk

    hybrid ROM algorithms which can be readily integrated into existing methods and offer higher computational efficiency and defendable accuracy of the reduced models. For example, the snapshots ROM algorithm is hybridized with the range finding algorithm to render reduction in the state space, e.g. the flux in reactor calculations. In another implementation, the perturbation theory used to calculate first order derivatives of responses with respect to parameters is hybridized with a forward sensitivity analysis approach to render reduction in the parameter space. Reduction at the state and parameter spaces can be combined to render further reduction at the interface between different physics codes in a multi-physics model with the accuracy quantified in a similar manner to the single physics case. Although the proposed algorithms are generic in nature, we focus here on radiation transport models used in support of the design and analysis of nuclear reactor cores. In particular, we focus on replacing the traditional assembly calculations by ROM models to facilitate the generation of homogenized cross-sections for downstream core calculations. The implication is that assembly calculations could be done instantaneously therefore precluding the need for the expensive evaluation of the few-group cross-sections for all possible core conditions. Given the generic natures of the algorithms, we make an effort to introduce the material in a general form to allow non-nuclear engineers to benefit from this work.

  20. Graphite design handbook

    SciTech Connect

    Ho, F.H.

    1988-09-01

    The objectives of the Graphite Design Handbook (GDH) are to provide and maintain a single source of graphite properties and phenomenological model of mechanical behavior to be used for design of MHTGR graphite components of the Reactor System, namely, core support, permanent side reflector, hexagonal reflector elements, and prismatic fuel elements; to provide a single source of data and material models for use in MHTGR graphite component design, performance, and safety analyses; to present properties and equations representing material models in a form which can be directly used by the designer or analyst without the need for interpretation and is compatible with analytical methods and structural criteria used in the MHTGR project, and to control the properties and material models used in the MHTGR design and analysis to proper Quality Assurance standards and project requirements. The reference graphite in the reactor internal components is the nuclear grade 2020. There are two subgrades of interest, the cylinder nuclear grade and the large rectangular nuclear grade. The large rectangular nuclear grade is molded in large rectangular blocks. It is the reference material for the permanent side reflector and the central column support structure. The cylindrical nuclear grade is isostatically pressed and is intended for use as the core support component. This report gives the design properties for both H-451 and 2020 graphite as they apply to their respective criteria. The properties are presented in a form for design, performance, and safety calculations that define or validate the component design. 103 refs., 20 figs., 19 tabs.

  1. 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.

  2. Simple Coupling of Reactor Physics Effects and Uncertain Nuances

    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 thatmore » 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.« less

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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. PMID:25197659

  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

    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.

  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 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

  10. 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.

  11. 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.

  12. Effects of compositions of filler, binder and porosity on elastic and fracture properties of nuclear graphite

    NASA Astrophysics Data System (ADS)

    Kyaw, S. T.; Sun, W.; Becker, A. A.

    2015-02-01

    Physical mechanisms at different length scales have to be taken into account while predicting the overall failure of nuclear graphite structures of advanced gas cooled graphite reactors. In this paper, the effect of composition of meso graphite phases and porosity on the aggregate elastic properties is predicted using the Eshelby homogenisation method. Results indicate an overall decrease in elastic modulus with an increase in porosity. Subsequently, the moduli at different porosity levels are used to predict the critical strain energy release rates for crack propagation of graphite, and fracture behaviour is studied using compact tension and four point bending tests. Compared to flexural strength at zero porosity level, significant reduction in strength of up to 80% at 30% porosity level is observed. Evolution of flexural strength due to porosity is also compared against available experimental values of graphite from UK nuclear plants.

  13. 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.

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

    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 iterativemore » 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.« less

  15. 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.

  16. 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...

  17. Research on acceleration method of reactor physics based on FPGA platforms

    SciTech Connect

    Li, C.; Yu, G.; Wang, K.

    2013-07-01

    The physical designs of the new concept reactors which have complex structure, various materials and neutronic energy spectrum, have greatly improved the requirements to the calculation methods and the corresponding computing hardware. Along with the widely used parallel algorithm, heterogeneous platforms architecture has been introduced into numerical computations in reactor physics. Because of the natural parallel characteristics, the CPU-FPGA architecture is often used to accelerate numerical computation. This paper studies the application and features of this kind of heterogeneous platforms used in numerical calculation of reactor physics through practical examples. After the designed neutron diffusion module based on CPU-FPGA architecture achieves a 11.2 speed up factor, it is proved to be feasible to apply this kind of heterogeneous platform into reactor physics. (authors)

  18. The ignition physics study group supports the compact ignition tokamak and engineering test reactor programs

    SciTech Connect

    Sheffield, J.

    1987-01-01

    This report presents a collection of Vugraphs dealing with the Compact Ignition Tokamak (CIT) and the Engineering Test Reactor (ETR). The role of the Ignition Physics Study Group is defined. Several design goals are presented. (JDH)

  19. 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.

  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. 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.

  2. 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.

  3. 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

    ... 10 Energy 2 2014-01-01 2014-01-01 false Requirements for physical protection of irradiated reactor... Transit § 73.35 Requirements for physical protection of irradiated reactor fuel (100 grams or less) in... quantity of irradiated reactor fuel weighing 100 grams (0.22 pounds) or less in net weight of...

  4. 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.

  5. 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.

  6. Physics Characterization of a Heterogeneous Sodium Fast Reactor Transmutation System

    SciTech Connect

    Samuel E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even mass number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both a non-flattened and a pancake core geometry. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of the same size.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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

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

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 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... achieve the following objectives: 1 For purposes of 10 CFR 73.37, the terms “irradiated reactor fuel”...

  13. 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 OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.60 Additional... nuclear material from theft or diversion pursuant to the requirements of paragraphs 73.67 (a), (b),...

  14. 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

  15. 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

    2014-08-01

    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

  16. 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

    2014-08-01

    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.

  17. 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.

  18. The impact of improved physics on commercial tokamak reactors

    SciTech Connect

    Galambos, J.D.; Perkins, L.J.; Haney, S.; Mandrekas, J.

    1994-01-01

    Improvements in the confinement and beta capability of tokamak devices have long been a goal of the fusion program. We examine the impact of improvements in present day confinement and beta capabilities on commercial tokamak reactors. We characterize confinement with the achievable enhancement factor (H) over the ITER89 Power scaling confinement time, and beta by the Troyon coefficient g. A surprisingly narrow range of plasma confinement and beta are found to be useful in minimizing the cost of electricity for a tokamak reactor. Improvements in only one of these quantities is not useful beyond some point, without accompanying improvements in the other. For the plasma beta limited by a Troyon coefficient (g) near 4.3 (%mT/MA), confinement levels characterized by H factor enhancements of only 2 are useful for our nominal steady-state driven tokamak. These confinement levels are similar to those observed in present day experiments. If the permissible Troyon beta coefficient is near 6, the useful H factor confinement range increases to 2.5, still close to present day confinement levels. Inductively driven, pulsed reactors have somewhat increased useful ranges of confinement, relative to the steady-state cases. For a Troyon beta limit coefficient g near 4.3, H factors up to 2.5 are useful, and for g near 6, H factors up to 3 are useful.

  19. Physical and engineering constraints for tokamak reactors with helical coils

    SciTech Connect

    Potok, R.E.; Becker, H.

    1983-09-01

    The authors present an analytical and numerical analysis of a tokamak reactor with a set of helical coils added in order to eliminate plasma disruptions. The optimal helical configuration was found to be a set of continuous, l = 2 stellarator coils which are made of copper and are internal to the toroidal field coils, l being the number of poloidal field periods. (The optimization process did not include evaluation of the viability of a modular stellarator reactor). Scaling laws were developed for this optimal configuration, and a series of parametric scans are performed with varying assumptions for the forces on the helical coils and the ratio of helical coil transform to plasma transform (M). The option space available for attractive reactor designs is strongly constrained and involves large forces on the helical coils, low q plasma operation (q being the plasma safety factor), and moderately low M (3 to 5). Numerical calculations showed that M must be > 3 in order to obtain well defined flux surfaces. This is in agreement with results from the JIPP-T-II Tokomak.

  20. 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.

  1. 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.

  2. 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.

  3. Reactor Physics Assessment of the Inclusion of Unseparated Neptunium in MOX Reactor Fuel

    SciTech Connect

    Ellis, Ronald James

    2009-01-01

    Reducing the number of actinide separation streams in a spent fuel recovery process would reduce the cost and complexity of the process, and lower the quantity and numbers of solvents needed. It is more difficult and costly to separate Np and recombine it with Am-Cm prior to co-conversion than to simply co-strip it with the U-Pu-Np. Inclusion of the Np in mixed oxide (MOX) fuel for light water reactor (LWR) applications should not seriously affect the operating behavior of the reactor, nor should it pose insurmountable fuel design issues. In this work, the U, Pu, and Np from typical discharged and cooled PWR spent nuclear fuel are assumed to be used together in the preparation of MOX fuel for use in a pressurized water reactor (PWR). The reactor grade Pu isotopic vector is used in the model and the relative mass ratio of the Pu and Np content (Np/Pu mass is 0.061) from the cooled spent fuel is maintained but the overall Pu-Np MOX wt% is adjusted with respect to the U content (assumed to be at 0.25 wt% 235U enrichment) to offset reactivity and cycle length effects. The SCALE 5.1 scientific package (especially modules TRITON, NEWT, ORIGEN-S, ORIGEN-ARP) was used for the calculations presented in this paper. A typical Westinghouse 17x17 fuel assembly design was modeled at nominal PWR operating conditions. It was seen that U-Pu-Np MOX fuel with NpO2 and PuO2 representing 11.5wt% of the total MOX fuel would be similar to standard MOX fuel in which PuO2 is 9wt% of the fuel. The reactivity, isotopic composition, and neutron and ? sources, and the decay heat details for the discharged MOX fuel are presented and discussed in this paper.

  4. 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.

  5. 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)

  6. 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.

  7. Fast burner reactor benchmark results from the NEA working party on physics of plutonium recycle

    SciTech Connect

    Hill, R.N.; Wade, D.C.; Palmiotti, G.

    1995-12-01

    As part of a program proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed; fuel cycle scenarios using either PUREX/TRUEX (oxide fuel) or pyrometallurgical (metal fuel) separation technologies were specified. These benchmarks were designed to evaluate the nuclear performance and radiotoxicity impact of a transuranic-burning fast reactor system. International benchmark results are summarized in this paper; and key conclusions are highlighted.

  8. 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.

  9. 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.

  10. Sealing nuclear graphite with pyrolytic carbon

    NASA Astrophysics Data System (ADS)

    Feng, Shanglei; Xu, Li; Li, Li; Bai, Shuo; Yang, Xinmei; Zhou, Xingtai

    2013-10-01

    Pyrolytic carbon (PyC) coatings were deposited on IG-110 nuclear graphite by thermal decomposition of methane at ∼1830 °C. The PyC coatings are anisotropic and airtight enough to protect IG-110 nuclear graphite against the permeation of molten fluoride salts and the diffusion of gases. The investigations indicate that the sealing nuclear graphite with PyC coating is a promising method for its application in Molten Salt Reactor (MSR).

  11. Students' assessment of interactive distance experimentation in nuclear reactor physics laboratory education

    NASA Astrophysics Data System (ADS)

    Malkawi, Salaheddin; Al-Araidah, Omar

    2013-10-01

    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 reactors limit their accessibility to few educational programmes around the world. The concept of the Internet Reactor Laboratory (IRL) was introduced earlier as a new approach that utilises distance education in nuclear reactor physics laboratory education. This paper presents an initial assessment of the implementation of the IRL between the PULSTAR research reactor at North Carolina State University in the USA and the Department of Nuclear Engineering at Jordan University of Science and Technology (JUST) in Jordan. The IRL was implemented in teaching the Nuclear Reactor laboratory course for two semesters. Feedback from surveyed students verifies that the outcomes attained from using IRL in experimentation are comparable to that attainable from other on-campus laboratories performed by the students.

  12. 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.

  13. Physics and Control Assessment of AN 850 Mw(e) - Leu-Candu Reactor.

    NASA Astrophysics Data System (ADS)

    Barbone, Michelangelo

    The physics and control assessment of an 850 MW(e) Low Enriched Uranium CANDU Pressurized Heavy Water (LEU -CANDU-PHW) reactor constitute the major objective of this thesis. The use of Low Enriched Uranium fuel in the present CANDU nuclear power generating stations is recognized as economically beneficial due to reduced fuelling costs. The LEU fuel cycle is also recognized as a stepping stone to transit from the present CANDU-PHW once-through natural Uranium cycle to advanced cycles such as those based on Plutonium recycle, once-through Th + U-235 cycle, Thorium with Uranium recycle and net U-235 feed, Thorium with Uranium recycle and Plutonium feed. However, although the use of Low Enriched Uranium in the present CANDU-PHW reactor has economic advantages, and it would act as a technical bridge between the present cycle and advanced cycles, technical problems in different areas of reactor physics and fuel management were anticipated. The present thesis research work adresses the areas of reactor physics, fuel management, and control (in particular, the spatial control of large CANDU-PHW reactors). The main conclusions that have been drawn following these studies are as follows: (1) The Low Enriched Uranium Cycle is feasible in a CANDU-PHW reactor of present design and provided that: (a) The enrichment is kept relatively low (that is, about 1% instead of 0.711%); (b) the number of bundles to be replaced at every refuelling operation is about one-half that of the natural Uranium fuel case; (c) The channels are refuelled in the same direction as the coolant. (2) The response of an LEU-CANDU-PHW reactor to reactivity perturbation such as single- and two-channel refuelling operation, shim transient, shutdown-start-up transient with enrichment levels of 0.9% and 1.2% is essentially very similar {provided that certain conditions in (1) are respected} to that of the natural uranium reactor core case without any reactor reoptimization. The general behaviour of the reactor

  14. Reactor physics calculations for {sup 99}Mo production at the annular core research reactor

    SciTech Connect

    Parma, E.J.

    1995-12-31

    The Isotope Production and Distribution Program at the U.S. Department of Energy has designated Sandia National Laboratories (SNL) as the most appropriate facility for the production of {sup 99}Mo, a radioisotope whose daughter, {sup 99m}Tc, is used in more than 36,000 medical procedures per day in the United States and is considered to be a vital medical diagnostic and treatment tool. The isotope would be produced at SNL using the annular core research reactor (ACRR) facility and collocated hot cell facility. The {sup 99}Mo would be produced using the fission process by irradiating {open_quotes}targets{close_quotes} coated with {sup 235}U in the form of highly enriched U{sub 3}O{sub 8}. After {approximately}7 days of continuous irradiation in the ACRR, a target would be re- moved from the reactor core for processing. The isotope would be extracted by chemically precipitating the molybdenum using the {open_quotes}Cintichem{close_quotes} process and would be shipped to the various pharmaceutical companies by commercial or chartered airline.

  15. 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.

  16. 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.

  17. 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.

  18. (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.

  19. 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…

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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. PMID:25551649

  5. 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.

  6. 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)

  7. Reactor physics analyses of the advanced neutron source three-element core

    SciTech Connect

    Gehin, J.C.

    1995-08-01

    A reactor physics analysis was performed for the Advanced Neutron Source reactor with a three-element core configuration. The analysis was performed with a two-dimensional r-z 20-energy-group finite-difference diffusion theory model of the 17-d fuel cycle. The model included equivalent r-z geometry representations of the central control rods, the irradiation and production targets, and reflector components. Calculated quantities include fuel cycle parameters, fuel element power distributions, unperturbed neutron fluxes in the reflector and target regions, reactivity perturbations, and neutron kinetics parameters.

  8. Transmission Electron Microscopy Study of Graphite under in situ Ion Irradiation

    NASA Astrophysics Data System (ADS)

    Hinks, J. A.; Jones, A. N.; Theodosiou, A.; van den Berg, J. A.; Donnelly, S. E.

    2012-07-01

    Graphite is employed as a moderator and structural component in 18 of the UK's fleet of Magnox and Advanced Gas-cooled Reactors (AGRs). During the operational lifetime of a reactor, graphite undergoes complex physical and mechanical property changes including dimensional modification, owing to the effects of temperature, oxidation and irradiation-induced atomic displacements. In order to safely extend the lifetime of the current fleet of AGRs, and also to develop materials for GenIV concepts such as the Very-High-Temperature Reactor (VHTR), it is important to gain a better understanding of the fundamental atomic processes which underpin the behaviour of graphite under current and future operational conditions. This study has focused on the effects of temperature and displacing radiation on the evolution of Mrozowski cracks in highly-orientated pyrolytic graphite (HOPG) using the new Microscope and Ion Accelerator for Materials Investigations (MIAMI) facility. This instrument allows transmission electron microscopy to be performed in situ whilst simultaneously ion irradiating to radiation damage levels typically reached in a reactor. By using this technique, it is possible to explore the development of radiation damage under a range of different conditions continuously from start-to-finish rather than just observing the end-states accessible in ex situ studies.

  9. 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)

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. 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

  16. 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.

  17. 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.

  18. Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems: Annual and Final Report

    SciTech Connect

    Kevan D. Weaver; Theron Marshall; James Parry

    2005-10-01

    The Idaho National Laboratory (INL) contribution to the Nuclear Energy Research Initiative (NERI) project number 2002-005 was divided into reactor physics, and thermal-hydraulics and plant design. The research targeted credible physics and thermal-hydraulics models for a gas-cooled fast reactor, analyzing various fuel and in-core fuel cycle options to achieve a true breed and burn core, and performing a design basis Loss of Coolant Accident (LOCA) analysis on that design. For the physics analysis, a 1/8 core model was created using different enrichments and simulated equilibrium fuel loadings. The model was used to locate the hot spot of the reactor, and the peak to average energy deposition at that location. The model was also used to create contour plots of the flux and energy deposition over the volume of the reactor. The eigenvalue over time was evaluated using three different fuel configurations with the same core geometry. The breeding capabilities of this configuration were excellent for a 7% U-235 model and good in both a plutonium model and a 14% U-235 model. Changing the fuel composition from the Pu fuel which provided about 78% U-238 for breeding to the 14% U-235 fuel with about 86% U-238 slowed the rate of decrease in the eigenvalue a noticeable amount. Switching to the 7% U-235 fuel with about 93% U-238 showed an increase in the eigenvalue over time. For the thermal-hydraulic analysis, the reactor design used was the one forwarded by the MIT team. This reactor design uses helium coolant, a Brayton cycle, and has a thermal power of 600 MW. The core design parameters were supplied by MIT; however, the other key reactor components that were necessary for a plausible simulation of a LOCA were not defined. The thermal-hydraulic and plant design research concentrated on determining reasonable values for those undefined components. The LOCA simulation was intended to provide insights on the influence of the Reactor Cavity Cooling System (RCCS), the

  19. 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

  20. 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.

  1. Diversion assumptions for high-powered research reactors

    SciTech Connect

    Binford, F.T.

    1984-01-01

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

  2. 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)

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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)

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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)

  1. 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

  2. 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)

  3. 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.

  4. 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

  5. Analysis of the NP-MHTGR concept: A comparison of reactor physics methods

    SciTech Connect

    Pelowitz, D B; Sapir, J L; Wing, J E

    1991-12-31

    Using the MCNP and ONEDANT analysis codes, we investigated basic neutronic characteristics of the NP-MHTGR preconceptual design. Exploratory steady-state analyses of k{sub eff}, neutron reaction rates, andtemperature reactivity coefficients were conducted to check die ability of our reactor physics methods to adequately model the highly heterogeneous NP-MHTGR reactor. Results of unit-fuel-cell analyses indicate that a three-region ONEDANT model adequately approximates the unit-fuel-cell lattice geometry. However, core-block analyses indicate that approximating an hexagonal heterogeneous block by a one-dimensional annular target cell can introduce significant calculational error. Investigating the core-block temperature coefficient of reactivity, we found that all components of the coefficient are negative and the delayed component contributes {approx}85% of the total temperature effect. Investigation of the full reactor temperature coefficient in the NP-MHTGR determined that all contributions from the active core are negative, with prompt effects again contributing {approx}15% of the total core coefficient Temperature-coefficient contributions from each of the reflector regions appear to be positive, but exhibit a smaller magnitude than those in the core. These positive contributions apparently are caused by reduced carbon and boron absorptions at the higher reflector temperatures. From a safety perspective, a conclusion as to the adequacy of the temperature coefficient cannot be drawn from its magnitude alone, but must be based on specific transient or accident analyses which incorporate all feedback effects. Calculational differences between MCNP and ONEDANT were as high as {approx} 1.2% for the reactor criticality eigenvalue and on the order of 20% for the core temperature coefficient.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  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

    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.

  14. Benchmarking thermal neutron scattering in graphite

    NASA Astrophysics Data System (ADS)

    Zhou, Tong

    A Slowing-Down-Time experiment was designed and performed at the Oak Ridge National Laboratory (ORNL) by using the Oak Ridge Electron Linear Accelerator (ORELA) as a neutron source to study the neutron thermalization in graphite at room and higher temperatures. The MCNP5 code was utilized to simulate the detector responses and help optimize the experimental design including the size of the graphite assembly, furnace, shielding system and detector position. To facilitate such analysis, MCNP5 version 1.30 was modified to enable perturbation calculation using point detector type tallies. By using the modified MCNP5 code, the sensitivity of the experimental models to the graphite total thermal neutron cross-sections was studied to optimize the design of the experiment. Measurements of slowing-down-time spectrum in graphite were performed at room temperature for a 70x70x70 cm graphite pile by using a Li-6 scintillator and a U-235 fission counter at different locations. The measurements were directly compared to Monte Carlo simulations that use different graphite thermal neutron scattering cross-section libraries. Simulations based on the ENDF/B-VI graphite library were found to have a 30%-40% disagreement with the measurements. In addition to the graphite SDT experiment, which provided the data in the energy region above the graphite Bragg-cutoff energy, transmission experiments were performed for different types of graphite samples using the NIST 8.9 A beam (located at NG-6) to investigating the energy region below the Bragg-cutoff energy. Measurements confirmed that reactor grade graphite, which is a two phase material (crystalline graphite and binder (amorphous-like) carbon), has different thermal neutron scattering cross section from pyrolytic graphite (crystalline graphite). The experiments presented in this work compliment each other and provide an experimental data set which can be used to benchmark graphite thermal neutron scattering cross section libraries that

  15. 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)

  16. 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.

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

    SciTech Connect

    Binford, F.T.

    1984-01-01

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

  18. Hybrid parallel code acceleration methods in full-core reactor physics calculations

    SciTech Connect

    Courau, T.; Plagne, L.; Ponicot, A.; Sjoden, G.

    2012-07-01

    When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)

  19. 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.

  20. 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.

  1. Personnel dosimetry intercomparison studies at the Health Physics Research Reactor: a summary (1974-80).

    PubMed

    Sims, C S; Swaja, R E

    1982-01-01

    Six personnel dosimetry intercomparison studies using the Health Physics Research Reactor at the Oak Ridge National Laboratory were conducted between 1974 and 1980. These studies allowed participants to test their neutron and gamma-ray dosimeters under a variety of mixed-field spectral conditions and to compare their results with those of others making measurements under identical conditions. Fifty-eight participant organizations, about half of which participated in more than one study, made approx. 2000 measurements of the neutron and gamma-dose-equivalent. Dose equivalents in the 0.1-12 mSv (i.e. 10-1200 mrem) range were determined for five different shielded reactor spectra using three basic types of dosimeters (thermoluminescent albedo, nuclear emulsion film and track etch) for neutron measurements and two basic types (film and thermoluminescent dosimeters) for the gamma-measurements. The data from the six studies are summarized, analyzed and explained. Intercomparison of the participants' results and consideration of reference dosimetry allows several conclusions to be made relative to the status of and trends in personnel neutron and gamma-ray dosimetry. PMID:7056645

  2. 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)

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

    DOE PAGES

    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 andmore » 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.« less

  4. 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.

  5. 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.

  6. 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.

  7. GUM Analysis for TIMS and SIMS Isotopic Ratios in Graphite

    SciTech Connect

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

    2007-04-01

    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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. Development and engineering plan for graphite spent fuels conditioning program

    SciTech Connect

    Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

    1993-09-01

    Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated.

  17. Physics design for the Brookhaven Medical Research Reactor epithermal neutron source.

    PubMed

    Wheeler, F J; Parsons, D K; Nigg, D W; Wessol, D E; Miller, L G; Fairchild, R G

    1990-01-01

    A collaborative effort by researchers at the Idaho National Engineering Laboratory and the Brookhaven National Laboratory has resulted in the design and implementation of an epithermal-neutron source at the Brookhaven Medical Research Reactor (BMRR). Large aluminum containers, filled with aluminum oxide tiles and aluminum spacers, were tailored to pre-existing compartments on the animal side of the reactor facility. A layer of cadmium was used to minimize the thermal-neutron component. Additional bismuth was added to the pre-existing bismuth shield to minimize the gamma component of the beam. Lead was also added to reduce gamma streaming around the bismuth. The physics design methods are outlined in this paper. Information available to date shows close agreement between calculated and measured beam parameters. The neutron spectrum is predominantly in the intermediate energy range (0.5 eV - 10 keV). The peak flux intensity is 6.4E + 12 n/(m2.s.MW) at the center of the beam on the outer surface of the final gamma shield. The corresponding neutron current is 3.8E + 12 n/(m2.s.MW). Presently, the core operates at a maximum of 3 MW. The fast-neutron KERMA is 3.6E-15 cGy/(n/m2) and the gamma KERMA is 5.0E-16 cGY/(n/m2) for the unperturbed beam. The neutron intensity falls off rapidly with distance from the outer shield and the thermal flux realized in phantom or tissue is strongly dependent on the beam-delimiter and target geometry.

  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. 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.

  20. 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.

  1. 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.

  2. Advances in implosion physics, alternative targets design, and neutron effects on heavy ion fusion reactors

    NASA Astrophysics Data System (ADS)

    Velarde, G.; Perlado, J. M.; Alonso, E.; Alonso, M.; Domínguez, E.; Rubiano, J. G.; Gil, J. M.; Gómez del Rio, J.; Lodi, D.; Malerba, L.; Marian, J.; Martel, P.; Martínez-Val, J. M.; Mínguez, E.; Piera, M.; Ogando, F.; Reyes, S.; Salvador, M.; Sanz, J.; Sauvan, P.; Velarde, M.; Velarde, P.

    2001-05-01

    The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DT x fuel with a small tritium initial content ( x<3%) could work in a catalytic regime in Inertial Fusion Targets, at very high burning temperatures (≫100 keV). Otherwise, the cross-section of DT remains much higher than that of DD and no internal breeding of tritium can take place. Improvements in the calculation model allow to properly simulate the effect of inverse Compton scattering which tends to lower Te and to enhance radiation losses, reducing the plasma temperature, Ti. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination barriers in Si

  3. 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.

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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.

  11. 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)

  12. 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.

  13. 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.

  14. Physics and thermal hydraulics design of a small water cooled reactor fuelled with plutonium in rock-like oxide (ROX) form

    SciTech Connect

    Gaultier, M.; Danguy, G.; Perry, A.; Williams, A.; Brushwood, J.; Thompson, A.; Beeley, P. A.

    2006-07-01

    This paper describes the Physics and Thermal Hydraulics areas of a design study for a small water-cooled reactor. The aim was to design a Pressurised Water Reactor (PWR) of maximum power 80 MWt, using a dispersed layout, capable of maximising primary natural circulation flow. The reactor fuel consists of plutonium contained in granular form within a Rock-like Oxide (ROX) pellet structure. (authors)

  15. Protection of nuclear graphite toward fluoride molten salt by glassy carbon deposit

    NASA Astrophysics Data System (ADS)

    Bernardet, V.; Gomes, S.; Delpeux, S.; Dubois, M.; Guérin, K.; Avignant, D.; Renaudin, G.; Duclaux, L.

    2009-02-01

    Molten salt reactor represents one of the promising future Generation IV nuclear reactors families where the fuel, a liquid molten fluoride salt, is circulating through the graphite reactor core. The interactions between nuclear graphite and fluoride molten salt and also the graphite surface protection were investigated in this paper by powder X-ray diffraction, micro-Raman spectroscopy and scanning electron microscopy coupled with X-ray microanalysis. Nuclear graphite discs were covered by two kinds of protection deposit: a glassy carbon coating and a double coating of pyrolitic carbon/glassy carbon. Different behaviours have been highlighted according to the presence and the nature of the coated protection film. Intercalation of molten salt between the graphite layers did not occur. Nevertheless the molten salt adhered more or less to the surface of the graphite disc, filled more or less the graphite surface porosity and perturbed more or less the graphite stacking order at the disc surface. The behaviour of unprotected graphite was far to be satisfactory after two days of immersion of graphite in molten salt at 500 °C. The best protection of the graphite disc surface, with the maximum of inertness towards molten salt, has been obtained with the double coating of pyrolitic carbon/glassy carbon.

  16. Development of a multi-physics calculation platform dedicated to irradiation devices in a material testing reactor

    SciTech Connect

    Bonaccorsi, T.; Di Salvo, J.; Aggery, A.; D'Aletto, C.; Doederlein, C.; Sireta, P.; Willermoz, G.; Daniel, M.

    2006-07-01

    The physical phenomena involved in irradiation devices within material testing reactors are complex (neutron and photon interactions, nuclear heating, thermal hydraulics, ...). However, the simulation of these phenomena requires a high precision in order to control the condition of the experiment and the development of predictive models. Until now, physicists use different tools with several approximations at each interface. The aim of this work is to develop a calculation platform dedicated to numerical multi-physics simulations of irradiation devices in the future European Jules Horowitz Reactor [1], This platform is based on a multi-physics data model which describes geometries, materials and state parameters associated with a sequence of thematic (neutronics, thermal hydraulics...) computations of these devices. Once the computation is carried out, the results can be returned to the data model (DM). The DM is encapsulated in a dedicated module of the SALOME platform [2] and exchanges data with SALOME native modules. This method allows a parametric description of a study, independent of the code used to perform the simulation. The application proposed in this paper concerns neutronic calculation of a fuel irradiation device with the new method of characteristics implemented in the APOLLO2 code [3]. The device is located at the periphery of the OSIRIS core. This choice is motivated by the possibility to compare the calculation with experimental results, which cannot be done for the Jules Horowitz Reactor, currently in design study phase. (authors)

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

    SciTech Connect

    Michael A. Pope

    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.

  18. 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

  19. 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

  20. 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.

  1. 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)

  2. 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)

  3. 75 FR 67636 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-03

    ... rule was published on October 13, 2010 (75 FR 62695), and the public comment period runs through... ``irradiated reactor fuel'' and ``spent nuclear fuel'' (SNF) are used interchangeably). The NRC has prepared...

  4. 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

  5. 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)

  6. 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.

  7. Sorption/desorption behavior of iodine on graphite. [HTGR

    SciTech Connect

    Lorenz, R.A.; Dyer, F.F.; Towns, R.L.

    1982-11-01

    An experimental program was undertaken to extend the data on the sorption and desorption of iodine on graphite to more realistic high-temperature gas-cooled reactor (HTGR) operating conditions. This was accomplished by heating compacts of H-451 or S-2020 graphite at 250 to 1000/sup 0/C in continuously flowing helium (at atmospheric pressure) that contained iodine at pressures of 10/sup -6/ Pa (10/sup -11/ to 10/sup -1/ to 10/sup -6/ bar). Equilibrium adsorption data were generally well-behaved and reproducible with the H-451 graphite. Type S-2020 graphite sorbed more iodine, especially in the temperature range 700 to 800/sup 0/C. Although desorption appeared to initiate as reversible sorption, the rate of loss of iodine decreased with time. During desorption from the S-2020 graphite at 700 and 800/sup 0/C, a temporary plateau was reached that was sensitive to helium flow and pressure.

  8. 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

  9. Biological Removal of Radiocarbon-14 from Irradiated Graphite

    SciTech Connect

    Molokwane, P.E.; Chirwa, E.M.N.

    2008-07-01

    This paper reports on the preliminary study on the bioseparation of radiocarbon-14 (C-14) using a mixed-culture of microorganisms. The bioseparation principle is critically analyzed and is benchmarked against existing C-14 separation methods using physical chemical processes. The preliminary study indicated probable bio-separate of C-14 from solution prepared from a nuclear graphite mixture, even though the findings need to be verified. The current experiment consisted of a growth vessel and a bio-filter operated in a closed loop. The bio-filter was not installed for the purpose of treatment but rather as a method of isolation of microorganisms for further processing. Significant amounts of C-14 were detected in the trapped cells in the bio-filter, significantly higher than in controls taken before adding carbon sources containing C-14. The microorganisms were grown under micro-aerobic conditions with graphite carbon and commercially purchased powdered carbon as the predominant supplied carbon sources. Small amounts of sucrose (500 mg/L) were added at 48 hour intervals to maintain culture balance. A proof of concept study is underway to determine the C-14 mass balance, characterize the microorganisms in the reactor, and establish the presence or absence of processes that might have affected the preliminary observations. This research represents an exploration into a new field using a new philosophy for treatment of C-14 in low-level waste. (authors)

  10. Method for producing dustless graphite spheres from waste graphite fines

    DOEpatents

    Pappano, Peter J; Rogers, Michael R

    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. 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.

  12. 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.

  13. 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.

  14. 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].

  15. 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.

  16. 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.

  17. 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.

  18. 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

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

    DOE PAGES

    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; et al

    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

  20. 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

    ..., 2010 (75 FR 62695). The public comment period for this proposed rule was scheduled to expire on February 11, 2011; however, on January 10, 2011 (76 FR 1376), the public comment period for the proposed... Irradiated Reactor Fuel in Transit.'' On November 3, 2010 (75 FR 67636), the NRC published for public...

  1. 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.

  2. 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.

  3. 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

  4. 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.

  5. 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)

  6. 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.

  7. 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

  8. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    NASA Astrophysics Data System (ADS)

    Cisneros, Anselmo Tomas, Jr.

    and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of

  9. 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.

  10. Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

    PubMed

    Henry, R; Tiselj, I; Snoj, L

    2015-03-01

    New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. PMID:25576735

  11. 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.

  12. 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).

  13. 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 ...

  14. 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...

  15. 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...

  16. 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

  17. Eugene Wigner, The First Nuclear Reactor Engineer

    NASA Astrophysics Data System (ADS)

    Weinberg, Alvin M.

    2002-04-01

    All physicists recognize Eugene Wigner as a theoretical physicist of the very first rank. Yet Wigner's only advanced degree was in Chemical Engineering. His physics was largely self-taught. During WWII, Wigner brilliantly returned to his original occupation as an engineer. He led the small team of theoretical physicists and engineers who designed, in remarkable detail, the original graphite-moderated, water-cooled Hanford reactor, which produced the Pu239 of the Trinity and Nagasaki bombs. With his unparalleled understanding of chain reactors (matched only by Fermi) and his skill and liking for engineering, Wigner can properly be called the Founder of Nuclear Engineering. The evidence for this is demonstrated by a summary of his 37 Patents on various chain reacting systems.

  18. 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

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. Relevance of β-delayed neutron data for reactor, nuclear physics and astrophysics applications

    SciTech Connect

    Kratz, Karl-Ludwig

    2015-02-24

    Initially, yields (or abundances) and branching ratios of β-delayed neutrons (βdn) from fission products (P{sub n}-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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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. PMID:16604642

  15. 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.

  16. 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.

  17. 78 FR 31821 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-28

    ... in the Federal Register on May 20, 2013 (78 FR 29519) (RIN 3150-AI64), that amended its security... public comment on November 3, 2010 (75 FR 67636). The NRC received comments from eight commenters during... acceptable for satisfying the requirements for the physical protection of spent nuclear fuel (SNF)...

  18. 78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-19

    ... safety. On May 20, 2013 (78 FR 29520), the NRC published the final rule for 10 CFR 73.37, ``Physical... security orders issued to licensees shipping SNF during the period of October 2003 through December 2010. These orders are collectively referred to as the ``SNF Transportation Orders.'' The SNF...

  19. 78 FR 29519 - Physical Protection of Irradiated Reactor Fuel in Transit

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-20

    ... acceptable security program for SNF shipments. On November 3, 2010 (75 FR 67636), the NRC published for... Register (44 FR 34466) an interim final rule that established requirements for the physical protection of... June 3, 1980 (45 FR 37399). Section 73.37 has changed little since its promulgation in 1980....

  20. 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.

  1. 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.

  2. The ablation of graphitic materials in the sublimation regime.

    NASA Technical Reports Server (NTRS)

    Lundell, J. H.; Dickey, R. R.

    1972-01-01

    A large variety of graphitic materials have been tested in an arc heated air stream at a surface pressure of 4.3 atm and a nominal surface temperature of 3925 K. Included were commercial and developmental grades of artificial graphites, both two and three dimensional carbon-carbon composites, composites seeded with refractory compounds, and several special materials such as pyrolytic graphite, mesophase graphite, glassy carbon, and natural graphite. ATJ graphite was used as a control material. Except for the seeded materials which had poor performance, the mass loss rate for all the man-made graphitic materials fell within the range of 17 per cent less to 30 per cent more than the rate for ATJ. Thus it is concluded that wide variations in constituents, processing, fabrication and structure have relatively little effect on the ablation performance of graphitic materials, at least under the conditions of the present tests. Particulate mass loss was observed for all the materials tested and is the dominant mechanism for mass removal at the present test conditions. It is suggested that this mechanism results from physical failure, primarily by compressive thermal stress.

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. 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.

  8. 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.

  9. 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.

  10. 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)

  11. Benzylamine-assisted noncovalent exfoliation of graphite-protecting Pt nanoparticles applied as catalyst for methanol oxidation.

    PubMed

    Hsu, Chun-Han; Liao, Hsin-Yi; Wu, Yi-Fang; Kuo, Ping-Lin

    2011-07-01

    A novel method has been developed to physically exfoliate graphite and uniformly disperse Pt nanoparticles on graphite nanoplates without damaging the graphene structures. A stable aqueous suspension of graphite nanoplates was achieved by benzylamine-assisted noncovalent fuctionalization to graphite and characterized by transmission electron microscopy, X-ray diffraction and Raman spectroscopy. A uniform dispersion of Pt nanoparticles was then prepared on the graphite nanoplates, where the benzylamine acts as a stabilizer. These Pt loaded graphite nanoplates were then prepared as an electrode, which significantly increased catalytic activity toward the methanol oxidation reaction, resulting in a 60% increment in mass activity compared to that of E-TEK. PMID:21668025

  12. 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.

  13. 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

  14. 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.

  15. a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

    NASA Astrophysics Data System (ADS)

    Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

    2009-08-01

    This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

  16. Radio Frequency-Thermal Plasma Synthesis of Metal Nano-Thin Exfoliated Graphite Hybrid.

    PubMed

    Shin, Myoung-Sun; Lee, Kyu-Hang; Kim, Jung-Gil; Choi, Sun-Yong; Cho, Guang-Sup; Kim, Seong-In

    2015-11-01

    Nano-metal with nano-thin exfoliated (NTE) graphite hybrid material has been synthesized by radio frequency (RF) thermal plasma. A micro-sized nickel powder and the NTE graphite powder were fed into the RF plasma and nano sized nickel particles attached to the surface of the NTE graphite were found. In the high temperature of RF thermal plasma that is of higher than 10,000 K, the NTE graphite was not vaporized or damaged, while the metal powder was vaporized. The size of nickel nanoparticles on the NTE graphite was 40 80 nme. The size and number density of produced metal nanoparticle can be controlled by the process pressure in a reactor, the feeding ratio of raw materials, and the flow rate of working gas. X-ray diffraction results of the produced hybrid nano material indicate that there was a bonding between the nano metal and the NTE graphite. The inert nature of surface of the NTE graphite has been a barrier for the NTE graphite to be used a compounding additive. The nano metal covered NTE graphite will open up many potential applications of NTE graphite and polymer compound materials.

  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 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

  18. 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

  19. Perchloric acid dissolution of graphite and pyrolytic carbon.

    PubMed

    Buzzelli, G; Mosen, A W

    1977-06-01

    Three procedures are described for the wet oxidation, with perchloric acid, of nuclear graphite, pyrocarbon-coated fuel particles, and other carbonaceous materials used in high-temperature gas-cooled nuclear reactors (HTGRs). The first procedure is for dissolution of graphite and pyrolytic carbon, the second for dissolution of milligram quantities of HTGR fuel particles, and the last for dissolution of more easily oxidized carbonaceous materials such as charcoal. These procedures were developed primarily for the dissolution of irradiated materials before fission-product or burn-up analyses, but they are also used for dissolution of unirradiated materials.

  20. Liquid deuterium cold source in graphite thermal column

    NASA Astrophysics Data System (ADS)

    Utsuro, M.; Kawai, T.; Maeda, Y.; Yamaoka, H.; Akiyoshi, T.; Okamoto, S.

    1989-01-01

    A liquid deuterium cold source with a non-spherical moderator chamber of about 4 litres was installed into the graphite thermal column of 5 MW Kyoto University Reactor (KUR). Three cold neutron holes and one very cold neutron hole are provided in the graphite for beam extractions. The operation tests with hydrogen liquefied in the condenser showed satisfactory performances and high gain factors of cold and very cold neutrons of more than 20 and 10, respectively. Neutron measurements with the deuterium moderator are now in progress.

  1. 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.

  2. 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.

  3. 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...

  4. Industrial Applications of Graphite Fluoride Fibers

    NASA Technical Reports Server (NTRS)

    Hung, Ching-Cheh; Kucera, Donald

    1991-01-01

    Based on fluorination technology developed during 1934 to 1959, and the fiber technology developed during the 1970s, a new process was developed to produce graphite fluoride fibers. In the process, pitch based graphitized carbon fibers are at first intercalated and deintercalated several times by bromine and iodine, followed by several cycles of nitrogen heating and fluorination at 350 to 370 C. Electrical, mechanical, and thermal properties of this fiber depend on the fluorination process and the fluorine content of the graphite fluoride product. However, these properties are between those of graphite and those of PTFE (Teflon). Therefore, it is considered to be a semiplastic. The physical properties suggest that this new material may have many new and unexplored applications. For example, it can be a thermally conductive electrical insulator. Its coefficient of thermal expansion (CTE) can be adjusted to match that of silicon, and therefore, it can be a heat sinking printed circuit board which is CTE compatible with silicon. Using these fibers in printed circuit boards may provide improved electrical performance and reliability of the electronics on the board over existing designs. Also, since it releases fluorine at 300 C or higher, it can be used as a material to store fluorine and to conduct fluorination. This application may simplify the fluorination process and reduce the risk of handling fluorine.

  5. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1957-09-17

    A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

  6. 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

  7. 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

  8. Bilirubin Oxidase from Myrothecium verrucaria Physically Absorbed on Graphite Electrodes. Insights into the Alternative Resting Form and the Sources of Activity Loss.

    PubMed

    Tasca, Federico; Farias, Diego; Castro, Carmen; Acuna-Rougier, Cristina; Antiochia, Riccarda

    2015-01-01

    The oxygen reduction reaction is one of the most important chemical processes in energy converting systems and living organisms. Mediator-less, direct electro-catalytic reduction of oxygen to water was achieved on spectrographite electrodes modified by physical adsorption of bilirubin oxidases from Myrothecium verrucaria. The existence of an alternative resting form of the enzyme is validated. The effect on the catalytic cycle of temperature, pH and the presence of halogens in the buffer was investigated. Previous results on the electrochemistry of bilirubin oxidase and on the impact of the presence of halogens are reviewed and reinterpreted.

  9. Bilirubin Oxidase from Myrothecium verrucaria Physically Absorbed on Graphite Electrodes. Insights into the Alternative Resting Form and the Sources of Activity Loss

    PubMed Central

    Tasca, Federico; Farias, Diego; Castro, Carmen; Acuna-Rougier, Cristina; Antiochia, Riccarda

    2015-01-01

    The oxygen reduction reaction is one of the most important chemical processes in energy converting systems and living organisms. Mediator-less, direct electro-catalytic reduction of oxygen to water was achieved on spectrographite electrodes modified by physical adsorption of bilirubin oxidases from Myrothecium verrucaria. The existence of an alternative resting form of the enzyme is validated. The effect on the catalytic cycle of temperature, pH and the presence of halogens in the buffer was investigated. Previous results on the electrochemistry of bilirubin oxidase and on the impact of the presence of halogens are reviewed and reinterpreted. PMID:26196288

  10. Graphite polyhedral crystals.

    PubMed

    Gogotsi, Y; Libera, J A; Kalashnikov, N; Yoshimura, M

    2000-10-13

    Polyhedral nano- and microstructures with shapes of faceted needles, rods, rings, barrels, and double-tipped pyramids, which we call graphite polyhedral crystals (GPCs), have been discovered. They were found in pores of glassy carbon. They have nanotube cores and graphite faces, and they can exhibit unusual sevenfold, ninefold, or more complex axial symmetry. Although some are giant radially extended nanotubes, Raman spectroscopy and transmission electron microscopy suggest GPCs have a degree of perfection higher than in multiwall nanotubes of similar size. The crystals are up to 1 micrometer in cross section and 5 micrometers in length, and they can probably be grown in much larger sizes. Preliminary results suggest a high electrical conductivity, strength, and chemical stability of GPC.

  11. 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

  12. 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.

  13. 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.

  14. Enhancing the oxidation resistance of graphite by applying an SiC coat with crack healing at an elevated temperature

    NASA Astrophysics Data System (ADS)

    Park, Jae-Won; Kim, Eung-Seon; Kim, Jae-Un; Kim, Yootaek; Windes, William E.

    2016-08-01

    The potential of reducing the oxidation of the supporting graphite components during normal and/or accident conditions in the Very High Temperature Reactor (VHTR) design has been studied. In this work efforts have been made to slow the oxidation process of the graphite with a thin SiC coating (∼ 10 μm). Upon heating at ≥ 1173 K in air, the spallations and cracks were formed in the dense columnar structured SiC coating layer grown on the graphite with a functionally gradient electron beam physical vapor deposition (EB-PVD. In accordance with the formations of these defects, the sample was vigorously oxidized, leaving only the SiC coating layer. Then, efforts were made to heal the surface defects using additional EB-PVD with ion beam bombardment and chemical vapor deposition (CVD). The EB-PVD did not effectively heal the cracks. But, the CVD was more appropriate for crack healing, likely due to its excellent crack line filling capability with a high density and high aspect ratio. It took ∼ 34 min for the 20% weight loss of the CVD crack healed sample in the oxidation test with annealing at 1173 K, while it took ∼ 8 min for the EB-PVD coated sample, which means it took ∼4 times longer at 1173 K for the same weight reduction in this experimental set-up.

  15. 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.

  16. 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.

  17. CMB-13 research on carbon and graphite

    NASA Technical Reports Server (NTRS)

    Smith, M. C.

    1972-01-01

    Preliminary results of the research on carbon and graphite accomplished during this report period are presented. Included are: particle characteristics of Santa Maria fillers, compositions and density data for hot-molded Santa Maria graphites, properties of hot-molded Santa Maria graphites, and properties of hot-molded anisotropic graphites. Ablation-resistant graphites are also discussed.

  18. Graphite fiber reinforced thermoplastic resins

    NASA Technical Reports Server (NTRS)

    Navak, R. C.

    1977-01-01

    The results of a program designed to optimize the fabrication procedures for graphite thermoplastic composites are described. The properties of the composites as a function of temperature were measured and graphite thermoplastic fan exit guide vanes were fabricated and tested. Three thermoplastics were included in the investigation: polysulfone, polyethersulfone, and polyarylsulfone. Type HMS graphite was used as the reinforcement. Bending fatigue tests of HMS graphite/polyethersulfone demonstrated a gradual shear failure mode which resulted in a loss of stiffness in the specimens. Preliminary curves were generated to show the loss in stiffness as a function of stress and number of cycles. Fan exit guide vanes of HMS graphite polyethersulfone were satisfactorily fabricated in the final phase of the program. These were found to have stiffness and better fatigue behavior than graphite epoxy vanes which were formerly bill of material.

  19. Graphite fluoride fibers and their applications in the space industry

    NASA Technical Reports Server (NTRS)

    Hung, Ching-Chen; Long, Martin; Dever, Therese

    1990-01-01

    Characterization and potential space applications of graphite fluoride fibers from commercially available graphitized carbon fibers are presented. Graphite fluoride fibers with fluorine to carbon ratios of 0.65 and 0.68 were found to have electrical resistivity values of 10(exp 4) and 10(exp 11) Ohms-cm, respectively, and thermal conductivity values of 24 and 5 W/m-K, respectively. At this fluorine content range, the fibers have tensile strength of 0.25 + or - 0.10 GPa (36 + or - 14 ksi), Young's modulus of 170 + or - 30 GPa (25 + or - 5 Msi). The coefficient of thermal expansion value of a sample with fluorine to carbon ratio of 0.61 was found to be 7 ppm/C. These properties change and approach the graphite value as the fluorine content approach 0. Electrically insulative graphite fluoride fiber is at least five times more thermally conductive than fiberglass. Therefore, it can be used as a heat sinking printed circuit board material for low temperature, long life power electronics in spacecraft. Also, partially fluorinated fiber with tailor-made physical properties to meet the requirements of certain engineering design can be produced. For example, a partially fluorinated fiber could have a predetermined CTE value in -1.5 to 7 ppm/C range and would be suitable for use in solar concentrators in solar dynamic power systems. It could also have a predetermined electrical resistivity value suitable for use as a low observable material. Experimental data indicate that slightly fluorinated graphite fibers are more durable in the atomic oxygen environment than pristine graphite. Therefore, fluorination of graphite used in the construction of spacecraft that would be exposed to the low Earth orbit atomic oxygen may protect defect sites in atomic oxygen protective coatings and therefore decrease the rate of degradation of graphite.

  20. 75 FR 6413 - Office of New Reactors; Proposed Revision to Standard Review Plan, Section 14.3.12 on Physical...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-09

    ... Regulations, Part 73, Power Reactor Security Rule (published in the Federal Register (FR) on March 27, 2009 (74 FR 13926)). The previous version of this SRP section was published in March 2007 as an initial... COMMISSION Office of New Reactors; Proposed Revision to Standard Review Plan, Section 14.3.12 on...

  1. 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.

  2. Simulation of water cluster assembly on a graphite surface.

    PubMed

    Lin, C S; Zhang, R Q; Lee, S T; Elstner, M; Frauenheim, Th; Wan, L J

    2005-07-28

    The assembly of small water clusters (H2O)n, n = 1-6, on a graphite surface is studied using a density functional tight-binding method complemented with an empirical van der Waals force correction, with confirmation using second-order Møller-Plesset perturbation theory. It is shown that the optimized geometry of the water hexamer may change its original structure to an isoenergy one when interacting with a graphite surface in some specific orientation, while the smaller water cluster will maintain its cyclic or linear configurations (for the water dimer). The binding energy of water clusters interacting with graphite is dependent on the number of water molecules that form hydrogen bonds, but is independent of the water cluster size. These physically adsorbed water clusters show little change in their IR peak position and leave an almost perfect graphite surface.

  3. 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.

  4. Heat exchanger using graphite foam

    DOEpatents

    Campagna, Michael Joseph; Callas, James John

    2012-09-25

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

  5. Towards graphene iodide: iodination of graphite oxide

    NASA Astrophysics Data System (ADS)

    Šimek, Petr; Klímová, Kateřina; Sedmidubský, David; Jankovský, Ondřej; Pumera, Martin; Sofer, Zdeněk

    2014-11-01

    Halogenated graphene derivatives are interesting owing to their outstanding physical and chemical properties. In this paper, we present various methods for the synthesis of iodinated graphene derivatives by the iodination of graphite oxides prepared according to either the Hummers or Hofmann method. Both graphite oxides were iodinated by iodine or hydroiodic acid under reflux or in an autoclave at elevated temperatures (240 °C) and pressures (over 100 bar). The influence of both graphite oxide precursors on the properties of resulting iodinated graphenes was investigated by various techniques, including SEM, SEM-EDS, high-resolution XPS, FTIR, STA, and Raman spectroscopy. Electrical resistivity was measured by a standard four point technique. In addition, the electrochemical properties were investigated by cyclic voltammetry. Although the iodinated graphenes were structurally similar, they had remarkably different concentrations of iodine. The most highly iodinated graphenes (iodine concentration above 30 wt%) exhibited relatively high C/O ratios, confirming high degrees of reduction. Iodine is incorporated in the form of covalent bonds to carbon atoms or as polyiodide anions non-covalently bonded through the charge transfer reaction with the graphene framework. Iodinated graphenes with such properties could be used as the starting material for further chemical modifications or as flame-retardant additives.Halogenated graphene derivatives are interesting owing to their outstanding physical and chemical properties. In this paper, we present various methods for the synthesis of iodinated graphene derivatives by the iodination of graphite oxides prepared according to either the Hummers or Hofmann method. Both graphite oxides were iodinated by iodine or hydroiodic acid under reflux or in an autoclave at elevated temperatures (240 °C) and pressures (over 100 bar). The influence of both graphite oxide precursors on the properties of resulting iodinated graphenes was

  6. Oxidation rate of graphitic matrix material in the kinetic regime for VHTR air ingress accident scenarios

    NASA Astrophysics Data System (ADS)

    Lee, Jo Jo; Ghosh, Tushar K.; Loyalka, Sudarshan K.

    2014-08-01

    Data on oxidation rates of matrix-grade graphite in the kinetically-controlled temperature regime of graphite oxidation are needed for safety analysis of High Temperature Gas Cooled Reactors and Very High Temperature Reactors. In this work, the oxidation rate of graphitic matrix material GKrS was measured thermogravimetrically for various oxygen concentrations and with temperatures from 873 to 1873 K. A semi-empirical Arrhenius rate equation was also developed for this temperature range. The activation energy of the graphitic material is found to be about 111.5 kJ/mol. The order of reaction was found to be about 0.89. The surface of oxidized GKrS was characterized by Scanning Electron Microscopy, Electron Dispersive Spectroscopy, Fourier Transform Infrared Spectroscopy and X-ray Photoelectron Spectroscopy.

  7. Graphitic packing removal tool

    SciTech Connect

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

    1996-12-31

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

  8. Graphitic packing removal tool

    DOEpatents

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

    1997-11-11

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

  9. Graphitic packing removal tool

    DOEpatents

    Meyers, Kurt Edward; Kolsun, George J.

    1997-01-01

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

  10. Preliminary analysis of graphite dust releasing behavior in accident for HTR

    SciTech Connect

    Peng, W.; Yang, X. Y.; Yu, S. Y.; Wang, J.

    2012-07-01

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

  11. Ion irradiation of 37Cl implanted nuclear graphite: Effect of the energy deposition on the chlorine behavior and consequences for the mobility of 36Cl in irradiated graphite

    NASA Astrophysics Data System (ADS)

    Toulhoat, N.; Moncoffre, N.; Bérerd, N.; Pipon, Y.; Blondel, A.; Galy, N.; Sainsot, P.; Rouzaud, J.-N.; Deldicque, D.

    2015-09-01

    Graphite is used in many types of nuclear reactors due to its ability to slow down fast neutrons without capturing them. Whatever the reactor design, the irradiated graphite waste management has to be faced sooner or later regarding the production of long lived or dose determining radioactive species such as 14C, 3H or 36Cl. The first carbon dioxide cooled, graphite moderated nuclear reactors resulted in a huge quantity of irradiated graphite waste for which the management needs a previous assessment of the radioactive inventory and the radionuclide's location and speciation. As the detection limits of usual spectroscopic methods are generally not adequate to detect the low concentration levels (<1 ppm) of the radionuclides, we used an indirect approach based on the implantation of 37Cl, to simulate the presence of 36Cl. Our previous studies show that temperature is one of the main factors to be considered regarding the structural evolution of nuclear graphite and chlorine mobility during reactor operation. However, thermal release of chlorine cannot be solely responsible for the depletion of the 36Cl inventory. We propose in this paper to study the impact of irradiation and its synergetic effects with temperature on chlorine release. Indeed, the collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic collisions. However, a small part of the recoil carbon atom energy is also transferred to the lattice through electronic excitation. This paper aims at elucidating the effects of the different irradiation regimes (ballistic and electronic) using ion irradiation, on the mobility of implanted 37Cl, taking into account the initial disorder level of the nuclear graphite.

  12. 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.

  13. Testing of Small Graphite Samples for Nuclear Qualification

    SciTech Connect

    Julie Chapman

    2010-11-01

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

  14. Thermal migration of deuterium implanted in graphite: Influence of free surface proximity and structure

    NASA Astrophysics Data System (ADS)

    Le Guillou, M.; Moncoffre, N.; Toulhoat, N.; Pipon, Y.; Ammar, M. R.; Rouzaud, J. N.; Deldicque, D.

    2016-03-01

    This paper is a contribution to the study of the behavior of activation products produced in irradiated nuclear graphite, graphite being the moderator of the first French generation of CO2 cooled nuclear fission reactors. This paper is focused on the thermal release of Tritium, a major contributor to the initial activity, taking into account the role of the free surfaces (open pores and graphite surface). Two kinds of graphite were compared. On one hand, Highly Oriented Pyrolitic Graphite (HOPG), a model well graphitized graphite, and on the other hand, SLA2, a porous less graphitized nuclear graphite. Deuterium ion implantation at three different energies 70, 200 and 390 keV allows simulating the presence of Tritium at three different depths, corresponding respectively to projected ranges Rp of 0.75, 1.7 and 3.2 μm. The D isotopic tracing is performed thanks to the D(3He,p)4He nuclear reaction. The graphite structure is studied by Raman microspectrometry. Thermal annealing is performed in the temperature range 200-1200 °C up to 300 h annealing time. As observed in a previous study, the results show that the D release occurs according to three kinetic regimes: a rapid permeation through open pores, a transient regime corresponding to detrapping and diffusion of D located at low energy sites correlated to the edges of crystallites and finally a saturation regime attributed to detrapping of interstitial D located at high energy sites inside the crystallites. Below 600 °C, D release is negligible whatever the implantation depth and the graphite type. The present paper clearly puts forward that above 600 °C, the D release decreases at deeper implantation depths and strongly depends on the graphite structure. In HOPG where high energy sites are more abundant, the D release is less dependent on the surface proximity compared to SLA2. In SLA2, in which the low energy sites prevail, the D release curves are clearly shifted towards lower temperatures when D is located

  15. The effect of neutron irradiation damage on the properties of grade NBG-10 graphite

    NASA Astrophysics Data System (ADS)

    Burchell, Timothy D.; Snead, Lance L.

    2007-09-01

    Nuclear block graphite-10 (NBG-10) is a medium-grain, near-isotropic graphite manufactured by SGL Carbon Company at their plant in Chedde, France. NBG-10 graphite was developed as a candidate core structural material for the pebble bed modular reactor (PBMR) currently being designed in South Africa, and for prismatic reactor concepts being developed in the USA and Europe. NBG-10 is one of several graphites included in the US-DOE Very High Temperature Reactor (VHTR) program. Thirty-six NBG-10 graphite flexure bars have been successfully irradiated in a series of 18 HFIR PTT capsules at ORNL. The capsule irradiation temperatures were 294 ± 25, 360 ± 25 and 691 ± 25 °C. The peak doses attained were 4.93, 6.67, and 6.69 × 10 25 n/m 2 [ E > 0.1 MeV] at ˜294, ˜360, and ˜691 °C, respectively. The high temperature irradiation volume and dimensional change behavior, and flexure strength and elastic modulus changes of NBG-10 were similar to other extruded, near-isotropic grades, such as H-451, which has been irradiated previously at ORNL. The low temperature (˜294 °C) irradiation volume and dimensional change behavior was also as expected for extruded graphites, i.e., exhibiting low dose swelling prior to shrinkage. This behavior was attributed to the relaxation of internal stress arising from the graphite manufacturing process and specimen machining. While the data reported here do not represent a complete database for NBG-10 graphite, they give a measure of confidence that the current generation of nuclear graphites will behave in a familiar and well understood manner.

  16. Effect of microstructure on air oxidation resistance of nuclear graphite

    SciTech Connect

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

    2012-01-01

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

  17. Analysis and Numerical Solution for Multi-Physics Coupling of Neutron Diffusion and Thermomechanics in Spherical Fast Burst Reactors

    SciTech Connect

    Samet Y. Kadioglu; Dana A. Knoll; Cassiano de Oliveira

    2009-05-01

    Coupling neutronics to thermomechanics is important for the analysis of fast burst reactors, because the criticality and safety study of fast burst reactors heavily depends on the thermomechanical behavior of fuel materials. For instance, the shut down mechanism or the transition between super and sub-critical states are driven by the fuel material expansion or contraction. The material expansion or contraction is due to temperature gradient which results from fission power. In this paper, we introduce a numerical model for coupling of neutron diffusion and thermomechanics in fast burst reactors. We also provide some analysis of the coupled system. We studied material behaviors corresponding to different levels of power pulses.

  18. Statistical Models of Fracture Relevant to Nuclear-Grade Graphite: Review and Recommendations

    NASA Technical Reports Server (NTRS)

    Nemeth, Noel N.; Bratton, Robert L.

    2011-01-01

    The nuclear-grade (low-impurity) graphite needed for the fuel element and moderator material for next-generation (Gen IV) reactors displays large scatter in strength and a nonlinear stress-strain response from damage accumulation. This response can be characterized as quasi-brittle. In this expanded review, relevant statistical failure models for various brittle and quasi-brittle material systems are discussed with regard to strength distribution, size effect, multiaxial strength, and damage accumulation. This includes descriptions of the Weibull, Batdorf, and Burchell models as well as models that describe the strength response of composite materials, which involves distributed damage. Results from lattice simulations are included for a physics-based description of material breakdown. Consideration is given to the predicted transition between brittle and quasi-brittle damage behavior versus the density of damage (level of disorder) within the material system. The literature indicates that weakest-link-based failure modeling approaches appear to be reasonably robust in that they can be applied to materials that display distributed damage, provided that the level of disorder in the material is not too large. The Weibull distribution is argued to be the most appropriate statistical distribution to model the stochastic-strength response of graphite.

  19. REFRACTORY COATING FOR GRAPHITE MOLDS

    DOEpatents

    Stoddard, S.D.

    1958-06-24

    Refractory coating for graphite molds used in the casting of uranium is described. The coating is an alumino-silicate refractory composition which may be used as a mold surface in solid form or as a coating applied to the graphite mold. The composition consists of a mixture of ball clay, kaolin, alumina cement, alumina, water, sodium silicate, and sodium carbonate.

  20. Cryotribology of diamond and graphite

    SciTech Connect

    Iwasa, Yukikazu; Ashaboglu, A.F.; Rabinowicz, E.R.

    1996-12-31

    An experimental study was carried out on the tribological behavior of materials of interest in cryogenic applications, focusing on diamond and graphite. Both natural diamond (referred in the text as diamond) and chemical-vapor-deposition (CVD) diamond (CVD-diamond) were used. The experiment was carried out using a pin-on-disk tribometer capable of operating at cryogenic temperatures, from 4.2 to 293 K. Two basic scenarios of testing were used: (1) frictional coefficient ({mu}) vs velocity (v) characteristics at constant temperatures; (2) {mu} vs temperature (T) behavior at fixed sliding speeds. For diamond/CVD-diamond, graphite/CVD-diamond, stainless steel/CVD-diamond pairs, {mu}`s are virtually velocity independent. For each of diamond/graphite, alumina/graphite, and graphite/graphite pairs, the {partial_derivative}{mu}/{partial_derivative}v characteristic is favorable, i.e., positive. For diamond/CVD-diamond and graphite/CVD-diamond pairs, {mu}`s are nearly temperature independent between in the range 77 - 293 K. Each {mu} vs T plot for pin materials sliding on graphite disks has a peak at a temperature in the range 100 - 200 K.

  1. Examining graphite reinforcement in composites

    NASA Technical Reports Server (NTRS)

    Sanders, R. E.; Yates, C. I.

    1980-01-01

    Structure of graphite layers in composite parts can be checked by pyrolizing epoxy portion of composite samples. After 2-3 hours in nitrogen atmosphere at 540 C, only graphite fibers remain. These can be separated and checked for proper number, thickness, and orientation.

  2. Multiaxial graphite test specimen

    SciTech Connect

    1988-09-01

    A multiaxial test program is to be conducted by Oak Ridge National Laboratory (ORNL) on the core component graphite. The objectives of the tests are to obtain failure data under uniaxial and biaxial states of stress in order to construct a failure surface in a two-dimensional stress space. These data will be used in verifying the accuracy of the maximum stress failure theory being proposed for use in designing the core graphite components. Tubular specimens are proposed to be used and are either loaded axially and/or subjected to internal pressure. This report includes a study on three specimen configurations. The conclusions of that study indicate that an elliptical transition geometry procedures the smallest discontinuity effects. Several loading combustions were studied using the elliptical transition specimen. The primary purpose is to establish the location of the highest stress state and its relation to the gage section for all of the loading conditions. The tension/internal pres sure loading condition (1:1) indicated that the high stress area is just outside the gage section but still should be acceptable. 5 refs., 18 figs.

  3. 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.

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

    SciTech Connect

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

    2014-10-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-10-01

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

  7. 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

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

    SciTech Connect

    Dunzik-Gougar, Mary Lou; Cleaver, James; LaBrier, Daniel; McCrory, Shilo; Smith, Tara E.

    2013-07-01

    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. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from some nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research presented last year and updated here is to identify the chemical form of C-14 in irradiated graphite and develop a practical method by which C-14 can be removed. A nuclear-grade graphite, NBG-18, and a high-surface-area graphite foam, POCOFoam{sup R}, were exposed to liquid nitrogen (to increase the quantity of C-14 precursor) and neutron-irradiated (10{sup 13} neutrons/cm{sup 2}/s). Finer grained NBG-25 was not exposed to liquid nitrogen prior to irradiation at a neutron flux on the order of 10{sup 14} /cm{sup 2}/s. Characterization of pre- and post-irradiation graphite was conducted to determine the chemical environment and quantity of C-14 and its precursors via the use of surface sensitive characterization techniques. Scanning Electron Microscopy (SEM) was used to evaluate the morphological features of graphite samples. The concentration, chemical composition, and bonding characteristics of C-14 and its precursors were determined through X-ray Photoelectron Spectroscopy (XPS), Time-of-Flight Secondary Ion Mass Spectrometry (SIMS), and Energy Dispersive X-ray Analysis Spectroscopy (EDX). Results of post-irradiation characterization of these materials indicate a variety of surface functional groups containing carbon, oxygen, nitrogen and hydrogen. During thermal treatment, irradiated graphite samples are heated in the presence of an inert carrier gas (with or without oxidant gas), which

  9. 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

    2015-07-01

    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).

  10. The molten salt reactor adventure

    SciTech Connect

    MacPherson, H.G.

    1985-08-01

    A personal history of the development of molten salt reactors in the United States is presented. The initial goal was an aircraft propulsion reactor, and a molten fluoride-fueled Aircraft Reactor Experiment was operated at Oak Ridge National Laboratory in 1954. In 1956, the objective shifted to civilian nuclear power, and reactor concepts were developed using a circulating UF4-ThF4 fuel, graphite moderator, and Hastelloy N pressure boundary. The program culminated in the successful operation of the Molten Salt Reactor Experiment in 1965 to 1969. By then the Atomic Energy Commission's goals had shifted to breeder development; the molten salt program supported on-site reprocessing development and study of various reactor arrangements that had potential to breed. Some commercial and foreign interest contributed to the program which, however, was terminated by the government in 1976. The current status of the technology and prospects for revived interest are summarized.

  11. Recent Advances in Preparation, Structure, Properties and Applications of Graphite Oxide.

    PubMed

    Srivastava, Suneel Kumar; Pionteck, Jürgen

    2015-03-01

    Graphite oxide, also referred as graphitic oxide or graphitic acid, is an oxidized bulk product of graphite with a variable composition. However, it did not receive immense attention until it was identified as an important and easily obtainable precursor for the preparation of graphene. This inspired many researchers to explore facts related to graphite oxide in exploiting its fascinating features. The present article culminates up-dated review on different preparative methods, morphology and characterization of physical/chemical properties of graphite oxide by XRD, XPS, FTIR, Raman, NMR, UV-visible, and DRIFT analyses. Finally, recent developments on intercalation and applications of GO in multifaceted areas of catalysis, sensor, supercapacitors, water purification, hydrogen storage and magnetic shielding etc. has also been reviewed.

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

    DOEpatents

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

    1995-07-04

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

  13. Tuning graphitic oxide for initiator- and metal-free aerobic epoxidation of linear alkenes

    NASA Astrophysics Data System (ADS)

    Pattisson, Samuel; Nowicka, Ewa; Gupta, Upendra N.; Shaw, Greg; Jenkins, Robert L.; Morgan, David J.; Knight, David W.; Hutchings, Graham J.

    2016-09-01

    Graphitic oxide has potential as a carbocatalyst for a wide range of reactions. Interest in this material has risen enormously due to it being a precursor to graphene via the chemical oxidation of graphite. Despite some studies suggesting that the chosen method of graphite oxidation can influence the physical properties of the graphitic oxide, the preparation method and extent of oxidation remain unresolved for catalytic applications. Here we show that tuning the graphitic oxide surface can be achieved by varying the amount and type of oxidant. The resulting materials differ in level of oxidation, surface oxygen content and functionality. Most importantly, we show that these graphitic oxide materials are active as unique carbocatalysts for low-temperature aerobic epoxidation of linear alkenes in the absence of initiator or metal. An optimum level of oxidation is necessary and materials produced via conventional permanganate-based methods are far from optimal.

  14. Tuning graphitic oxide for initiator- and metal-free aerobic epoxidation of linear alkenes

    PubMed Central

    Pattisson, Samuel; Nowicka, Ewa; Gupta, Upendra N.; Shaw, Greg; Jenkins, Robert L.; Morgan, David J.; Knight, David W.; Hutchings, Graham J.

    2016-01-01

    Graphitic oxide has potential as a carbocatalyst for a wide range of reactions. Interest in this material has risen enormously due to it being a precursor to graphene via the chemical oxidation of graphite. Despite some studies suggesting that the chosen method of graphite oxidation can influence the physical properties of the graphitic oxide, the preparation method and extent of oxidation remain unresolved for catalytic applications. Here we show that tuning the graphitic oxide surface can be achieved by varying the amount and type of oxidant. The resulting materials differ in level of oxidation, surface oxygen content and functionality. Most importantly, we show that these graphitic oxide materials are active as unique carbocatalysts for low-temperature aerobic epoxidation of linear alkenes in the absence of initiator or metal. An optimum level of oxidation is necessary and materials produced via conventional permanganate-based methods are far from optimal. PMID:27687877

  15. Chemical Characterization and Removal of C-14 from Irradiated Graphite-12010

    SciTech Connect

    Cleaver, James; McCrory, Shilo; Smith, Tara E.; Dunzik-Gougar, Mary Lou

    2012-07-01

    Quantities of irradiated graphite waste are expected to drastically increase, which indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research described here is to identify the chemical form of C-14 in irradiated graphite and develop a practical method by which C-14 can be removed. Characterization of pre- and post-irradiation graphite was conducted to determine bond type, functional groups, location and concentration of C-14 and its precursors via the use of surface sensitive characterization techniques. Because most surface C-14 originates from neutron activation of nitrogen, an understanding of nitrogen bonding to graphite may lead to a greater understanding of the formation pathway of C-14. However, no single technique provides a complete picture. Therefore, a portfolio of techniques has been developed, with each technique providing another piece to the puzzle that is the chemical nature of the C-14. Scanning Electron Microscopy (SEM), X-Ray Diffraction (XRD), and Raman Spectroscopy were used to evaluate the morphological features of graphite samples. The concentration, chemical composition, and bonding characteristics of C-14 and its precursors were determined through X-ray Photoelectron Spectroscopy (XPS), Time-of-Flight Secondary Ion Mass Spectrometry (SIMS), and Auger and Energy Dispersive X-ray Analysis Spectroscopy (EDX). High-surface-area graphite foam, POCOFoam{sup R}, was exposed to liquid nitrogen and irradiated. Characterization of this material has shown C-14 to C-12 ratios of 0.035. This information was used to optimize the thermal treatment of graphite. Thermal treatment of irradiated graphite as reported by Fachinger et al. (2007) uses naturally adsorbed oxygen complexes to

  16. 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)

  17. The Effect of Neutron Irradiation on the Fracture Toughness of Graphite

    SciTech Connect

    Burchell, Timothy D; Strizak, Joe P

    2012-01-01

    As part of our irradiated graphite recycle program a small quantity of PCEA grade graphite was irradiated in the High Flux Isotope Reactor (HFIR) at ORNL. The graphite will provide the raw material for future recycle experiments. The geometry of the irradiated graphite allowed us to study the effects of neutron irradiation on the Critical Stress Intensity Factor, KIc, of graphite. The specimens where irradiated in two groups of 6 at an irradiation temperature of 900 C in rabbit capsules to doses of 6.6 and 10.2 DPA, respectively. Following a full suite of pre-and post-irradiation examination, which included dimensions, mass, electrical resistivity, elastic constants, and thermal expansion (to 800 C) the samples were notched and tested to determine their KIc using the newly approved ATSM test method for SENB fracture toughness of graphite. Here we report the irradiation induced changes in the dimensions, elastic constants, resistivity, and coefficient of thermal expansion of PCEA graphite. Moreover, irradiation induced changes in the Critical Stress Intensity Factor, KIc, or fracture toughness, are reported and discussed. Very little work on the effect of neutron irradiation on the fracture toughness of graphite has previously be performed or reported.

  18. Interaction of boron with graphite: A van der Waals density functional study

    NASA Astrophysics Data System (ADS)

    Liu, Juan; Wang, Chen; Liang, Tongxiang; Lai, Wensheng

    2016-08-01

    Boron doping has been widely investigated to improve oxidation resistance of graphite. In this work the interaction of boron with graphite is investigated by a van der Waals density-functional approach (vdW-DF). The traditional density-functional theory (DFT) is well accounted for the binding in boron-substituted graphite. However, to investigate the boron atom on graphite surface and the interstitial impurities require use of a description of graphite interlayer binding. Traditional DFT cannot describe the vdW physics, for instance, GGA calculations show no relevant binding between graphite sheets. LDA shows some binding, but they fail to provide an accurate account of vdW forces. In this paper, we compare the calculation results of graphite lattice constant and cohesive energy by several functionals, it shows that vdW-DF such as two optimized functionals optB88-vdW and optB86b-vdW give much improved results than traditional DFT. The vdW-DF approach is then applied to study the interaction of boron with graphite. Boron adsorption, substitution, and intercalation are discussed in terms of structural parameters and electronic structures. When adsorbing on graphite surface, boron behaves as π electron acceptor. The π electron approaches boron atom because of more electropositive of boron than carbon. For substitution situation, the hole introduced by boron mainly concentrates on boron and the nearest three carbon atoms. The B-doped graphite system with the hole has less ability to offer electrons to oxygen, ultimately resulted in the inhibition of carbon oxidation. For interstitial doping, vdW-DFs show more accurate formation energy than LDA. PBE functional cannot describe the interstitial boron in graphite reasonably because of the ignoring binding of graphite sheets. The investigation of electron structures of boron doped graphite will play an important role in understanding the oxidation mechanism in further study.

  19. Imaging of tritium implanted into graphite

    SciTech Connect

    Malinowski, M.E.; Causey, R.A.

    1988-05-01

    The extensive use of graphite in plasma-facing surfaces of tokamaks such as the Tokamak Fusion Test Reactor, which has planned tritium discharges, makes two-dimensional tritium detection techniques important in helping to determine torus tritium inventories. We have performed experiments in which highly oriented pyrolytic graphite (HOPG) samples were first tritium implanted with fluences of approx.10/sup 16/ T/cm/sup 2/ at energies approx. <25 eV and then the near-surface implant distributions were detected in two dimensions using tritium imaging. A portion of the sample was masked off during the implant in order to produce a well-defined implant boundary. Heating of the HOPG samples to temperatures as high as 500 /sup 0/C resulted in no discernible motion of tritium along the basal plane, but did show that significant desorption of the implanted tritium occurred. The current results indicate that tritium in quantities of 10/sup 12/ T/cm/sup 2/ in tritiated components could be readily detected by imaging at lower magnifications.

  20. 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.

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

    SciTech Connect

    Pinto, A.P.

    1994-08-01

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

  2. 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

  3. Density of intercalated graphite fibers

    NASA Technical Reports Server (NTRS)

    Gaier, James R.; Slabe, Melissa E.

    1990-01-01

    The density of Amoco P-55, P-75, P-100, and P-120 pitch-based graphite fibers and their intercalation compounds with bromine, iodine monochloride, and copper (II) chloride have been measured using a density gradient column. The distribution of densities within a fiber type is found to be a sensitive indicator of the quality of the intercalation reaction. In all cases the density was found to increase, indicating that the mass added to the graphite is dominant over fiber expansion. Density increases are small (less than 10 percent) giving credence to a model of the intercalated graphite fibers which have regions which are intercalated and regions which are not.

  4. Density of intercalated graphite fibers

    NASA Technical Reports Server (NTRS)

    Gaier, James R.; Slabe, Melissa E.

    1989-01-01

    The density of Amoco P-55, P-75, P-100, and P-120 pitch-based graphite fibers and their intercalation compounds with bromine, iodine monochloride, and copper (II) chloride have been measured using a density gradient column. The distribution of densities within a fiber type is found to be a sensitive indicator of the quality of the intercalation reaction. In all cases the density was found to increase, indicating that the mass added to the graphite is dominant over fiber expansion. Density increases are small (less than 10 percent) giving credence to a model of the intercalated graphite fibers which have regions which are intercalated and regions which are not.

  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. 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. PMID:27573503

  7. 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.

  8. 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.

  9. METHOD OF OBTAINING UNIFORM COATINGS ON GRAPHITE

    DOEpatents

    Campbell, I.E.

    1961-04-01

    A method is given for obtaining uniform carbide coatings on graphite bodies. According to the invention a metallic halide in vapor form is passed over the graphite body under such conditions of temperature and pressure that the halide reacts with the graphite to form a coating of the metal carbide on the surface of the graphite.

  10. Method of Obtaining Uniform Coatings on Graphite

    DOEpatents

    Campbell, I. E.

    1961-04-01

    A method is given for obtaining uniform carbide coatings on graphite bodies. According to the invention a metallic halide in vapor form is passed over the graphite body under such conditions of temperature and pressure that the halide reacts with the graphite to form a coating of the metal carbide on the surface of the graphite.

  11. Mineral resource of the month: graphite

    USGS Publications Warehouse

    ,

    2008-01-01

    The article presents facts about graphite ideal for industrial applications. Among the characteristics of graphite are its metallic luster, softness, perfect basal cleavage and electrical conductivity. Batteries, brake linings and powdered metals are some of the products that make use of graphite. It attributes the potential applications for graphite in high-technology fields to innovations in thermal technology and acid-leaching techniques.

  12. 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)

  13. 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

  14. 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)

  15. Graphite-reinforced bone cement

    NASA Technical Reports Server (NTRS)

    Knoell, A. C.

    1976-01-01

    Chopped graphite fibers added to surgical bone cement form bonding agent with mechanical properties closely matched to those of bone. Curing reaction produces less heat, resulting in reduced traumatization of body tissues. Stiffness is increased without affecting flexural strength.

  16. Intercalated hybrid graphite fiber composite

    NASA Technical Reports Server (NTRS)

    Gaier, James R. (Inventor)

    1993-01-01

    The invention is directed to a highly conductive lightweight hybrid material and methods of producing the same. The hybrid composite is obtained by weaving strands of a high strength carbon or graphite fiber into a fabric-like structure, depositing a layer of carbon onto the structure, heat treating the structure to graphitize the carbon layer, and intercalating the graphitic carbon layer structure. A laminate composite material useful for protection against lightning strikes comprises at least one layer of the hybrid material over at least one layer of high strength carbon or graphite fibers. The composite material of the present invention is compatible with matrix compounds, has a coefficient of thermal expansion which is the same as underlying fiber layers, and is resistant to galvanic corrosion in addition to being highly conductive. These materials are useful in the aerospace industry, in particular as lightning strike protection for airplanes.

  17. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... integration of systems, technologies, programs, equipment, supporting processes, and implementing...

  18. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... integration of systems, technologies, programs, equipment, supporting processes, and implementing...

  19. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... integration of systems, technologies, programs, equipment, supporting processes, and implementing...

  20. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... access into the protected area, the licensee shall: (i) Confirm the identity of individuals. (ii)...

  1. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... access into the protected area, the licensee shall: (i) Confirm the identity of individuals. (ii)...

  2. Study of neutron physics: conversion of the University of Missouri-Rolla reactor to low-enriched fuel

    SciTech Connect

    Straka, M.; Covington, L.

    1987-01-01

    A detailed study of a fuel conversion (using LEU) has been undertaken for the University of Missouri-Rolla reactor. Results achieved with the available code package have been compared with the measured data whenever possible. The neutronic codes LEOPARD and 2DB-UM provided adequate results in most cases examined.

  3. A physics study for negative void reactivity in compact supercritical CO{sub 2}-cooled fast reactor

    SciTech Connect

    Kim, Y.; Hartanto, D.; Lee, J. I.

    2013-07-01

    A compact S-CO{sub 2}-cooled fast reactor which has negative Coolant Void Reactivity (CVR) has been investigated. A negative CVR is important for the gas cooled fast reactor as an inherent safety mechanism to prevent the sudden positive reactivity insertion when the loss of coolant accident happens. An alternative solution to reduce the CVR is investigated in this study by using O-17 instead of O-16 in UO{sub 2} fuel. By using O-17 in the fuel, it is found that the CVR can even be negative. Impacts of the radial reflector on the CVR are also evaluated for the small SCO{sub 2} cooled fast reactor in this study. We have considered a pure lead (Pb) reflector and a lead magnesium eutectic (LME) reflector as alternative radial reflectors of the S-CO 2-cooled fast reactor. It has been shown that, with the LME radial reflector, the CVR can be negative, while the pure lead reflector provides a slightly positive CVR. (authors)

  4. Irradiation Induced Creep of Graphite

    SciTech Connect

    Burchell, Timothy D; Murty, Prof K.L.; Eapen, Dr. Jacob

    2010-01-01

    The current status of graphite irradiation induced creep strain prediction is reviewed and the major creep models are described. The ability of the models to quantitatively predict the irradiation induced creep strain of graphite is reported. Potential mechanisms of in-crystal creep are reviewed as are mechanisms of pore generation under stress. The case for further experimental work is made and the need for improved creep models across multi-scales is highlighted.

  5. 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.

  6. Transport of atomic hydrogen through graphite and its reaction with azoaromatic compounds.

    PubMed

    Ye, Jianchang; Chiu, Pei C

    2006-06-15

    Graphite is a major non-iron component in commercial iron granules that are typically used for groundwater remediation. Recent studies suggest graphite inclusions in commercial iron may serve as both adsorption and reaction sites for nitrogenous pollutants such as nitroaromatics, nitrate esters, and heterocyclic nitramines. In this study, we investigated graphite-mediated reduction of azoaromatic compounds with elemental iron in dialysis cells, where azo compounds and iron were physically separated by graphite foil. Both the nonpolar azobenzene and the water-soluble orange G were reduced to aniline, suggesting that exposed graphite in granular iron may mediate reduction of both polar and nonpolar compounds. Orange G reduction was zero-order and commenced after a long initial lag. Both the lag time and the zero-order rate constant varied with graphite thickness, consistent with the explanation that orange G reduction was limited by atomic hydrogen, which was formed via anaerobic iron corrosion and spilled over to graphite. Involvement of atomic hydrogen was confirmed by detection of deuterated aniline when iron was placed in a D2O-based buffer. Our results indicate that atomic hydrogen is mobile in graphite at room temperature, is reactive toward azoaromatic compounds, and may be consumed during transport in graphite.

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

    NASA Astrophysics Data System (ADS)

    Nyathi, Mhlwazi S.

    2011-12-01

    Graphite is utilized as a neutron moderator and structural component in some nuclear reactor designs. During the reactor operaction the structure of graphite is damaged by collision with fast neutrons. Graphite's resistance to this damage determines its lifetime in the reactor. On neutron irradiation, isotropic or near-isotropic graphite experiences less structural damage than anisotropic graphite. The degree of anisotropy in a graphite artifact is dependent on the structure of its precursor coke. Currently, there exist concerns over a short supply of traditional precursor coke, primarily due to a steadily increasing price of petroleum. The main goal of this study was to study the anisotropic and isotropic properties of graphitized co-cokes and anthracites as a way of investigating the possibility of synthesizing isotropic or near-isotropic graphite from co-cokes and anthracites. Demonstrating the ability to form isotropic or near-isotropic graphite would mean that co-cokes and anthracites have a potential use as filler material in the synthesis of nuclear graphite. The approach used to control the co-coke structure was to vary the reaction conditions. Co-cokes were produced by coking 4:1 blends of vacuum resid/coal and decant oil/coal at temperatures of 465 and 500 °C for reaction times of 12 and 18 hours under autogenous pressure. Co-cokes obtained were calcined at 1420 °C and graphitized at 3000 °C for 24 hours. Optical microscopy, X-ray diffraction, temperature-programmed oxidation and Raman spectroscopy were used to characterize the products. It was found that higher reaction temperature (500 °C) or shorter reaction time (12 hours) leads to an increase in co-coke structural disorder and an increase in the amount of mosaic carbon at the expense of textural components that are necessary for the formation of anisotropic structure, namely, domains and flow domains. Characterization of graphitized co-cokes showed that the quality, as expressed by the degree of

  8. Microstructural characterization and pore structure analysis of nuclear graphite

    NASA Astrophysics Data System (ADS)

    Kane, J.; Karthik, C.; Butt, D. P.; Windes, W. E.; Ubic, R.

    2011-08-01

    Graphite will be used as a structural and moderator material in next-generation nuclear reactors. While the overall nature of the production of nuclear graphite is well understood, the historic nuclear grades of graphite are no longer available. This paper reports the virgin microstructural characteristics of filler particles and macro-scale porosity in virgin nuclear graphite grades of interest to the Next Generation Nuclear Plant program. Optical microscopy was used to characterize filler particle size and shape as well as the arrangement of shrinkage cracks. Computer aided image analysis was applied to optical images to quantitatively determine the variation of pore structure, area, eccentricity, and orientation within and between grades. The overall porosity ranged between ˜14% and 21%. A few large pores constitute the majority of the overall porosity. The distribution of pore area in all grades was roughly logarithmic in nature. The average pore was best fit by an ellipse with aspect ratio of ˜2. An estimated 0.6-0.9% of observed porosity was attributed to shrinkage cracks in the filler particles. Finally, a preferred orientation of the porosity was observed in all grades.

  9. Microstructural Characterization and Pore Structure Analysis of Nuclear Graphite

    SciTech Connect

    J. Kane; C. Karthik; D. P. Butt; W. E. Windes; R. Ubic

    2011-08-01

    Graphite will be used as a structural and moderator material in next-generation nuclear reactors. While the overall nature of the production of nuclear graphite is well understood, the historic nuclear grades of graphite are no longer available. This paper reports the virgin microstructural characteristics of filler particles and macro-scale porosity in virgin nuclear graphite grades of interest to the Next Generation Nuclear Plant program. Optical microscopy was used to characterize filler particle size and shape as well as the arrangement of shrinkage cracks. Computer aided image analysis was applied to optical images to quantitatively determine the variation of pore structure, area, eccentricity, and orientation within and between grades. The overall porosity ranged between {approx}14% and 21%. A few large pores constitute the majority of the overall porosity. The distribution of pore area in all grades was roughly logarithmic in nature. The average pore was best fit by an ellipse with aspect ratio of {approx}2. An estimated 0.6-0.9% of observed porosity was attributed to shrinkage cracks in the filler particles. Finally, a preferred orientation of the porosity was observed in all grades.

  10. ICP-MS measurement of iodine diffusion in IG-110 graphite for HTGR/VHTR

    NASA Astrophysics Data System (ADS)

    Carter, L. M.; Brockman, J. D.; Robertson, J. D.; Loyalka, S. K.

    2016-05-01

    Graphite functions as a structural material and as a barrier to fission product release in HTGR/VHTR designs, and elucidation of transport parameters for fission products in reactor-grade graphite is thus required for reactor source terms calculations. We measured iodine diffusion in spheres of IG-110 graphite using a release method based on Fickain diffusion kinetics. Two sources of iodine were loaded into the graphite spheres; molecular iodine (I2) and cesium iodide (CsI). Measurements of the diffusion coefficient were made over a temperature range of 873-1293 K. We have obtained the following Arrhenius expressions for iodine diffusion:DI , CsI infused =(6 ×10-12 2/s) exp(30,000 J/mol RT) And,DI , I2 infused =(4 ×10-10 m2/s) exp(-11,000 J/mol RT ) The results indicate that iodine diffusion in IG-110 graphite is not well-described by Fickan diffusion kinetics. To our knowledge, these are the first measurements of iodine diffusion in IG-110 graphite.

  11. The Characterization of Grade PCEA Recycle Graphite Pilot Scale Billets

    SciTech Connect

    Burchell, Timothy D; Pappano, Peter J

    2010-10-01

    Here we report the physical properties of a series specimens machined from pilot scale (~ 152 mm diameter x ~305 mm length) grade PCEA recycle billets manufactured by GrafTech. The pilot scale billets were processed with increasing amounts of (unirradiated) graphite (from 20% to 100%) introduced to the formulation with the goal of determining if large fractions of recycle graphite have a deleterious effect on properties. The properties determined include Bulk Density, Electrical Resistivity, Elastic (Young s) Modulus, and Coefficient of Thermal Expansion. Although property variations were observed to be correlated with the recycle fraction, the magnitude of the variations was noted to be small.

  12. 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.

  13. Reinforcement of cement-based matrices with graphite nanomaterials

    NASA Astrophysics Data System (ADS)

    Sadiq, Muhammad Maqbool

    Cement-based materials offer a desirable balance of compressive strength, moisture resistance, durability, economy and energy-efficiency; their tensile strength, fracture energy and durability in aggressive environments, however, could benefit from further improvements. An option for realizing some of these improvements involves introduction of discrete fibers into concrete. When compared with today's micro-scale (steel, polypropylene, glass, etc.) fibers, graphite nanomaterials (carbon nanotube, nanofiber and graphite nanoplatelet) offer superior geometric, mechanical and physical characteristics. Graphite nanomaterials would realize their reinforcement potential as far as they are thoroughly dispersed within cement-based matrices, and effectively bond to cement hydrates. The research reported herein developed non-covalent and covalent surface modification techniques to improve the dispersion and interfacial interactions of graphite nanomaterials in cement-based matrices with a dense and well graded micro-structure. The most successful approach involved polymer wrapping of nanomaterials for increasing the density of hydrophilic groups on the nanomaterial surface without causing any damage to the their structure. The nanomaterials were characterized using various spectrometry techniques, and SEM (Scanning Electron Microscopy). The graphite nanomaterials were dispersed via selected sonication procedures in the mixing water of the cement-based matrix; conventional mixing and sample preparation techniques were then employed to prepare the cement-based nanocomposite samples, which were subjected to steam curing. Comprehensive engineering and durability characteristics of cement-based nanocomposites were determined and their chemical composition, microstructure and failure mechanisms were also assessed through various spectrometry, thermogravimetry, electron microscopy and elemental analyses. Both functionalized and non-functionalized nanomaterials as well as different

  14. Energy corrugation in atomic-scale friction on graphite revisited by molecular dynamics simulations

    NASA Astrophysics Data System (ADS)

    Sun, Xiao-Yu; Qi, Yi-Zhou; Ouyang, Wengen; Feng, Xi-Qiao; Li, Qunyang

    2016-08-01

    Although atomic stick-slip friction has been extensively studied since its first demonstration on graphite, the physical understanding of this dissipation-dominated phenomenon is still very limited. In this work, we perform molecular dynamics (MD) simulations to study the frictional behavior of a diamond tip sliding over a graphite surface. In contrast to the common wisdom, our MD results suggest that the energy barrier associated lateral sliding (known as energy corrugation) comes not only from interaction between the tip and the top layer of graphite but also from interactions among the deformed atomic layers of graphite. Due to the competition of these two subentries, friction on graphite can be tuned by controlling the relative adhesion of different interfaces. For relatively low tip-graphite adhesion, friction behaves normally and increases with increasing normal load. However, for relatively high tip-graphite adhesion, friction increases unusually with decreasing normal load leading to an effectively negative coefficient of friction, which is consistent with the recent experimental observations on chemically modified graphite. Our results provide a new insight into the physical origins of energy corrugation in atomic scale friction.

  15. Use of high-volume outdoor smog chamber photo-reactors for studying physical and chemical atmospheric aerosol formation and composition

    NASA Astrophysics Data System (ADS)

    Borrás, E.; Ródenas, M.; Vera, T.; Muñoz, A.

    2015-12-01

    The atmospheric particulate matter has a large impact on climate, biosphere behaviour and human health. Its study is complex because of large number of species are present at low concentrations and the continuous time evolution, being not easily separable from meteorology, and transport processes. Closed systems have been proposed by isolating specific reactions, pollutants or products and controlling the oxidizing environment. High volume simulation chambers, such as EUropean PHOtoREactor (EUPHORE), are an essential tool used to simulate atmospheric photochemical reactions. This communication describes the last results about the reactivity of prominent atmospheric pollutants and the subsequent particulate matter formation. Specific experiments focused on organic aerosols have been developed at the EUPHORE photo-reactor. The use of on-line instrumentation, supported by off-line techniques, has provided well-defined reaction profiles, physical properties, and up to 300 different species are determined in particulate matter. The application fields include the degradation of anthropogenic and biogenic pollutants, and pesticides under several atmospheric conditions, studying their contribution on the formation of secondary organic aerosols (SOA). The studies performed at the EUPHORE have improved the mechanistic studies of atmospheric degradation processes and the knowledge about the chemical and physical properties of atmospheric particulate matter formed during these processes.

  16. Ferric chloride graphite intercalation compounds prepared from graphite fluoride

    NASA Technical Reports Server (NTRS)

    Hung, Ching-Cheh

    1994-01-01

    The reaction between graphite fluoride and ferric chloride was observed in the temperature range of 300 to 400 C. The graphite fluorides used for this reaction have an sp3 electronic structure and are electrical insulators. They can be made by fluorinating either carbon fibers or powder having various degrees of graphitization. Reaction is fast and spontaneous and can occur in the presence of air. The ferric chloride does not have to be predried. The products have an sp2 electronic structure and are electrical conductors. They contain first stage FeCl3 intercalated graphite. Some of the products contain FeCl2*2H2O, others contain FeF3 in concentrations that depend on the intercalation condition. The graphite intercalated compounds (GIC) deintercalated slowly in air at room temperature, but deintercalated quickly and completely at 370 C. Deintercalation is accompanied by the disappearing of iron halides and the formation of rust (hematite) distributed unevenly on the fiber surface. When heated to 400 C in pure N2 (99.99 vol %), this new GIC deintercalates without losing its molecular structure. However, when the compounds are heated to 800 C in quartz tube, they lost most of its halogen atoms and formed iron oxides (other than hematite), distributed evenly in or on the fiber. This iron-oxide-covered fiber may be useful in making carbon-fiber/ceramic-matrix composites with strong bonding at the fiber-ceramic interface.

  17. Calculation of 3s photoemission spectra of vanadium on graphite

    SciTech Connect

    Krueger, P.; Taguchi, M.; Parlebas, J.C.; Kotani, A.

    1997-06-01

    A few years ago, a satellite structure in the vanadium 3s x-ray photoemission spectroscopy (XPS) spectrum of V clusters upon graphite was observed and attributed to the presence of magnetic moments on the V surface. Here, we present calculations of these spectra using a cluster model that takes into account intra-atomic d-d and d{endash}core electron correlation and hybridization between V d and graphite {pi} states. When the V-graphite distance is increased from 1.5 to 2.0 {Angstrom} the system undergoes a low-to-high spin transition, which is clearly evidenced in the evolution of the XPS line shape. Although direct comparison with experiment is difficult, our study suggests that the observed satellite is due to core hole screening rather than a magnetic moment on the V atom. {copyright} {ital 1997} {ital The American Physical Society}

  18. PMR Graphite Engine Duct Development

    NASA Technical Reports Server (NTRS)

    Stotler, C. L.; Yokel, S. A.

    1989-01-01

    The objective was to demonstrate the cost and weight advantages that could be obtained by utilizing the graphite/PMR15 material system to replace titanium in selected turbofan engine applications. The first component to be selected as a basis for evaluation was the outer bypass duct of the General Electric F404 engine. The operating environment of this duct was defined and then an extensive mechanical and physical property test program was conducted using material made by processing techniques which were also established by this program. Based on these properties, design concepts to fabricate a composite version of the duct were established and two complete ducts fabricated. One of these ducts was proof pressure tested and then run successfully on a factory test engine for over 1900 hours. The second duct was static tested to 210 percent design limit load without failure. An improved design was then developed which utilized integral composite end flanges. A complete duct was fabricated and successfully proof pressure tested. The net results of this effort showed that a composite version of the outer duct would be 14 percent lighter and 30 percent less expensive that the titanium duct. The other type of structure chosen for investigation was the F404 fan stator assembly, including the fan stator vanes. It was concluded that it was feasible to utilize composite materials for this type structure but that the requirements imposed by replacing an existing metal design resulted in an inefficient composite design. It was concluded that if composites were to be effectively used in this type structure, the design must be tailored for composite application from the outset.

  19. Fluorine interaction with defects on graphite surface by a first-principles study

    NASA Astrophysics Data System (ADS)

    Wang, Song; Xuezhi, Ke; Zhang, Wei; Gong, Wenbin; Huai, Ping; Zhang, Wenqing; Zhu, Zhiyuan

    2014-02-01

    The interaction between fluorine atom and graphite surface has been investigated in the framework of density functional theory. Due to the consideration of molten salt reactor system, only carbon adatoms and vacancies are chemical reactive for fluorine atoms. Fluorine adsorption on carbon adatom will enhance the mobility of carbon adatom. Carbon adatom can also be removed easily from graphite surface in form of CF2 molecule, explaining the formation mechanism of CF2 molecule in previous experiment. For the interaction between fluorine and vacancy, we find that fluorine atoms which adsorb at vacancy can hardly escape. Both pristine surface and vacancy are impossible for fluorine to penetrate due to the high penetration barrier. We believe our result is helpful to understand the compatibility between graphite and fluorine molten salt in molten salt reactor system.

  20. Residual stress measurements in polycrystalline graphite with micro-Raman spectroscopy

    NASA Astrophysics Data System (ADS)

    Krishna, Ram; Jones, Abbie N.; Edge, Ruth; Marsden, Barry J.

    2015-06-01

    Micro-Raman microscopy technique is applied to evaluate unevenly distributed residual stresses in the various constituents of polygranular reactor grades graphite. The wavenumber based Raman shift (cm-1) corresponds to the local residual stress and measurements of stress dependent first order Raman spectra in graphite have enabled localized residual stress values to be determined. The bulk polygranular graphite of reactor grades - Gilsocarbon, NBG-18 and PGA - are examined to illustrate the residual stress variations in their constituents. Binder phase and filler particles have shown to be under compressive and tensile stresses, respectively. Among the studied graphite grades, the binder phase in Gilsocarbon has the highest residual stress and NBG-18 has the lowest value. Filler particles in Gilsocarbon have the highest residual stress and PGA showed the lowest, this is most likely due to the morphology of the coke particles used in the manufacturing and applied processing techniques for fabrications. Stresses have also been evaluated along the peripheral of pores and at the tips of the cracks. Cracks in filler and binder phases have shown mixed behaviour, compressive as well as tensile, whereas pores in binder and filler particles have shown compressive behaviour. The stresses in these graphitic constituents are of the order of MPa. Non-destructive analyses presented in this study make the current state-of-the-art technique a powerful method for the study of stress variations near the graphite surface and are expected to increase its use further in property determination analysis of low to highly fluence irradiated graphite samples from the material test reactors.

  1. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    NASA Astrophysics Data System (ADS)

    Cisneros, Anselmo Tomas, Jr.

    and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of

  2. Destruction of nuclear graphite using closed chamber incineration

    SciTech Connect

    Senor, D.J.; Hollenberg, G.W.; Morgan, W.C.; Marianowski, L.G.

    1994-03-01

    Closed chamber incineration (CCI) is a novel technique by which irradiated nuclear graphite may be destroyed without the risk of radioactive cation release into the environment. The process utilizes an enclosed combustion chamber coupled with molten carbonate fuel cells (MCFCs). The transport of cations is intrinsically suppressed by the MCFCs, such that only the combustion gases are conducted through for release to the environment. An example CCI design was developed which had as its goal the destruction of graphite fuel elements from the Fort St. Vrain reactor (FSVR). By employing CCI, the volume of high level waste from the FSVR will be reduced by approximately 87 percent. Additionally, the incineration process will convert the SiC coating on the FSVR fuel particles to SiO{sub 2}, thus creating a form potentially suitable for direct incorporation in a vitrification process stream. The design is compact, efficient, and makes use of currently available technology.

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

    SciTech Connect

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

    2013-06-12

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

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

    NASA Astrophysics Data System (ADS)

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

    2013-06-01

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

  5. Graphite Formation in Cast Iron

    NASA Technical Reports Server (NTRS)

    Stefanescu, D. M.

    1985-01-01

    In the first phase of the project it was proven that by changing the ratio between the thermal gradient and the growth rate for commercial cast iron samples solidifying in a Bridgman type furnace, it is possible to produce all types of graphite structures, from flake to spheroidal, and all types of matrices, from ferritic to white at a certain given level of cerium. KC-135 flight experiments have shown that in a low-gravity environment, no flotation occurs even in spheroidal graphite cast irons with carbon equivalent as high as 5%, while extensive graphite flotation occurred in both flake and spheroidal graphite cast irons, in high carbon samples solidified in a high gravity environment. This opens the way for production of iron-carbon composite materials, with high carbon content (e.g., 10%) in a low gravity environment. By using KC-135 flights, the influence of some basic elements on the solidification of cast iron will be studied. The mechanism of flake to spheroidal graphite transition will be studied, by using quenching experiments at both low and one gravity for different G/R ratios.

  6. Structural, chemical, and isotopic microanalytical investigations of graphite from supernovae

    NASA Astrophysics Data System (ADS)

    Croat, T. Kevin; Bernatowicz, Thomas; Amari, Sachiko; Messenger, Scott; Stadermann, Frank J.

    2003-12-01

    value of 0.122. Significant variations about the mean V/Ti ratio were also seen among TiCs in the same graphite, likely indicating chemical equilibration with the surrounding gas over a range of temperatures. In general, the diversity in internal TiC properties suggests that TiCs formed first and had substantially diverse histories before incorporation into the graphite, implying some degree of turbulent mixing in the SN outflows. In most graphites, there is a decrease in the number density of TiCs as a function of increasing radial dis- tance, caused by either preferential depletion of TiCs from the gas or an acceleration of graphite growth with decreasing ambient temperature. In several graphites, TiCs showed a trend of larger V/Ti ratios with increasing distance from the graphite center, an indication of progressive equilibration with the surrounding gas before they were sequestered in the graphites. In all but one graphite, no trend was seen in the TiC size vs. distance from the graphite center, implying that appreciable TiC growth had effectively stopped before the graphites formed, or else that graphite growth was rapid compared to TiC growth. Taken together, the chemical variations among internal grains as well as the presence of partially amorphous rims and epitaxial Fe phases on some TiCs clearly indicate that the phase condensation sequence was TiC, followed by the iron phases (only found in some graphites) and finally graphite. Since graphite typically condenses at a higher temperature than iron at low pressures (<10 -3 bars) in a gas with C > O and otherwise solar composition, the observed condensation sequence implies a relative iron enrichment in the gas or greater supersaturation of graphite relative to iron. The TEM observations allow inferences to be made about the physical conditions in the gas from which the grains condensed. Given the TiC sizes and abundances, the gas was evidently quite dusty. From the observed TiC size range of ˜20 nm to ˜500 nm

  7. Structural, chemical, and isotopic microanalytical investigations of graphite from supernovae

    NASA Astrophysics Data System (ADS)

    Croat, T. Kevin; Bernatowicz, Thomas; Amari, Sachiko; Messenger, Scott; Stadermann, Frank J.

    2003-12-01

    value of 0.122. Significant variations about the mean V/Ti ratio were also seen among TiCs in the same graphite, likely indicating chemical equilibration with the surrounding gas over a range of temperatures. In general, the diversity in internal TiC properties suggests that TiCs formed first and had substantially diverse histories before incorporation into the graphite, implying some degree of turbulent mixing in the SN outflows. In most graphites, there is a decrease in the number density of TiCs as a function of increasing radial dis- tance, caused by either preferential depletion of TiCs from the gas or an acceleration of graphite growth with decreasing ambient temperature. In several graphites, TiCs showed a trend of larger V/Ti ratios with increasing distance from the graphite center, an indication of progressive equilibration with the surrounding gas before they were sequestered in the graphites. In all but one graphite, no trend was seen in the TiC size vs. distance from the graphite center, implying that appreciable TiC growth had effectively stopped before the graphites formed, or else that graphite growth was rapid compared to TiC growth. Taken together, the chemical variations among internal grains as well as the presence of partially amorphous rims and epitaxial Fe phases on some TiCs clearly indicate that the phase condensation sequence was TiC, followed by the iron phases (only found in some graphites) and finally graphite. Since graphite typically condenses at a higher temperature than iron at low pressures (<10 -3 bars) in a gas with C > O and otherwise solar composition, the observed condensation sequence implies a relative iron enrichment in the gas or greater supersaturation of graphite relative to iron. The TEM observations allow inferences to be made about the physical conditions in the gas from which the grains condensed. Given the TiC sizes and abundances, the gas was evidently quite dusty. From the observed TiC size range of ˜20 nm to ˜500 nm

  8. NGNP Graphite Testing and Qualification Specimen Selection Strategy

    SciTech Connect

    Robert Bratton

    2005-05-01

    The available grades of graphite for the NGNP are reviewed. A selection matrix is presented outlining the available grades for the NGNP graphite irradiation program based upon input from potential NGNP vendors, graphite manufactures, and graphite experts.

  9. Thermal Pyrolytic Graphite Enhanced Components

    NASA Technical Reports Server (NTRS)

    Hardesty, Robert E. (Inventor)

    2015-01-01

    A thermally conductive composite material, a thermal transfer device made of the material, and a method for making the material are disclosed. Apertures or depressions are formed in aluminum or aluminum alloy. Plugs are formed of thermal pyrolytic graphite. An amount of silicon sufficient for liquid interface diffusion bonding is applied, for example by vapor deposition or use of aluminum silicon alloy foil. The plugs are inserted in the apertures or depressions. Bonding energy is applied, for example by applying pressure and heat using a hot isostatic press. The thermal pyrolytic graphite, aluminum or aluminum alloy and silicon form a eutectic alloy. As a result, the plugs are bonded into the apertures or depressions. The composite material can be machined to produce finished devices such as the thermal transfer device. Thermally conductive planes of the thermal pyrolytic graphite plugs may be aligned in parallel to present a thermal conduction path.

  10. Design analysis of the upgraded TREAT reactor

    SciTech Connect

    Bhattacharyya, S.K.

    1982-01-01

    The TREAT reactor, fueled by a dilute dispersion of fully enriched UO/sub 2/ in graphite, has been a premier transient testing facility since 1959. A major Upgrade of the reactor is in progress to enhance its transient testing capability in support of the LMFBR safety program. The TREAT Upgrade (TU) reactor features a modified central zone of the core with higher fissile loadings of the same fuel, clad in Inconel to allow operation at higher temperatures. The demanding functional requirements on the reactor necessitated the use of unique features in the core design which, in turn, presented major calculational complexities in the analysis. Special design methods had to be used in many cases to treat these complexities. The addition of an improved Reactor Control System, a safety grade Plant Protection System and an enhanced Coolant/Filtration System produces a reactor that can meet the functional requirements on the reactor in a safe manner.

  11. Graphite filament wound pressure vessels

    NASA Technical Reports Server (NTRS)

    Feldman, A.; Damico, J. J.

    1972-01-01

    Filament wound NOL rings, 4-inch and 8-inch diameter closed-end vessels involving three epoxy resin systems and three graphite fibers were tested to develop property data and fabrication technology for filament wound graphite/epoxy pressure vessels. Vessels were subjected to single-cycle burst tests at room temperature. Manufacturing parameters were established for tooling, winding, and curing that resulted in the development of a pressure/vessel performance factor (pressure x volume/weight) or more than 900,000 in. for an oblate spheroid specimen.

  12. Graphite fiber reinforced thermoplastic resins

    NASA Technical Reports Server (NTRS)

    Novak, R. C.

    1975-01-01

    Mechanical properties of neat resin samples and graphite fiber reinforced samples of thermoplastic resins were characterized with particular emphasis directed to the effects of environmental exposure (humidity, temperature and ultraviolet radiation). Tensile, flexural, interlaminar shear, creep and impact strengths were measured for polysulfone, polyarylsulfone and a state-of-the-art epoxy resin samples. In general, the thermoplastic resins exhibited environmental degradation resistance equal to or superior to the reference epoxy resin. Demonstration of the utility and quality of a graphite/thermoplastic resin system was accomplished by successfully thermoforming a simulated compressor blade and a fan exit guide vane.

  13. Pyrolytic graphite collector development program

    NASA Technical Reports Server (NTRS)

    Wilkins, W. J.

    1982-01-01

    Pyrolytic graphite promises to have significant advantages as a material for multistage depressed collector electrodes. Among these advantages are lighter weight, improved mechanical stiffness under shock and vibration, reduced secondary electron back-streaming for higher efficiency, and reduced outgassing at higher operating temperatures. The essential properties of pyrolytic graphite and the necessary design criteria are discussed. This includes the study of suitable electrode geometries and methods of attachment to other metal and ceramic collector components consistent with typical electrical, thermal, and mechanical requirements.

  14. Jacobian-free Newton Krylov discontinuous Galerkin method and physics-based preconditioning for nuclear reactor simulations

    SciTech Connect

    HyeongKae Park; Robert R. Nourgaliev; Richard C. Martineau; Dana A. Knoll

    2008-09-01

    We present high-order accurate spatiotemporal discretization of all-speed flow solvers using Jacobian-free Newton Krylov framework. One of the key developments in this work is the physics-based preconditioner for the all-speed flow, which makes use of traditional semi-implicit schemes. The physics-based preconditioner is developed in the primitive variable form, which allows a straightforward separation of physical phenomena. Numerical examples demonstrate that the developed preconditioner effectively reduces the number of the Krylov iterations, and the efficiency is independent of the Mach number and mesh sizes under a fixed CFL condition.

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

    SciTech Connect

    Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

    2012-10-01

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

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

    DOEpatents

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

    2012-09-25

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

  17. ANALYSIS OF WIGNER ENERGY IN BGRR GRAPHITE.

    SciTech Connect

    FUHRMANN, M.

    2006-10-31

    Wigner Energy was determined by DSC analysis in cored graphite from the Brookhaven Graphite Research Reactor. Eight segments (4-inch long slugs) of cores were obtained from BGRR for analysis of Wigner Energy retained in the graphite. Graphite was scraped from each end of each slug giving two samples from each specimen. Between 10 and 20 mg of this graphite powder were weighed into platinum analysis cells and subjected to thermal analysis on a Shimadzu Differential Scanning Calorimeter (DSC-50). The samples were annealed in nitrogen up to 700 C at a scan rate of 20 C/minute with data recorded at one second intervals. Each sample was run twice; the first scan provided the energy profile of the ''as received'' material and the second scan provided the background energy profile of the specimen, as the Wigner Energy had been removed during the first annealing. An example is shown in Figure 1. The blank was subtracted from the initial scan to give the Wigner energy profile. The appendix contains two graphs for each sample. One graph presents the data in J/s/g and shows the results of the two scans described above; the energy measurement of the ''as received'' and the same sample after annealing. The other graph presents the data in J/g/K, which was calculated by subtracting the background scan data from the first scan and dividing by the heating rate. The heating rate was nominally 20 K /minute (0.333K/s), however regression analysis provided a more accurate heating rate of 0.3506 K/s and this was used to determine J/g/K. These values were plotted against temperature in C. From these plots the temperature at which energy release increases can be determined. The data (J/s/g) were summed providing a measure of total Wigner energy in the sample in Joules per gram. The DSC analysis gives energy content of the graphite that ranges from around 0 (actual measurements of samples from Loc 4 slug 3 were -2.9 and -21.2 J/g) to 212 J/g. Table 1 gives the location and the total

  18. Composition and method for brazing graphite to graphite

    DOEpatents

    Taylor, Albert J.; Dykes, Norman L.

    1984-01-01

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

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

    SciTech Connect

    Chang H. Oh; Eung S. Kim

    2009-12-01

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

  20. RECOVERY OF VALUABLE MATERIAL FROM GRAPHITE BODIES

    DOEpatents

    Fromm, L.W. Jr.

    1959-09-01

    An electrolytic process for recovering uranium from a graphite fuel element is described. The uraniumcontaining graphite body is disposed as the anode of a cell containing a nitric acid electrolyte and a 5 amp/cm/sup 2/ current passed to induce a progressive disintegration of the graphite body. The dissolved uranium is quickly and easily separated from the resulting graphite particles by simple mechanical means, such as centrifugation, filtration, and decontamination.

  1. Method of Joining Graphite Fibers to a Substrate

    NASA Technical Reports Server (NTRS)

    Beringer, Durwood M. (Inventor); Caron, Mark E. (Inventor); Taddey, Edmund P. (Inventor); Gleason, Brian P. (Inventor)

    2014-01-01

    A method of assembling a metallic-graphite structure includes forming a wetted graphite subassembly by arranging one or more layers of graphite fiber material including a plurality of graphite fibers and applying a layer of metallization material to ends of the plurality of graphite fibers. At least one metallic substrate is secured to the wetted graphite subassembly via the layer of metallization material.

  2. High speed hydrogen/graphite interaction

    NASA Technical Reports Server (NTRS)

    Kelly, A. J.; Hamman, R.; Sharma, O. P.; Harrje, D. T.

    1974-01-01

    Various aspects of a research program on high speed hydrogen/graphite interaction are presented. Major areas discussed are: (1) theoretical predictions of hydrogen/graphite erosion rates; (2) high temperature, nonequilibrium hydrogen flow in a nozzle; and (3) molecular beam studies of hydrogen/graphite erosion.

  3. Separation medium containing thermally exfoliated graphite oxide

    NASA Technical Reports Server (NTRS)

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

    2012-01-01

    A separation medium, such as a chromatography filling or packing, containing a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 m.sup.2/g to 2600 m.sup.2/g, wherein the thermally exfoliated graphite oxide has a surface that has been at least partially functionalized.

  4. CMB-13 research on carbon and graphite

    NASA Technical Reports Server (NTRS)

    Smith, M. C.

    1972-01-01

    The research on graphite and carbon for this period is reported. Topics discussed include: effects of grinding on the Santa Marie graphites, properties and purities of coal-tar, resin-bonded graphite, carbonization of resin components, and glass-like carbon filler.

  5. 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.

  6. AN INTEGRAL REACTOR PHYSICS EXPERIMENT TO INFER ACTINIDE CAPTURE CROSS-SECTIONS FROM THORIUM TO CALIFORNIUM WITH ACCELERATOR MASS SPECTROMETRY

    SciTech Connect

    G. Youinou; M. Salvatores; M. Paul; R. Pardo; G. Palmiotti; F. Kondev; G. Imel

    2010-04-01

    The principle of the proposed experiment is to irradiate very pure actinide samples in the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation will allow inference of effective neutron capture cross-sections. This approach has been used in the past and the novelty of this experiment is that the atom densities of the different transmutation products will be determined using the Accelerator Mass Spectroscopy (AMS) technique at the ATLAS facility located at ANL. It is currently planned to irradiate the following isotopes: 232Th, 235U, 236U, 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am and 248Cm.

  7. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel, Progress Report for Work through August 31, 2002, First Annual/4th Quarterly Report

    SciTech Connect

    Anderson, William J.; Ake, Timothy N.; Punatar, Mahendra; Pitts, Michelle L.; Harms, Gary A.; Rearden, Bradley T.; Parks, Cecil V.; Tulenko, James S.; Dugan, Edward; Smith, Robert M.

    2002-09-23

    OAK B204 The objective of this Nuclear Energy Research Initiative (NERI) project is to design, perform, and analyze critical benchmark experiments for validating reactor physics methods and models for fuel enrichments greater than 5-wt% 235U. These experiments will also provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5-wt% 235U fuel. These experiments are designed as reactor physics benchmarks, to include measurements of critical boron concentration, burnable absorber worth, relative pin powers, and relative average powers.The first year focused primarily on designing the experiments using available fuel, preparing the necessary plans, procedures and authorization basis for performing the experiments, and preparing for the transportation, receipt and storage of the Pathfinder fuel currently stored at Pennsylvania State University.Framatome ANP, Inc. leads the project with the collaboration of Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and the University of Florida (UF). The project is organized into 5 tasks:Task 1: Framatome ANP, Inc., ORNL, and SNL will design the specific experiments, establish the safety authorization, and obtain approvals to perform these experiments at the SNL facility. ORNL will apply their sensitivity/uncertainty methodology to verify the need for particular experiments and the parameters that these experiments need to explore.Task 2: Framatome ANP, Inc., ORNL, and UF will analyze the proposed experiments using a variety of reactor-physics methods employed in the nuclear industry. These analyses will support the operation of the experiments by predicting the expected experimental values for the criticality and physics parameters.Task 3: This task encompasses the experiments to be performed. The Pathfinder fuel will be transported from Penn State to SNL for use in the experiments. The experiments will be performed and the

  8. Management of Spent Nuclear Fuel of Nuclear Research Reactor VVR-S at the National Institute of Physics and Nuclear Engineering, Bucharest, Romania

    NASA Astrophysics Data System (ADS)

    Biro, Lucian

    2009-05-01

    The Nuclear Research Reactor VVR-S (RR-VVR-S) located in Magurele-Bucharest, Romania, was designed for research and radioisotope production. It was commissioned in 1957 and operated without any event or accident for forty years until shut down in 1997. In 2002, by government decree, it was permanently shutdown for decommissioning. The National Institute of Physics and Nuclear Engineering (IFIN-HH) is responsible for decommissioning the RR-VVR-S, the first nuclear decommissioning project in Romania. In this context, IFIN-HH prepared and obtained approval from the Romanian Nuclear Regulatory Body for the Decommissioning Plan. One of the most important aspects for decommissioning the RR-VVR-S is solving the issue of the fresh and spent nuclear fuel (SNF) stored on site in wet storage pools. In the framework of the Russian Research Reactor Fuel Return Program (RRRFR), managed by the U.S. Department of Energy and in cooperation with the International Atomic Energy Agency and the Rosatom State Corporation, Romania repatriated all fresh HEU fuel to the Russian Federation in 2003 and the HEU SNF will be repatriated to Russia in 2009. With the experience and lessons learned from this action and with the financial support of the Romanian Government it will be possible for Romania to also repatriate the LEU SNF to the Russian Federation before starting the dismantling and decontamination of the nuclear facility. [4pt] In collaboration with K. Allen, Idaho National Laboratory, USA; L. Biro, National Commission for Nuclear Activities Control, Romania; and M. Dragusin, National Institute of Physics and Nuclear Engineering, Bucharest-Magurele, Romania.

  9. Graphite oral tattoo: case report.

    PubMed

    Moraes, Renata Mendonça; Gouvêa Lima, Gabriela de Morais; Guilhermino, Marinaldo; Vieira, Mayana Soares; Carvalho, Yasmin Rodarte; Anbinder, Ana Lia

    2015-10-16

    Pigmented oral lesions compose a large number of pathological entities, including exogenous pigmentat oral tattoos, such as amalgam and graphite tattoos. We report a rare case of a graphite tattoo on the palate of a 62-year-old patient with a history of pencil injury, compare it with amalgam tattoos, and determine the prevalence of oral tattoos in our Oral Pathology Service. We also compare the clinical and histological findings of grafite and amalgam tattoos. Oral tattoos affect women more frequently in the region of the alveolar ridge. Graphite tattoos occur in younger patients when compared with the amalgam type. Histologically, amalgam lesions represent impregnation of the reticular fibers of vessels and nerves with silver, whereas in cases of graphite tattoos, this impregnation is not observed, but it is common to observe a granulomatous inflammatory response, less evident in cases of amalgam tattoos. Both types of lesions require no treatment, but in some cases a biopsy may be done to rule out melanocytic lesions.

  10. Igneous Graphite in Enstatite Chondrites

    NASA Technical Reports Server (NTRS)

    Rubin, Alan E.

    1997-01-01

    Igneous graphite. a rare constituent in terrestrial mafic and ultramafic rocks. occurs in three EH and one EL enstatite chondrite impact-melt breccias as 2-150 Ilm long euhedrallaths. some with pyramidal terminations. In contrast. graphite in most enstatite chondrites exsolved from metallic Fe-Ni as polygonal. rounded or irregular aggregates. Literature data for five EH chondrites on C combusting at high temperatures show that Abee contains the most homogeneous C isotopes (i.e. delta(sup 13)C = -8.1+/-2.1%); in addition. Abee's mean delta(sup l3)C value is the same as the average high-temperature C value for the set of five EH chondrites. This suggests that Abee scavenged C from a plurality of sources on its parent body and homogenized the C during a large-scale melting event. Whereas igneous graphite in terrestrial rocks typically forms at relatively high pressure and only moderately low oxygen fugacity (e.g., approx. 5 kbar. logfO2, approx. -10 at 1200 C ). igneous graphite in asteroidal meteorites formed at much lower pressures and oxygen fugacities.

  11. Fracture toughness of anisotropic graphites

    SciTech Connect

    Kennedy, C.R.; Kehne, M.T.

    1985-01-01

    Fracture toughness measurements have been made at 0, 30, 45, 60, and 90/sup 0/ from the extrusion axis on a reasonably anisotropic graphite, grade AGOT. It was found that the fracture toughness did not vary appreciably with orientation. An observed variation in strength was found to be the result of defect orientation.

  12. Graphite oral tattoo: case report.

    PubMed

    Moraes, Renata Mendonça; Gouvêa Lima, Gabriela de Morais; Guilhermino, Marinaldo; Vieira, Mayana Soares; Carvalho, Yasmin Rodarte; Anbinder, Ana Lia

    2015-10-01

    Pigmented oral lesions compose a large number of pathological entities, including exogenous pigmentat oral tattoos, such as amalgam and graphite tattoos. We report a rare case of a graphite tattoo on the palate of a 62-year-old patient with a history of pencil injury, compare it with amalgam tattoos, and determine the prevalence of oral tattoos in our Oral Pathology Service. We also compare the clinical and histological findings of grafite and amalgam tattoos. Oral tattoos affect women more frequently in the region of the alveolar ridge. Graphite tattoos occur in younger patients when compared with the amalgam type. Histologically, amalgam lesions represent impregnation of the reticular fibers of vessels and nerves with silver, whereas in cases of graphite tattoos, this impregnation is not observed, but it is common to observe a granulomatous inflammatory response, less evident in cases of amalgam tattoos. Both types of lesions require no treatment, but in some cases a biopsy may be done to rule out melanocytic lesions. PMID:26632800

  13. Direct Preparation of Few Layer Graphene Epoxy Nanocomposites from Untreated Flake Graphite.

    PubMed

    Throckmorton, James; Palmese, Giuseppe

    2015-07-15

    The natural availability of flake graphite and the exceptional properties of graphene and graphene-polymer composites create a demand for simple, cost-effective, and scalable methods for top-down graphite exfoliation. This work presents a novel method of few layer graphite nanocomposite preparation directly from untreated flake graphite using a room temperature ionic liquid and laminar shear processing regimen. The ionic liquid serves both as a solvent and initiator for epoxy polymerization and is incorporated chemically into the matrix. This nanocomposite shows low electrical percolation (0.005 v/v) and low thickness (1-3 layers) graphite/graphene flakes by TEM. Additionally, the effect of processing conditions by rheometry and comparison with solvent-free conditions reveal the interactions between processing and matrix properties and provide insight into the theory of the chemical and physical exfoliation of graphite crystals and the resulting polymer matrix dispersion. An interaction model that correlates the interlayer shear physics of graphite flakes and processing parameters is proposed and tested.

  14. Reactor neutrino monitoring

    NASA Astrophysics Data System (ADS)

    Lhuillier, D.

    2009-03-01

    Nuclear reactors are the most intense man-controlled sources of antineutrinos and as such have hosted number of key physics experiments, from the antineutrino discovery to modern oscillation measurements. At the present time, both detection technology and understanding of fundamental physics are mature enough to think about antineutrinos as a new tool for reactor monitoring. We describe below how antineutrinos can provide online information on reactor operation and amount of plutonium accumulated in the core. Reactors are the only sources of plutonium on earth and this element can be chemically separated from the rest of the nuclear fuel and diverted into nuclear weapons. We present in the next sections the unique features antineutrino detectors could provide to safeguards agencies such as IAEA. We review the worldwide efforts to develop small ( 1m scale) antineutrino detectors dedicated to automated and non-intrusive reactor monitoring.

  15. Physics.

    ERIC Educational Resources Information Center

    Bromley, D. Allan

    1980-01-01

    The author presents the argument that the past few years, in terms of new discoveries, insights, and questions raised, have been among the most productive in the history of physics. Selected for discussion are some of the most important new developments in physics research. (Author/SA)

  16. Investigating the effects of stress on the pore structures of nuclear grade graphites

    NASA Astrophysics Data System (ADS)

    Taylor, Joshua E. L.; Hall, Graham N.; Mummery, Paul M.

    2016-03-01

    Graphite is used as a moderating material and as a structural component in a number of current generation nuclear reactors. During reactor operation stresses develop in the graphite components, causing them to deform. It is important to understand how the microstructure of graphite affects the material's response to these stresses. A series of experiments were performed to investigate how the pore structures of Pile Grade A and Gilsocarbon graphites respond to loading stresses. A compression rig was used to simulate the build-up of operational stresses in graphite components, and a confocal laser microscope was used to study variation of a number of important pore properties. Values of elastic modulus and Poisson's ratio were calculated and compared to existing literature to confirm the validity of the experimental techniques. Mean pore areas were observed to decrease linearly with increasing applied load, mean pore eccentricity increased linearly, and a small amount of clockwise pore rotation was observed. The response to build-up of stresses was dependent on the orientation of the pores and basal planes and the shapes of the pores with respect to the loading axis. It was proposed that pore closure and pore reorientation were competing processes. Pore separation was quantified using 'nearest neighbour' and Voronoi techniques, and non-pore regions were found to shrink linearly with increasing applied load.

  17. HTR Fuel Waste Management: TRISO separation and acid-graphite intercalation compounds preparation

    NASA Astrophysics Data System (ADS)

    Guittonneau, Fabrice; Abdelouas, Abdesselam; Grambow, Bernd

    2010-12-01

    Considering the need to reduce waste production and greenhouse emissions and still keeping high energy efficiency, various 4th generation nuclear energy systems have been proposed. As far as graphite-moderated reactors are concerned (future high temperature fast or thermal reactors), one of the key issues is the large volumes of irradiated graphite encountered. With the objective to reduce volume of waste in the HTR concept, it is very important to be able to separate the fuel from low level activity graphite representing a large volume. The separated TRISO particles can then be reprocessed for waste separation or disposed off in geological repository. In addition, preparation of acid-GICs from the separated graphite may constitute a way to recycle this waste. We used HTR-type compact fuel with ZrO 2 TRISO particles to test two separation methods: low (H 2SO 4 + H 2O 2) and high (H 2SO 4 + HNO 3) temperature acid treatments. In both cases the TRISO separation was complete but some TRISO layers oxidized at high temperature. At low temperature, the desegregation of graphite grains is facilitated by intercalation of sulfuric acid between the graphene layers. The acid-GIC obtained consists of pure phases of high quality suggesting their potential industrial recycling.

  18. Graphite nanoreinforcements in polymer nanocomposites

    NASA Astrophysics Data System (ADS)

    Fukushima, Hiroyuki

    Nanocomposites composed of polymer matrices with clay reinforcements of less than 100 nm in size, are being considered for applications such as interior and exterior accessories for automobiles, structural components for portable electronic devices, and films for food packaging. While most nanocomposite research has focused on exfoliated clay platelets, the same nanoreinforcement concept can be applied to another layered material, graphite, to produce nanoplatelets and nanocomposites. Graphite is the stiffest material found in nature (Young's Modulus = 1060 GPa), having a modulus several times that of clay, but also with excellent electrical and thermal conductivity. The key to utilizing graphite as a platelet nanoreinforcement is in the ability to exfoliate this material. Also, if the appropriate surface treatment can be found for graphite, its exfoliation and dispersion in a polymer matrix will result in a composite with not only excellent mechanical properties but electrical properties as well, opening up many new structural applications as well as non-structural ones where electromagnetic shielding and high thermal conductivity are requirements. In this research, a new process to fabricate exfoliated nano-scale graphite platelets was established (Patent pending). The size of the resulted graphite platelets was less than 1 um in diameter and 10 nm in thickness, and the surface area of the material was around 100 m2/g. The reduction of size showed positive effect on mechanical properties of composites because of the increased edge area and more functional groups attached with it. Also various surface treatment techniques were applied to the graphite nanoplatelets to improve the surface condition. As a result, acrylamide grafting treatment was found to enhance the dispersion and adhesion of graphite flakes in epoxy matrices. The resulted composites showed better mechanical properties than those with commercially available carbon fibers, vapor grown carbon fibers

  19. JANUS reactor d and d project.

    SciTech Connect

    Fellhauer, C. R.

    1998-02-16

    Argonne National Laboratory (ANL-E) has recently completed the decontamination and decommissioning (D and D) of the JANUS Reactor Facility located in Building 202. The 200 KW reactor operated from August 1963 to March 1992. The facility was used to study the effects of both high and low doses of fission neutrons in animals. There were two exposure rooms on opposite sides of the reactor and the reactor was therefore named after the two-faced Roman god. The High Dose Room was capable of specimen exposure at a dose rate of 3,600 rads per hour. During calendar year 1996 a detailed characterization of the facility was performed by ANL-E Health Physics personnel. ANL-E Analytical Services performed the required sample analysis. An Auditable Safety Analysis and an Environmental Assessment were completed. D and D plans, procedures and procurement documents were prepared and approved. A D and D subcontractor was selected and a firm, fixed price contract awarded for the field work and final survey effort. The D and D subcontractor was mobilized to ANL-E in January 1997. Electrical isolation of all reactor equipment and control panels was accomplished and the equipment removed. A total of 207,230 pounds (94,082 Kg) of lead shielding was removed, surveyed and sampled, and free-released for recycle. All primary and secondary piping was removed, size reduced and packaged for disposal or recycled as appropriate. The reactor vessel was removed, sized reduced and packaged as radioactive waste in April. The activated graphite block reflector was removed next, followed by the bioshield concrete and steel. All of this material was packaged as low level waste. Total low level radioactive waste generation was 4002.1 cubic feet (113.3 cubic meters). Mixed waste generation was 538 cubic feet (15.2 cubic meters). The Final Release Survey was completed in September. The project field work was completed in 38 weeks without any lost-time accidents, personnel contaminations or unplanned

  20. Research on graphite reinforced glass matrix composites

    NASA Technical Reports Server (NTRS)

    Bacon, J. F.; Prewo, K. M.

    1977-01-01

    The results of research for the origination of graphite-fiber reinforced glass matrix composites are presented. The method selected to form the composites consisted of pulling the graphite fiber through a slurry containing powdered glass, winding up the graphite fiber and the glass it picks up on a drum, drying, cutting into segments, loading the tape segment into a graphite die, and hot pressing. During the course of the work, composites were made with a variety of graphite fibers in a glass matrix.