Science.gov

Sample records for structural reactor components

  1. Impact of conversion to mixed-oxide fuels on reactor structural components

    SciTech Connect

    Yahr, G.T.

    1997-04-01

    The use of mixed-oxide (MOX) fuel to replace conventional uranium fuel in commercial light-water power reactors will result in an increase in the neutron flux. The impact of the higher flux on the structural integrity of reactor structural components must be evaluated. This report briefly reviews the effects of radiation on the mechanical properties of metals. Aging degradation studies and reactor operating experience provide a basis for determining the areas where conversion to MOX fuels has the potential to impact the structural integrity of reactor components.

  2. Structural Integrity of Water Reactor Pressure Boundary Components.

    DTIC Science & Technology

    1980-08-01

    tests of reference steels of the NRC light water reactor, pressure vessel irradiation dosimetry program. SECURITY CLAS5IICATION 0PHiS PA6GMbn" Dfat ...multiple specimen R- curve approach; NRL emphasis was on the SSC procedure as it is being developed for hot- cell testing of irradiated materials. MULTIPLE...a second autoclave, capable of testing 50 or 100 mm (2T or 4T) thick CT or WOL specimens, was installed in a hot cell and a test was started on 2T-CT

  3. Structural Integrity of Water Reactor Pressure Boundary Components.

    DTIC Science & Technology

    1981-02-20

    RES-79-103 UNCLASSIFIED NRL--- 400 NURE-CR-17B3 NL mnmmnuunin -’El-.--. IIIIIIINI ., *q. - - ,aM T? * NUREG /CI 73 NIL Iteof AW, SOIituA 1 nert of Water...Progress Report for July-September 1979," NUREG /CR-1197, Oak Ridge National Labora- tory, Oak Ridge, Tn., Oct. 1978. 2. F. J. Loss, Ed., "Structural...Progress Report for April-June 1976," ORNL/ NUREG /TM-49, Oak Ridge National Labora- tory, Oak Ridge, Tn., Oct. 1976, pp. 27-38. 5. R. G. Berggren

  4. Reactor component automatic grapple

    SciTech Connect

    Greenaway, P.R.

    1982-12-07

    A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment.

  5. Reactor component automatic grapple

    DOEpatents

    Greenaway, Paul R.

    1982-01-01

    A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment.

  6. Evaluation of hot isostatic pressing for joining of fusion reactor structural components

    NASA Astrophysics Data System (ADS)

    Ivanov, A. D.; Sato, S.; Le Marois, G.

    2000-12-01

    Hot isostatic pressing (HIP) is a promising technology to fabricate the blanket structure of fusion reactors. HIP joining of solid materials has been selected as a reference fabrication method for the shielding blanket/first wall of the international thermonuclear experimental reactor (ITER). On the basis of experimental results obtained in Europe, Japan and Russia, an evaluation of HIP joining for fusion reactor structural components has been carried out. The parameters of HIP fabrication for copper alloys and stainless steels are given. The results of microscopic observations, X-ray microanalysis, tensile, impact toughness, fracture toughness and fatigue tests are presented. Material science criteria for an estimation of quality for joints fabricated by HIP are discussed.

  7. Fundamental Understanding of Crack Growth in Structural Components of Generation IV Supercritical Light Water Reactors

    SciTech Connect

    Iouri I. Balachov; Takao Kobayashi; Francis Tanzella; Indira Jayaweera; Palitha Jayaweera; Petri Kinnunen; Martin Bojinov; Timo Saario

    2004-11-17

    This work contributes to the design of safe and economical Generation-IV Super-Critical Water Reactors (SCWRs) by providing a basis for selecting structural materials to ensure the functionality of in-vessel components during the entire service life. During the second year of the project, we completed electrochemical characterization of the oxide film properties and investigation of crack initiation and propagation for candidate structural materials steels under supercritical conditions. We ranked candidate alloys against their susceptibility to environmentally assisted degradation based on the in situ data measure with an SRI-designed controlled distance electrochemistry (CDE) arrangement. A correlation between measurable oxide film properties and susceptibility of austenitic steels to environmentally assisted degradation was observed experimentally. One of the major practical results of the present work is the experimentally proven ability of the economical CDE technique to supply in situ data for ranking candidate structural materials for Generation-IV SCRs. A potential use of the CDE arrangement developed ar SRI for building in situ sensors monitoring water chemistry in the heat transport circuit of Generation-IV SCWRs was evaluated and proved to be feasible.

  8. End-of-life irradiation performance of core structural components in the Shippingport Light Water Breeder Reactor

    SciTech Connect

    Clayton, J.C.; Smith, B.C.

    1991-12-31

    Nondestructive and destructive end-of-life examinations of Light Water Breeder Reactor (LWBR) core structural components were performed following operation in the Shippingport Atomic Power Station for 29,047 effective full power hours. The Shippingport LWBR demonstrated that breeding can be achieved in a light water reactor with thorium and uranium-233 oxide fuel pellets contained in Zircaloy-4 tubes. The purpose of this presentation is to report results of LWBR core structural component examinations that were carried out to assess the effects of irradiation on support structure and to provide a data base for the evaluation of design procedures. The postirradiation nondestructive examinations included visual inspection and, in some cases, dye penetrant testing to assess structural integrity and surface conditions of the components. Destructive metallography was performed to assess cracking, corrosion buildup, and microstructural condition.

  9. Application of two component biodegradable carriers in a particle-fixed biofilm airlift suspension reactor: development and structure of biofilms.

    PubMed

    Hille, Andrea; He, Mei; Ochmann, Clemens; Neu, Thomas R; Horn, Harald

    2009-01-01

    Two component biodegradable carriers for biofilm airlift suspension (BAS) reactors were investigated with respect to development of biofilm structure and oxygen transport inside the biofilm. The carriers were composed of PHB (polyhydroxybutyrate), which is easily degradable and PCL (caprolactone), which is less easily degradable by heterotrophic microorganisms. Cryosectioning combined with classical light microscopy and CLSM was used to identify the surface structure of the carrier material over a period of 250 days of biofilm cultivation in an airlift reactor. Pores of 50 to several hundred micrometers depth are formed due to the preferred degradation of PHB. Furthermore, microelectrode studies show the transport mechanism for different types of biofilm structures, which were generated under different substrate conditions. At high loading rates, the growth of a rather loosely structured biofilm with high penetration depths of oxygen was found. Strong changes of substrate concentration during fed-batch mode operation of the reactor enhance the growth of filamentous biofilms on the carriers. Mass transport in the outer regions of such biofilms was mainly driven by advection.

  10. Lagrangian three-dimensional finite-element formulation for the nonlinear fluid-structural response of reactor components. [LMFBR

    SciTech Connect

    Kulak, R. F.; Fiala, C.

    1980-03-01

    This report presents the formulations used in the NEPTUNE code. Specifically, it describes the finite-element formulation of a three-dimensional hexahedral element for simulating the behavior of either fluid or solid continua. Since the newly developed hexahedral element and the original triangular plate element are finite elements, they are compatible in the sense that they can be combined arbitrarily to simulate complex reactor components in three-dimensional space. Because rate-type constitutive relations are used in conjunction with a velocity-strain tensor, the formulation is applicable to large deformation problems. This development can be used to simulate (1) the fluid adjacent to reactor components and (2) the concrete fill found in large reactor head closures.

  11. Nonlinear seismic analysis of a reactor structure impact between core components

    NASA Technical Reports Server (NTRS)

    Hill, R. G.

    1975-01-01

    The seismic analysis of the FFTF-PIOTA (Fast Flux Test Facility-Postirradiation Open Test Assembly), subjected to a horizontal DBE (Design Base Earthquake) is presented. The PIOTA is the first in a set of open test assemblies to be designed for the FFTF. Employing the direct method of transient analysis, the governing differential equations describing the motion of the system are set up directly and are implicitly integrated numerically in time. A simple lumped-nass beam model of the FFTF which includes small clearances between core components is used as a "driver" for a fine mesh model of the PIOTA. The nonlinear forces due to the impact of the core components and their effect on the PIOTA are computed.

  12. Development of Nb-1%Zr-0.1%C alloy as structural components for high temperature reactors

    NASA Astrophysics Data System (ADS)

    Vishwanadh, B.; Vaibhav, K.; Jha, S. K.; Mirji, K. V.; Samajdar, I.; Srivastava, D.; Tewari, R.; Saibaba, N.; Dey, G. K.

    2012-08-01

    The Nb-1Zr-0.1C (wt.%) alloy is being considered for structural components in the proposed Compact High-Temperature-Reactors (HTR). The present work reports on the development of 30-50 kg ingots of the alloy in correct composition as well as technology for forming the material in various shapes. The work deals with the deformation behavior of as-cast material at different temperatures and strain rates, recrystallization behavior at different temperature and time and evolution of microstructures at different processing conditions (as-cast, deformed and recrystallized). The as-cast Nb alloys were deformed up to 35% at different temperatures. The deformation results showed that the flow stress of the as-cast Nb alloy increases with increasing temperature from 800 °C to 1000 °C. Beyond 1200 °C, substantial decrease in the strength of the alloy was noticed. To determine the optimum recrystallization temperature and time for the alloy, several heat treatments were conducted by systematically varying temperature and time. It was found that the deformed Nb alloy could be recrystallized by annealing at 1300 °C for 3 h. The microstructures of the as-cast, deformed and recrystallized samples of Nb-1%Zr-0.1%C alloy were systematically characterized by optical, electron back scattered diffraction (EBSD) and transmission electron microscopy techniques. The Nb-1Zr-0.1C alloy showed significant differences in the microstructure after different thermo-mechanical treatments. Microstructures of the Nb alloy showed two phases: the matrix (bcc) phase and the carbide phase. Electron Microscopy and energy dispersive spectroscopic analyses revealed that the carbide precipitation undergoes various phase transformations. The as-cast structure of Nb alloy had hexagonal Nb2C precipitates in the Nb matrix and after extrusion, the deformed microstructure had two types of carbide precipitates: needle and rectangular morphology precipitates. The needle shape precipitates were of (Nb, Zr)2C

  13. Energy deposition in STARFIRE reactor components

    SciTech Connect

    Gohar, Y.; Brooks, J.N.

    1985-04-01

    The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry.

  14. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    SciTech Connect

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

  15. Mechanical cutting of irradiated reactor internal components

    SciTech Connect

    Anderson, Michael G.

    2008-01-15

    Mechanical cutting methods to volume reduce and package reactor internal components are now a viable solution for stakeholders challenged with the retirement of first generation nuclear facilities. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods, inclusive of plasma arc and abrasive water-jet cutting, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. Reactor internal components were segmented, packaged, and removed from the reactor building for shipment or storage, allowing the reactor cavity to be drained and follow-on reactor segmentation activities to proceed in the dry state. Area exposure rates at the work positions during the segmentation process were generally 1 mR per hr. Radiological exposure documented for the underwater segmentation processes totaled 13 person rem. The reactor internals weighing 343,000 pounds were segmented into over 200 pieces for maximum shipping package efficiency and produced 5,600 lb of stainless steel chips and shavings which were packaged in void spaces of existing disposal containers, therefore creating no additional disposal volume. Because no secondary waste was driven into suspension in the reactor cavity water, the water was free released after one pass through a charcoal bed and ion exchange filter system. Mechanical cutting techniques are capable of underwater segmentation of highly radioactive components on a large scale. This method minimized radiological exposure and costly water cleanup while creating no secondary waste.

  16. Protective structures on the surface of zirconium components of light water reactor cores: Formation, testing, and prototype equipment

    SciTech Connect

    Begrambekov, L. B.; Gordeev, A. A.; Evsin, A. E. Ivanova, S. V.; Kaplevsky, A. S.; Sadovskiy, Ya. A.

    2015-12-15

    The results of tests of plasma treatment of zirconium and deposition of protective yttrium coatings used as the methods of protection of zirconium components of light water reactor cores against hydrogenation are detailed. The amount of hydrogen in the treated sample exposed to superheated steam for 2500 h at temperature T = 400°C and pressure p = 1 atm was five times lower than the corresponding value for the untreated one. The amount of hydrogen in the sample coated with yttrium remained almost unchanged in 4000 h of exposure. A plasma method for rapid testing for hydrogen resistance is proposed. The hydrogenation rate provided by this method is 700 times higher than that in tests with superheated steam. The results of preliminary experiments confirm the possibility of constructing a unit for batch processing of the surfaces of fuel rod claddings.

  17. Protective structures on the surface of zirconium components of light water reactor cores: Formation, testing, and prototype equipment

    NASA Astrophysics Data System (ADS)

    Begrambekov, L. B.; Gordeev, A. A.; Evsin, A. E.; Ivanova, S. V.; Kaplevsky, A. S.; Sadovskiy, Ya. A.

    2015-12-01

    The results of tests of plasma treatment of zirconium and deposition of protective yttrium coatings used as the methods of protection of zirconium components of light water reactor cores against hydrogenation are detailed. The amount of hydrogen in the treated sample exposed to superheated steam for 2500 h at temperature T = 400°C and pressure p = 1 atm was five times lower than the corresponding value for the untreated one. The amount of hydrogen in the sample coated with yttrium remained almost unchanged in 4000 h of exposure. A plasma method for rapid testing for hydrogen resistance is proposed. The hydrogenation rate provided by this method is 700 times higher than that in tests with superheated steam. The results of preliminary experiments confirm the possibility of constructing a unit for batch processing of the surfaces of fuel rod claddings.

  18. Mechanical Cutting of Irradiated Reactor Internal Components

    SciTech Connect

    Anderson, M.G.; Fennema, J.A.

    2007-07-01

    This paper discusses the use of mechanical cutting methods to volume reduce and package irradiated reactor internal components. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods used for similar projects, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. (authors)

  19. Heavy Water Components Test Reactor Decommissioning - Major Component Removal

    SciTech Connect

    Austin, W.; Brinkley, D.

    2010-05-05

    The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these

  20. REACTOR MODERATOR STRUCTURE

    DOEpatents

    Fraas, A.P.; Tudor, J.J.

    1963-08-01

    An improved moderator structure for nuclear reactors consists of moderator blocks arranged in horizontal layers to form a multiplicity of vertically stacked columns of blocks. The blocks in each vertical column are keyed together, and a ceramic grid is disposed between each horizontal layer of blocks. Pressure plates cover- the lateral surface of the moderator structure in abutting relationship with the peripheral terminal lengths of the ceramic grids. Tubular springs are disposed between the pressure plates and a rigid external support. The tubular springs have their axes vertically disposed to facilitate passage of coolant gas through the springs and are spaced apart a selected distance such that at sonae preselected point of spring deflection, the sides of the springs will contact adjacent springs thereby causing a large increase in resistance to further spring deflection. (AEC)

  1. HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

    SciTech Connect

    Austin, W.; Brinkley, D.

    2011-10-13

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  2. NEUTRONIC REACTOR STRUCTURE

    DOEpatents

    Daniels, F.

    1961-10-24

    A reactor core, comprised of vertical stacks of hexagonal blocks of beryllium oxide having axial cylindrical apertures extending therethrough and cylindrical rods of a sintered mixture of uranium dioxide and beryllium oxide, is described. (AEC)

  3. Neutronic Reactor Structure

    DOEpatents

    Vernon, H. C.; Weinberg, A. M.

    1961-05-30

    The neutronic reactor is comprised of a core consisting of natural uranium and heavy water with a K-factor greater than unity. The core is surrounded by a reflector consisting of natural uranium and ordinary water with a Kfactor less than unity. (AEC)

  4. NEUTRONIC REACTOR STRUCTURE

    DOEpatents

    Weinberg, A.M.; Vernon, H.C.

    1961-05-30

    A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.

  5. Preliminary study on nano- and micro-composite sol-gel based alumina coatings on structural components of lead-bismuth eutectic cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Dou, Peng; Kasada, Ryuta

    2011-02-01

    In order to protect the structural components of lead-bismuth eutectic cooled fast breeder reactors from liquid metal corrosion, Al 2O 3 nano- and micro-composite coatings were developed using an improved sol-gel process, which includes dipping specimens in a sol-gel solution dispersed with fine α-Al 2O 3 powders prepared by mechanical milling. Accelerated corrosion tests were conducted on coated specimens in liquid lead-bismuth eutectic at 500 °C under dynamic conditions. Scanning electron microscopy (SEM) and X-ray diffraction (XRD) analyses revealed that the coatings are composed of α-Al 2O 3 and they are about 10 μm thick. After the corrosion tests, no spallation occurred on the coatings, and neither Pb nor Bi penetrated into the coatings, which indicates that the coatings possess an enhanced dynamic LBE corrosion resistance to lead-bismuth eutectic corrosion. The nano-structured composite particles integrated into the coatings play an important role in achieving such superior lead-bismuth eutectic corrosion resistance.

  6. Scale modeling flow-induced vibrations of reactor components

    SciTech Connect

    Mulcahy, T M

    1982-06-01

    Similitude relationships currently employed in the design of flow-induced vibration scale-model tests of nuclear reactor components are reviewed. Emphasis is given to understanding the origins of the similitude parameters as a basis for discussion of the inevitable distortions which occur in design verification testing of entire reactor systems and in feature testing of individual component designs for the existence of detrimental flow-induced vibration mechanisms. Distortions of similitude parameters made in current test practice are enumerated and selected example tests are described. Also, limitations in the use of specific distortions in model designs are evaluated based on the current understanding of flow-induced vibration mechanisms and structural response.

  7. Repair welding of fusion reactor components. Final technical report

    SciTech Connect

    Chin, B.A.; Wang, C.A.

    1997-09-30

    The exposure of metallic materials, such as structural components of the first wall and blanket of a fusion reactor, to neutron irradiation will induce changes in both the material composition and microstructure. Along with these changes can come a corresponding deterioration in mechanical properties resulting in premature failure. It is, therefore, essential to expect that the repair and replacement of the degraded components will be necessary. Such repairs may require the joining of irradiated materials through the use of fusion welding processes. The present ITER (International Thermonuclear Experimental Reactor) conceptual design is anticipated to have about 5 km of longitudinal welds and ten thousand pipe butt welds in the blanket structure. A recent study by Buende et al. predict that a failure is most likely to occur in a weld. The study is based on data from other large structures, particularly nuclear reactors. The data used also appear to be consistent with the operating experience of the Fast Flux Test Facility (FFTF). This reactor has a fuel pin area comparable with the area of the ITER first wall and has experienced one unanticipated fuel pin failure after two years of operation. The repair of irradiated structures using fusion welding will be difficult due to the entrapped helium. Due to its extremely low solubility in metals, helium will diffuse and agglomerate to form helium bubbles after being trapped at point defects, dislocations, and grain boundaries. Welding of neutron-irradiated type 304 stainless steels has been reported with varying degree of heat-affected zone cracking (HAZ). The objectives of this study were to determine the threshold helium concentrations required to cause HAZ cracking and to investigate techniques that might be used to eliminate the HAZ cracking in welding of helium-containing materials.

  8. Structural materials and components

    NASA Technical Reports Server (NTRS)

    Gagliani, John (Inventor); Lee, Raymond (Inventor)

    1982-01-01

    High density structural (blocking) materials composed of a polyimide filled with glass microballoons. Structural components such as panels which have integral edgings and/or other parts made of the high density materials.

  9. Structural materials and components

    NASA Technical Reports Server (NTRS)

    Gagliani, John (Inventor); Lee, Raymond (Inventor)

    1983-01-01

    High density structural (blocking) materials composed of a polyimide filled with glass microballoons. Structural components such as panels which have integral edgings and/or other parts made of the high density materials.

  10. Structural materials and components

    NASA Technical Reports Server (NTRS)

    Gagliani, John (Inventor); Lee, Raymond (Inventor)

    1982-01-01

    High density structural (blocking) materials composed of a polyimide filled with glass microballoons and methods for making such materials. Structural components such as panels which have integral edgings and/or other parts made of the high density materials.

  11. Structural materials for breeder reactor cores and coolant circuits

    SciTech Connect

    Diercks, D.R.

    1984-02-01

    The structural components of principal interest in LMFBR cores and cooling circuits include the reactor vessel, primary and secondary piping, intermediate heat exchanger (IHX), and steam generator. Load-bearing components inside the vessel, among these the fuel cladding and duct, are also included. The operating conditions present in a fast-breeder nuclear reactor impose a number of requirements on the mechanical, physical, and neutronic properties of the materials used to construct these components.

  12. Generalized Structured Component Analysis

    ERIC Educational Resources Information Center

    Hwang, Heungsun; Takane, Yoshio

    2004-01-01

    We propose an alternative method to partial least squares for path analysis with components, called generalized structured component analysis. The proposed method replaces factors by exact linear combinations of observed variables. It employs a well-defined least squares criterion to estimate model parameters. As a result, the proposed method…

  13. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    SciTech Connect

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950`s are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  14. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    SciTech Connect

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  15. 78 FR 28896 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-16

    ... COMMISSION Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components... Combinations for Metal Primary Reactor Containment System Components,'' in which there are no substantive... loading combinations for metal primary reactor containment system components. ADDRESSES: Please refer...

  16. Austenitic alloy and reactor components made thereof

    DOEpatents

    Bates, John F.; Brager, Howard R.; Korenko, Michael K.

    1986-01-01

    An austenitic stainless steel alloy is disclosed, having excellent fast neutron irradiation swelling resistance and good post irradiation ductility, making it especially useful for liquid metal fast breeder reactor applications. The alloy contains: about 0.04 to 0.09 wt. % carbon; about 1.5 to 2.5 wt. % manganese; about 0.5 to 1.6 wt. % silicon; about 0.030 to 0.08 wt. % phosphorus; about 13.3 to 16.5 wt. % chromium; about 13.7 to 16.0 wt. % nickel; about 1.0 to 3.0 wt. % molybdenum; and about 0.10 to 0.35 wt. % titanium.

  17. Designed porosity materials in nuclear reactor components

    DOEpatents

    Yacout, A. M.; Pellin, Michael J.; Stan, Marius

    2016-09-06

    A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

  18. CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR

    SciTech Connect

    Vinson, Dennis

    2010-06-01

    The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

  19. Nondestructive Measurements for Diagnostics of Advanced Reactor Passive Components

    SciTech Connect

    Prowant, Matthew S.; Dib, Gerges; Roy, Surajit; Luzi, Lorenzo; Ramuhalli, Pradeep

    2016-09-20

    Information on advanced reactor (AdvRx) component condition and failure probability is necessary to maintaining adequate safety margins and avoiding unplanned shutdowns, both of which have regulatory and economic consequences. Prognostic health management (PHM) technologies provide one approach to addressing these needs by providing the technical means for lifetime management of significant passive components and reactor internals. However, such systems require measurement data that are sensitive to degradation of the component. This paper describes results to date of ongoing research on nondestructive measurements of component condition for degradation mechanisms of relevance to AdvRx concepts. The focus of this paper is on in-situ ultrasonic measurements during high-temperature creep degradation. The data were analyzed to assess the sensitivity of the measurements to creep degradation, with the specific objective of assessing the suitability of the resulting correlations for remaining life prediction. The details of the measurements, results of data analysis, and ongoing research in this area are discussed.

  20. Hydraulic balancing of a control component within a nuclear reactor

    DOEpatents

    Marinos, D.; Ripfel, H.C.F.

    1975-10-14

    A reactor control component includes an inner conduit, for instance containing neutron absorber elements, adapted for longitudinal movement within an outer guide duct. A transverse partition partially encloses one end of the conduit and meets a transverse wall within the guide duct when the conduit is fully inserted into the reactor core. A tube piece extends from the transverse partition and is coaxially aligned to be received within a tubular receptacle which extends from the transverse wall. The tube piece and receptacle cooperate in engagement to restrict the flow and pressure of coolant beneath the transverse partition and thereby minimize upward forces tending to expel the inner conduit.

  1. Emersion Testing of Phenix Reactor Components From Liquid Sodium

    SciTech Connect

    Baque, F.

    2002-07-01

    The life extension of the Phenix LMFR involved the inspection of reactor vessel internal structures: among other techniques, a visual inspection was performed of the above core structure, fuel assembly heads and upper components. To make this inspection possible, a partial draining of the main vessel from primary liquid sodium was carried out (sodium at 180 and argon cover at 150 ). The test program aimed at obtaining further knowledge on the process of wetting of sodium - as pure metal - on Phenix Plant assembly heads - made of stainless steel -, as well as on the internal structure welding, was carried out from November 1998 to January 1999. The main results were as follows: - the sodium meniscus measured during sodium lowering against the non-wet vertical structures reaches 10 mm in height. On wetted structures, it reaches only 5.3 mm. - when sodium level decreases, the process if very regular. However, re-flooding is carried out in stages. - a difference of 0.2 mm between two heads altitudes is enough to observe successively each of the heads. - the quality of sodium does not modify the wetting process (in the range of cold trap temperature: 110-140 deg. C). - the influence of lighting is important. - the visibility limit of emerging electro-eroded cracks (from 0.17 to 1.0 mm) is at 0.20 mm. - the visibility of a horizontal welding, machined or not, is good when the lighting is sufficient. - the superficial flow of sodium only modifies the wetting process for the closest heads. A final test allowed to observe that the global inclination of the assembly head mock-up does not modify the wetting process. These experimental results were part of the feasibility demonstration of the visual inspection within the actual Phenix Plant that was undertaken in 2001. (authors)

  2. Component and system simulation models for High Flux Isotope Reactor

    SciTech Connect

    Sozer, A.

    1989-08-01

    Component models for the High Flux Isotope Reactor (HFIR) have been developed. The models are HFIR core, heat exchangers, pressurizer pumps, circulation pumps, letdown valves, primary head tank, generic transport delay (pipes), system pressure, loop pressure-flow balance, and decay heat. The models were written in FORTRAN and can be run on different computers, including IBM PCs, as they do not use any specific simulation languages such as ACSL or CSMP. 14 refs., 13 figs.

  3. Carbon structural materials for fusion reactors

    SciTech Connect

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

    1993-12-31

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

  4. Nuclear reactor spacer grid and ductless core component

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1989-01-01

    The invention relates to a nuclear reactor spacer grid member for use in a liquid cooled nuclear reactor and to a ductless core component employing a plurality of these spacer grid members. The spacer grid member is of the egg-shell type and is constructed so that the walls of the cell members of the grid member are formed of a single thickness of metal to avoid tolerance problems. Within each cell member is a hydraulic spring which laterally constrains the nuclear material bearing rod which passes through each cell member against a hardstop in response to coolant flow through the cell member. This hydraulic spring is also suitable for use in a water cooled nuclear reactor. A core component constructed of, among other components, a plurality of these spacer grid members, avoids the use of a full length duct by providing spacer sleeves about the sodium tubes passing through the spacer grid members at locations between the grid members, thereby maintaining a predetermined space between adjacent grid members.

  5. Prognostics Health Management for Advanced Small Modular Reactor Passive Components

    SciTech Connect

    Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Mitchell, Mark R.; Wootan, David W.; Hirt, Evelyn H.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-10-18

    In the United States, sustainable nuclear power to promote energy security is a key national energy priority. Advanced small modular reactors (AdvSMR), which are based on modularization of advanced reactor concepts using non-light-water reactor (LWR) coolants such as liquid metal, helium, or liquid salt may provide a longer-term alternative to more conventional LWR-based concepts. The economics of AdvSMRs will be impacted by the reduced economy-of-scale savings when compared to traditional LWRs and the controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance costs. Therefore, achieving the full benefits of AdvSMR deployment requires a new paradigm for plant design and management. In this context, prognostic health management of passive components in AdvSMRs can play a key role in enabling the economic deployment of AdvSMRs. In this paper, the background of AdvSMRs is discussed from which requirements for PHM systems are derived. The particle filter technique is proposed as a prognostics framework for AdvSMR passive components and the suitability of the particle filter technique is illustrated by using it to forecast thermal creep degradation using a physics-of-failure model and based on a combination of types of measurements conceived for passive AdvSMR components.

  6. Passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  7. Natural circulating passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  8. Structural materials challenges for advanced reactor systems

    NASA Astrophysics Data System (ADS)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

  9. Method of detecting leakage of reactor core components of liquid metal cooled fast reactors

    DOEpatents

    Holt, Fred E.; Cash, Robert J.; Schenter, Robert E.

    1977-01-01

    A method of detecting the failure of a sealed non-fueled core component of a liquid-metal cooled fast reactor having an inert cover gas. A gas mixture is incorporated in the component which includes Xenon-124; under neutron irradiation, Xenon-124 is converted to radioactive Xenon-125. The cover gas is scanned by a radiation detector. The occurrence of 188 Kev gamma radiation and/or other identifying gamma radiation-energy level indicates the presence of Xenon-125 and therefore leakage of a component. Similarly, Xe-126, which transmutes to Xe-127 and Kr-84, which produces Kr-85.sup.m can be used for detection of leakage. Different components are charged with mixtures including different ratios of isotopes other than Xenon-124. On detection of the identifying radiation, the cover gas is subjected to mass spectroscopic analysis to locate the leaking component.

  10. Nuclear reactor heat transport system component low friction support system

    DOEpatents

    Wade, Elman E.

    1980-01-01

    A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

  11. An Assessment of Remote Visual Methods to Detect Cracking in Reactor Components

    SciTech Connect

    Cumblidge, Stephen E.; Anderson, Michael T.; Doctor, Steven R.; Simonen, Fredric A.; Elliot, Anthony J.

    2008-01-01

    Recently, the U.S. nuclear industry has proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, “Inservice Inspection of Nuclear Power Plant Components,” with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and time to perform the examination than do volumetric examinations such as ultrasonic testing. The issues relative to the reliability of VT in determining the structural integrity of reactor components were examined. Some piping and pressure vessel components in a nuclear power station are examined using VT as they are either in high radiation fields or component geometry precludes the use of ultrasonic testing (UT) methodology. Remote VT with radiation-hardened video systems has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, core shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote VT use submersible closed-circuit video cameras to examine reactor components and welds. PNNL conducted a parametric study that examined the important variables influencing the effectiveness of a remote visual test. Tested variables included lighting techniques, camera resolution, camera movement, and magnification. PNNL also conducted a limited laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to detect cracks of various widths under ideal conditions. The results of these studies and their implications are presented in this paper.

  12. 77 FR 39521 - Application for a License To Export Nuclear Reactor Major Components and Equipment

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-03

    ... COMMISSION Application for a License To Export Nuclear Reactor Major Components and Equipment Pursuant to 10... Reactor internals, Components and For use in Braka nuclear power Company LLC reactor coolant equipment for... plant in Braka. 110060011 control equipment, auxiliary equipment and emergency cooling systems. Dated...

  13. MPP: A modular library of models of nuclear reactor components

    SciTech Connect

    Abdalla, M.A.; Guimaraes, L.; Ugolini, D. ); March-Leuba, C.; Nypaver, D.J. ); Ford, C.E. )

    1992-01-01

    This paper presents the Modular Power Plant (MPP) library and its application to simulate the Advanced Liquid Metal Reactor. The MPP library is being developed as part of the Advanced Controls Program of the Oak Ridge National Laboratory. The general purpose of the library is to provide a set of modular models of components needed to simulate nuclear power plants. To give the MPP models modularity characteristics, each model is developed as a stand-alone system. The MPP contains 28 models coded in either the Advanced Continuous Simulation Language (ACSL), or the Generalized Object-Oriented Simulation Environment (GOOSE). The MPP development is parallel to the GOOSE development, and we are currently translating the MPP components from ACSL to GOOSE.

  14. MPP: A modular library of models of nuclear reactor components

    SciTech Connect

    Abdalla, M.A.; Guimaraes, L.; Ugolini, D.; March-Leuba, C.; Nypaver, D.J.; Ford, C.E.

    1992-05-01

    This paper presents the Modular Power Plant (MPP) library and its application to simulate the Advanced Liquid Metal Reactor. The MPP library is being developed as part of the Advanced Controls Program of the Oak Ridge National Laboratory. The general purpose of the library is to provide a set of modular models of components needed to simulate nuclear power plants. To give the MPP models modularity characteristics, each model is developed as a stand-alone system. The MPP contains 28 models coded in either the Advanced Continuous Simulation Language (ACSL), or the Generalized Object-Oriented Simulation Environment (GOOSE). The MPP development is parallel to the GOOSE development, and we are currently translating the MPP components from ACSL to GOOSE.

  15. Regularized Generalized Structured Component Analysis

    ERIC Educational Resources Information Center

    Hwang, Heungsun

    2009-01-01

    Generalized structured component analysis (GSCA) has been proposed as a component-based approach to structural equation modeling. In practice, GSCA may suffer from multi-collinearity, i.e., high correlations among exogenous variables. GSCA has yet no remedy for this problem. Thus, a regularized extension of GSCA is proposed that integrates a ridge…

  16. Regularized Generalized Structured Component Analysis

    ERIC Educational Resources Information Center

    Hwang, Heungsun

    2009-01-01

    Generalized structured component analysis (GSCA) has been proposed as a component-based approach to structural equation modeling. In practice, GSCA may suffer from multi-collinearity, i.e., high correlations among exogenous variables. GSCA has yet no remedy for this problem. Thus, a regularized extension of GSCA is proposed that integrates a ridge…

  17. Reactor Materials Program -- weldment component toughness of SRS PWS piping materials. [Process Water System

    SciTech Connect

    Sindelar, R.L.

    1993-02-01

    The mechanical properties of austenitic stainless steel materials from the reactor systems in the unirradiated (baseline) and the irradiated conditions have been developed previously for structural and fracture analyses of the pressure boundary of the SRS reactor Process Water System (PWS) components. Individual mechanical specimen test results were compiled into three separate weldment components or regions, namely, the base, weld, and weld heat-affected-zone (HAZ), for two orientations (L-C and C-L) with respect to the pipe axis of the source materials and for two test temperatures of 25 and 125[degrees]C. Twelve separate categories were thus defined to assess the effect of test conditions on the mechanical properties and to facilitate selection of properties for structural and fracture analyses. The testing results show high fracture toughness of the materials and support the demonstration of PWS pressure boundary structural integrity under all conditions of reactor operation. The fracture toughness of a fourth weldment component, namely, the weld fusion line region, has been measured to evaluate the potential for a region of low toughness in the interface between the Type 308 stainless steel weld metal and the Type 304 stainless steel pipe. The testing details and results of the weld fusion line toughness are contained in this report.

  18. Work Breakdown Structure and Plant/Equipment Designation System Numbering Scheme for the High Temperature Gas- Cooled Reactor (HTGR) Component Test Capability (CTC)

    SciTech Connect

    Jeffrey D Bryan

    2009-09-01

    This white paper investigates the potential integration of the CTC work breakdown structure numbering scheme with a plant/equipment numbering system (PNS), or alternatively referred to in industry as a reference designation system (RDS). Ideally, the goal of such integration would be a single, common referencing system for the life cycle of the CTC that supports all the various processes (e.g., information, execution, and control) that necessitate plant and equipment numbers be assigned. This white paper focuses on discovering the full scope of Idaho National Laboratory (INL) processes to which this goal might be applied as well as the factors likely to affect decisions about implementation. Later, a procedure for assigning these numbers will be developed using this white paper as a starting point and that reflects the resolved scope and outcome of associated decisions.

  19. Preloading of bolted connections in nuclear reactor component supports

    SciTech Connect

    Yahr, G T

    1984-10-01

    A number of failures of threaded fasteners in nuclear reactor component supports have been reported. Many of those failures were attributed to stress corrosion cracking. This report discusses how stress corrosion cracking can be avoided in bolting by controlling the maximum bolt preloads so that the sustained stresses in the bolts are below the level required to cause stress corrosion cracking. This is a basic departure from ordinary bolted joint design where the only limits on preload are on the minimum preload. Emphasis is placed on the importance of detailed analysis to determine the acceptable range of preload and the selection of a method for measuring the preload that is sufficiently accurate to ensure that the preload is actually within the acceptable range. Procedures for determining acceptable preload range are given, and the accuracy of various methods of measuring preload is discussed.

  20. NDE Assessments of Cast Stainless Steel Reactor Piping Components

    SciTech Connect

    Diaz, Aaron A.; Anderson, Michael T.; Cumblidge, Stephen E.; Doctor, Steven R.; Mathews, Royce

    2006-02-01

    Studies conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, have focused on developing and evaluating the effectiveness and reliability of novel NDE approaches for the inspection of coarse-grained, cast stainless steel reactor components. The primary objective of this work is to provide information to the United States Nuclear Regulatory Commission (US NRC) on the utility, effectiveness and reliability of ultrasonic testing (UT) and eddy current testing (ET) inspection techniques as related to the in-service inspection of primary piping components in pressurized water reactors (PWRs). This paper describes recent developments and results from assessments of three different NDE approaches including an ultrasonic phased array inspection methodology, an eddy current testing technique and a low-frequency ultrasonic inspection methodology coupled with a synthetic aperture focusing technique (SAFT). Westinghouse Owner’s Group (WOG) cast stainless steel pipe segments with thermal and mechanical fatigue cracks located close to the weld roots, were used for assessing the inspection methods. ET studies were conducted on the inner diameter (ID) surface of piping specimens while the ultrasonic inspection methods were performed from the outer diameter (OD) surface of the specimens. The ET technique employed a ZETEC MIZ-27SI Eddy Current instrument and a ZETEC Z0000857-1 cross point spot probe with an operating frequency of 250 kHz. On some samples where noise levels were high, degaussing of the sample resulted in significant improvements. The phased array approach was implemented using an RD Tech Tomoscan III system operating at 1 MHz and composite volumetric images of the samples were generated. The low-frequency ultrasonic method employs a zone-focused, multi-incident angle; inspection protocol (operating at 250-450 kHz) coupled with a synthetic aperture focusing technique (SAFT) for improved signal-to-noise and advanced imaging

  1. PIPING FOR COOLANT WATER IS INSTALLED INSIDE REACTOR STRUCTURE PRIOR ...

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

    PIPING FOR COOLANT WATER IS INSTALLED INSIDE REACTOR STRUCTURE PRIOR TO EMBEDMENT IN CONCRETE. HIGHER PIPE IS INLET; THE OTHER, THE OUTLET LOOP. INLET PIPE WILL CONNECT TO TOP SECTION OF REACTOR VESSEL. INL NEGATIVE NO. 1287. Unknown Photographer, 1/18/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  2. 78 FR 57904 - Request for a License To Export; Reactor Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-20

    ... COMMISSION Request for a License To Export; Reactor Components Pursuant to 10 CFR 110.70 (b) ``Public Notice... 28, 2013, August 29, 2013, systems, related reactors. operation of AP- XR177, 11006121. equipment, and 1000 (design) spare parts. nuclear reactors. Dated this 16th day of September 2013 in...

  3. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false General license for the export of nuclear reactor... NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...) of this section any nuclear reactor component of U.S. origin described in paragraphs (5) through...

  4. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false General license for the export of nuclear reactor... NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...) of this section any nuclear reactor component of U.S. origin described in paragraphs (5) through...

  5. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false General license for the export of nuclear reactor... NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...) of this section any nuclear reactor component of U.S. origin described in paragraphs (5) through...

  6. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false General license for the export of nuclear reactor... NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...) of this section any nuclear reactor component of U.S. origin described in paragraphs (5) through...

  7. Induced Radioactivity and Waste Classification of Reactor Zone Components of the Chernobyl Nuclear Power Plant Unit 1 After Final Shutdown

    SciTech Connect

    Bylkin, Boris K.; Davydova, Galina B.; Zverkov, Yuri A.; Krayushkin, Alexander V.; Neretin, Yuri A.; Nosovsky, Anatoly V.; Seyda, Valery A.; Short, Steven M.

    2001-10-15

    The dismantlement of the reactor core materials and surrounding structural components is a major technical concern for those planning closure and decontamination and decommissioning of the Chernobyl Nuclear Power Plant (NPP). Specific issues include when and how dismantlement should be accomplished and what the radwaste classification of the dismantled system would be at the time it is disassembled. Whereas radiation levels and residual radiological characteristics of the majority of the plant systems are directly measured using standard radiation survey and radiochemical analysis techniques, actual measurements of reactor zone materials are not practical due to high radiation levels and inaccessibility. For these reasons, neutron transport analysis was used to estimate induced radioactivity and radiation levels in the Chernobyl NPP Unit 1 reactor core materials and structures.Analysis results suggest that the optimum period of safe storage is 90 to 100 yr for the Unit 1 reactor. For all of the reactor components except the fuel channel pipes (or pressure tubes), this will provide sufficient decay time to allow unlimited worker access during dismantlement, minimize the need for expensive remote dismantlement, and allow for the dismantled reactor components to be classified as low- or medium-level radioactive waste. The fuel channel pipes will remain classified as high-activity waste requiring remote dismantlement for hundreds of years due to the high concentration of induced {sup 63}Ni in the Zircaloy pipes.

  8. GENERIC, COMPONENT FAILURE DATA BASE FOR LIGHT WATER AND LIQUID SODIUM REACTOR PRAs

    SciTech Connect

    S. A. Eide; S. V. Chmielewski; T. D. Swantz

    1990-02-01

    A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs) . The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates . Using this approach, most of the failure rates are based on actual plant data rather than existing estimates .

  9. Eddy current position indicating apparatus for measuring displacements of core components of a liquid metal nuclear reactor

    DOEpatents

    Day, Clifford K.; Stringer, James L.

    1977-01-01

    Apparatus for measuring displacements of core components of a liquid metal fast breeder reactor by means of an eddy current probe. The active portion of the probe is located within a dry thimble which is supported on a stationary portion of the reactor core support structure. Split rings of metal, having a resistivity significantly different than sodium, are fixedly mounted on the core component to be monitored. The split rings are slidably positioned around, concentric with the probe and symmetrically situated along the axis of the probe so that motion of the ring along the axis of the probe produces a proportional change in the probes electrical output.

  10. Inelastic behavior of structural components

    NASA Technical Reports Server (NTRS)

    Hussain, N.; Khozeimeh, K.; Toridis, T. G.

    1980-01-01

    A more accurate procedure was developed for the determination of the inelastic behavior of structural components. The actual stress-strain curve for the mathematical of the structure was utilized to generate the force-deformation relationships for the structural elements, rather than using simplified models such as elastic-plastic, bilinear and trilinear approximations. relationships were generated for beam elements with various types of cross sections. In the generational of these curves, stress or load reversals, kinematic hardening and hysteretic behavior were taken into account. Intersections between loading and unloading branches were determined through an iterative process. Using the inelastic properties obtained, the plastic static response of some simple structural systems composed of beam elements was computed. Results were compared with known solutions, indicating a considerable improvement over response predictions obtained by means of simplified approximations used in previous investigations.

  11. Structural Studies of Ciliary Components

    PubMed Central

    Mizuno, Naoko; Taschner, Michael; Engel, Benjamin D.; Lorentzen, Esben

    2012-01-01

    Cilia are organelles found on most eukaryotic cells, where they serve important functions in motility, sensory reception, and signaling. Recent advances in electron tomography have facilitated a number of ultrastructural studies of ciliary components that have significantly improved our knowledge of cilium architecture. These studies have produced nanometer‐resolution structures of axonemal dynein complexes, microtubule doublets and triplets, basal bodies, radial spokes, and nexin complexes. In addition to these electron tomography studies, several recently published crystal structures provide insights into the architecture and mechanism of dynein as well as the centriolar protein SAS-6, important for establishing the 9-fold symmetry of centrioles. Ciliary assembly requires intraflagellar transport (IFT), a process that moves macromolecules between the tip of the cilium and the cell body. IFT relies on a large 20-subunit protein complex that is thought to mediate the contacts between ciliary motor and cargo proteins. Structural investigations of IFT complexes are starting to emerge, including the first three‐dimensional models of IFT material in situ, revealing how IFT particles organize into larger train-like arrays, and the high-resolution structure of the IFT25/27 subcomplex. In this review, we cover recent advances in the structural and mechanistic understanding of ciliary components and IFT complexes. PMID:22683354

  12. Performance and lifetime assessment of reactor wall and nearby components during plasma instabilities.

    SciTech Connect

    Hassanein, A.

    1998-03-10

    Surface and structural damage to plasma-facing components due to the frequent loss of plasma confinement is a serious problem for the tokamak reactor concept. The plasma energy deposited on these components during loss of confinement causes significant surface erosion, possible structural failure, and frequent plasma contamination. Surface damage consists of vaporization, spallation, and liquid splatter of metallic materials. Comprehensive multidimensional models that include thermodynamics and thermal hydraulics of plasma-facing materials, eroded-debris/vapor atomic physics and magnetohydrodynamics, resulting photon radiation and photon transport, as well as liquid splashing and brittle destruction of materials, are used self-consistently to evaluate and assess our current understanding of the lifetime of plasma-facing materials and the various forms of damage they experience. Models are developed to study the stability of the vapor shielding layer, erosion of the melt-layer, brittle destruction/explosive erosion, and the issues involved therein.

  13. Development of Underwater Laser Cladding and Underwater Laser Seal Welding Techniques for Reactor Components (II)

    SciTech Connect

    Masataka Tamura; Shohei Kawano; Wataru Kouno; Yasushi Kanazawa

    2006-07-01

    Stress corrosion cracking (SCC) is one of the major reasons to reduce the reliability of aged reactor components. Toshiba has been developing underwater laser welding onto surface of the aged components as maintenance and repair techniques. Because most of the reactor internal components to apply this underwater laser welding technique have 3-dimensional shape, effect of welding positions and welded shapes are examined and presented in this report. (authors)

  14. Code qualification of structural materials for AFCI advanced recycling reactors.

    SciTech Connect

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L.

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the

  15. 76 FR 16842 - Request for a License To Export Reactor Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-25

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Request for a License To Export Reactor Components Pursuant to 10 CFR 110.70 (b) ``Public Notice... systems, related reactors. operation of AP- equipment, and 1000 (design) spare parts. nuclear...

  16. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    SciTech Connect

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  17. Structural integrity of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Knott, John F.

    2013-09-01

    The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify 'Incredibility of Failure' and how these are incorporated in assessments of the integrity of existing plant and proposed 'new build' PWR pressure vessels.

  18. THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS

    SciTech Connect

    David S. Duncan; Vondell J. Balls; Stephanie L. Austad

    2008-09-01

    The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

  19. Structural Components of Oriboca Virus

    PubMed Central

    Rosato, Robert R.; Robbins, Mary Louise; Eddy, Gerald A.

    1974-01-01

    Analysis of purified Oriboca virions by neutral, sodium dodecyl sulfate polyacrylamide-gel electrophoresis indicated the presence of three structural polypeptides designated V-1, V-2, and V-3 on the basis of their relative electrophoretic mobilities in 8% gels. Polypeptides V-2 and V-3 are glycopeptides associated with the virion envelope as demonstrated by the preferential incorporation of labeled glucosamine into the polypeptides and by release of the polypeptides from the intact virion by the nonionic detergent NP-40. Polypeptide V-1 is the protein component of the nucleoprotein core of Oriboca virus as evidenced by the specific incorporation of uridine into the nucleoprotein, its release from the intact virion by NP-40 treatment, and its separation by both rate-zonal and isopycnic density gradient centrifugation from both the intact virion and envelope components. Molecular weights have been tentatively assigned to the polypeptides by extrapolation from the structural polypeptides of Sindbis virus when both are run in the same gel. Polypeptide V-1 has an apparent molecular weight of 20,000 to 23,000; V-2, 30,000 to 32,000; and V-3, 83,000 to 85,000. PMID:4821489

  20. Structural clay tile component behavior

    SciTech Connect

    Columber, Christopher Eugene

    1994-12-18

    The basic properties of structural clay tile walls were determined through component and composite testing of structural clay tile and mortar. The fundamental material parameters and strengths of clay tile coupons were determined through compression, tension, modulus of rupture and absorption tests. Mortar cylinders were tested in both compression and split cylinder fashion. Stress-strain curves for mortar under compression were determined. Four miniature prisms were tested in compression. These prisms were made from two 8 inches x 12 inches x 12 inches structural clay tiles, using a stack bond with a 3/4 inches mortar joint. Stress strain curves as well as material property values were obtained. These results were compared with previous tests on larger (2 feet x 4 feet) prisms. Twenty five bond wrench samples were tested. Two series of bond wrench samples were run. The first series (six tests) were fitted with LVDTs so that load deflection curves as well as flexural strengths could be obtained. A shifting of the neutral axis towards the compression face was observed. The second series were made with different mortar types: type N masonry cement mortar, type S masonry cement mortar, type N portland cement lime (PCL) mortar, and type S PCL mortar. Type S mortar and portland cement lime mortar were found to improve the bond strength.

  1. Functional Generalized Structured Component Analysis.

    PubMed

    Suk, Hye Won; Hwang, Heungsun

    2016-12-01

    An extension of Generalized Structured Component Analysis (GSCA), called Functional GSCA, is proposed to analyze functional data that are considered to arise from an underlying smooth curve varying over time or other continua. GSCA has been geared for the analysis of multivariate data. Accordingly, it cannot deal with functional data that often involve different measurement occasions across participants and a large number of measurement occasions that exceed the number of participants. Functional GSCA addresses these issues by integrating GSCA with spline basis function expansions that represent infinite-dimensional curves onto a finite-dimensional space. For parameter estimation, functional GSCA minimizes a penalized least squares criterion by using an alternating penalized least squares estimation algorithm. The usefulness of functional GSCA is illustrated with gait data.

  2. Flow-induced vibration and instability of some nuclear-reactor-system components. [PWR

    SciTech Connect

    Chen, S.S.

    1983-01-01

    The high-velocity coolant flowing through a reactor system component is a source of energy that can induce component vibration and instability. In fact, many reactor components have suffered from excessive vibration and/or dynamic instability. The potential for detrimental flow-induced vibration makes it necessary that design engineers give detailed considerations to the flow-induced vibration problems. Flow-induced-vibration studies have been performed in many countries. Significant progress has been made in understanding the different phenomena and development of design guidelines to avoid damaging vibration. The purpose of this paper is to present an overview of the recent progress in several selected areas, to discuss some new results and to indentify future research needs. Specifically, the following areas will be presented: examples of flow-induced-vibration problems in reactor components; excitation mechanisms and component response characteristics; instability mechanisms and stability criteria; design considerations; and future research needs.

  3. Liquid metal systems development: reactor vessel support structure evaluation

    SciTech Connect

    McEdwards, J.A.

    1981-01-01

    Results of an evaluation of support structures for the reactor vessel are reported. The U ring, box ring, integral ring, tee ring and tangential beam supports were investigated. The U ring is the recommended vessel support structure configuration.

  4. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    SciTech Connect

    Holcomb, David Eugene; Cetiner, Sacit M; Flanagan, George F; Peretz, Fred J; Yoder Jr, Graydon L

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  5. Component-Level Prognostics Health Management Framework for Passive Components - Advanced Reactor Technology Milestone: M2AT-15PN2301043

    SciTech Connect

    Ramuhalli, Pradeep; Roy, Surajit; Hirt, Evelyn H.; Prowant, Matthew S.; Pitman, Stan G.; Tucker, Joseph C.; Dib, Gerges; Pardini, Allan F.

    2015-06-19

    This report describes research results to date in support of the integration and demonstration of diagnostics technologies for prototypical advanced reactor passive components (to establish condition indices for monitoring) with model-based prognostics methods. Achieving this objective will necessitate addressing several of the research gaps and technical needs described in previous technical reports in this series.

  6. Catalyst support structure, catalyst including the structure, reactor including a catalyst, and methods of forming same

    DOEpatents

    Van Norman, Staci A.; Aston, Victoria J.; Weimer, Alan W.

    2017-05-09

    Structures, catalysts, and reactors suitable for use for a variety of applications, including gas-to-liquid and coal-to-liquid processes and methods of forming the structures, catalysts, and reactors are disclosed. The catalyst material can be deposited onto an inner wall of a microtubular reactor and/or onto porous tungsten support structures using atomic layer deposition techniques.

  7. Space Fission Reactor Structural Materials: Choices Past, Present and Future

    SciTech Connect

    Busby, Jeremy T; Leonard, Keith J

    2007-01-01

    Nuclear powered spacecraft will enable missions well beyond the capabilities of current chemical, radioisotope thermal generator and solar technologies. The use of fission reactors for space applications has been considered for over 50 years, although, structural material performance has often limited the potential performance of space reactors. Space fission reactors are an extremely harsh environment for structural materials with high temperatures, high neutron fields, potential contact with liquid metals, and the need for up to 15-20 year reliability with no inspection or preventative maintenance. Many different materials have been proposed as structural materials. While all materials meet many of the requirements for space reactor service, none satisfy all of them. However, continued development and testing may resolve these issues and provide qualified materials for space fission reactors.

  8. Influence of Natural Convection and Thermal Radiation Multi-Component Transport in MOCVD Reactors

    NASA Technical Reports Server (NTRS)

    Lowry, S.; Krishnan, A.; Clark, I.

    1999-01-01

    The influence of Grashof and Reynolds number in Metal Organic Chemical Vapor (MOCVD) reactors is being investigated under a combined empirical/numerical study. As part of that research, the deposition of Indium Phosphide in an MOCVD reactor is modeled using the computational code CFD-ACE. The model includes the effects of convection, conduction, and radiation as well as multi-component diffusion and multi-step surface/gas phase chemistry. The results of the prediction are compared with experimental data for a commercial reactor and analyzed with respect to the model accuracy.

  9. Reactor Materials Program -- weldment component toughness of SRS PWS piping materials. Task number: 89-023-1

    SciTech Connect

    Sindelar, R.L.

    1993-02-01

    The mechanical properties of austenitic stainless steel materials from the reactor systems in the unirradiated (baseline) and the irradiated conditions have been developed previously for structural and fracture analyses of the pressure boundary of the SRS reactor Process Water System (PWS) components. Individual mechanical specimen test results were compiled into three separate weldment components or regions, namely, the base, weld, and weld heat-affected-zone (HAZ), for two orientations (L-C and C-L) with respect to the pipe axis of the source materials and for two test temperatures of 25 and 125{degrees}C. Twelve separate categories were thus defined to assess the effect of test conditions on the mechanical properties and to facilitate selection of properties for structural and fracture analyses. The testing results show high fracture toughness of the materials and support the demonstration of PWS pressure boundary structural integrity under all conditions of reactor operation. The fracture toughness of a fourth weldment component, namely, the weld fusion line region, has been measured to evaluate the potential for a region of low toughness in the interface between the Type 308 stainless steel weld metal and the Type 304 stainless steel pipe. The testing details and results of the weld fusion line toughness are contained in this report.

  10. The Capabilities and Limitation of Remote Visual Methods to Detect Service-Induced Cracks in Reactor Components

    SciTech Connect

    Cumblidge, Stephen E.; Doctor, Steven R.; Anderson, Michael T.

    2006-11-01

    Since 1977, the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research has funded a multiyear program at the Pacific Northwest National Laboratory (PNNL) to evaluate the reliability and accuracy of nondestructive evaluation (NDE) techniques employed for inservice inspection (ISI). Recently, the U.S. nuclear industry proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by ASME Boiler and Pressure Vessel Code Section XI, with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and examination times than do volumetric examinations such as ultrasonic testing (UT). However, for industry to justify supplamenting volumetric metods with VT, and analysis of pertinent issues is needed to support the reliability of VT in determining the structural intefrity of reactor components. As piping and pressure vessel compoents in a nuclear power station are generally underwater and in high radiation field, they need to be examined by VT from a distance with radiation-hardened video systems. Remote visual testing has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, for shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote visual testing use submersible closed-circuit video cameras to examine reactor components and welds. PNNL has conducted a parametric study that examines the important variables that affect the effectiveness of a remote visual test. Tested variables include lighting techniques, camera resolution, camera movement, and magnification. PNNL has also conductrd a laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to

  11. Research in nondestructive evaluation techniques for nuclear reactor concrete structures

    NASA Astrophysics Data System (ADS)

    Clayton, Dwight; Smith, Cyrus

    2014-02-01

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

  12. Corrosion of structural materials by lead-based reactor coolants.

    SciTech Connect

    Abraham, D. P.; Leibowitz, L.; Maroni, V. A.; McDeavitt, S. M.; Raraz, A. G.

    2000-11-16

    Advanced nuclear reactor design has, in recent years, focused increasingly on the use of heavy-liquid-metal coolants, such as lead and lead-bismuth eutectic. Similarly, programs on accelerator-based transmutation systems have also considered the use of such coolants. Russian experience with heavy-metal coolants for nuclear reactors has lent credence to the validity of this approach. Of significant concern is the compatibility of structural materials with these coolants. We have used a thermal convection-based test method to allow exposure of candidate materials to molten lead and lead-bismuth flowing under a temperature gradient. The gradient was deemed essential in evaluating the behavior of the test materials in that should preferential dissolution of components of the test material occur we would expect dissolution in the hotter regions and deposition in the colder regions, thus promoting material transport. Results from the interactions of a Si-rich mild steel alloy, AISI S5, and a ferritic-martensitic stainless steel, HT-9, with the molten lead-bismuth are presented.

  13. Research in nondestructive evaluation techniques for nuclear reactor concrete structures

    SciTech Connect

    Clayton, Dwight; Smith, Cyrus

    2014-02-18

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

  14. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    SciTech Connect

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers.

  15. Disposal of Large Reactor Components - Rulemaking to Address Funding of Disposal Costs

    SciTech Connect

    Greeves, J.T.; Lieberman, J.; Magette, T.E.

    2008-07-01

    The paper will explore the current challenges of financing the disposal costs for disposing of large reactor components such as reactor pressure vessel heads and steam generators and the resulting delays in disposal caused by the current regulatory requirements. The paper also will discuss a recent rulemaking petition submitted by EnergySolutions to the US Nuclear Regulatory Commission designed to improve the regulatory process by providing a process to permit funds from decommissioning trust funds to be used to fund disposal of large reactor components. If granted, the disposal of these large components could be expedited where reactor licensees have sufficient decommissioning trust funds available. Perspectives on the rulemaking will be addressed. In conclusion: NRC should provide serious consideration to this Petition. There is support in the industry for granting this Petition. As of the date this paper was submitted, there have been no negative comments. Granting this Petition is prudent and consistent with the underlying purpose of 10 C.F.R. 50.82(a)(8) and 10 CFR 20.1406. It provides flexibility without any adverse impact on the public health and safety. It should facilitate the decommissioning process by providing a regulatory framework to allow removing MRCs from sites, resulting in (1) the source term at the site being reduced, (2) the site workers being exposed to less radiation, (3) eliminating an unnecessary regulatory burden as the costs associated with maintaining the MRCs on-site and providing protection to the workers as a result of those components can be avoided, (4) the overall cost to decommission the site being reduced, and (5) more funds being made available to decommission the reactor at the time the reactor ceases operation. Finally, the framework would provide the demonstration by a site-specific decommissioning cost estimate and the associated funding program that adequate funds are available to dispose of these components as well as

  16. Development of plasma facing components for fusion experimental reactors

    SciTech Connect

    Onozuka, M.; Fujiya, Y.; Inoue, M.; Morimoto, M.

    1995-12-31

    The divertor structure and fabrication process have been investigated, including the structures of the divertor elements and support, fundamental brazing techniques, brazing of large divertor tiles and fabrication method of large divertor modules. Using direct brazing, a partial divertor module with large CFC tiles was fabricated and tested. It was shown that the model had sufficient structural integrity against thermal shocks of {approximately}17MW/m{sup 2} {times} 4 sec for up to 1,600 times. A fabrication technique for large and complex-shaped divertor module has been developed and successfully applied to a 1m-long linear and 0.8m-long curved divertor modules. In addition, preliminary investigation of direct brazing of beryllium to the copper substrate has been conducted. It was found that the bending strength of the bonded materials was around 40 MPa. Furthermore, boron coating on the CFC and Mo has been examined. Using the boron ion implantation technique, boron ions were implanted to the CFC and Mo plates prior to the boron atoms deposition. The samples fabricated with this method were found to have a sufficient thermal shock resistance.

  17. Thermal aging of some decommissioned reactor components and methodology for life prediction

    SciTech Connect

    Chung, H.M.

    1989-03-01

    Since a realistic aging of cast stainless steel components for end-of-life or life-extension conditions cannot be produced, it is customary to simulate the thermal aging embrittlement by accelerated aging at /approximately/400/degree/C. In this investigation, field components obtained from decommissioned reactors have been examined after service up to 22 yr to provide a benchmark of the laboratory simulation. The primary and secondary aging processes were found to be identical to those of the laboratory-aged specimens, and the kinetic characteristics were also similar. The extent of the aging embrittlement processes and other key factors that are known to influence the embrittlement kinetics have been compared for the decommissioned reactor components and materials aged under accelerated conditions. On the basis of the study, a mechanistic understanding of the causes of the complex behavior in kinetics and activation energy of aging (i.e., the temperature dependence of aging embrittlement between the accelerated and reactor-operating conditions) is presented. A mechanistic correlation developed thereon is compared with a number of available empirical correlations to provide an insight for development of a better methodology of life prediction of the reactor components. 18 refs., 18 figs., 5 tabs.

  18. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    SciTech Connect

    Owen, M.B.

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR.

  19. High heat flux issues for plasma-facing components in fusion reactors

    NASA Astrophysics Data System (ADS)

    Watson, Robert D.

    1993-02-01

    Plasma facing components in tokamak fusion reactors are faced with a number of difficult high heat flux issues. These components include: first wall armor tiles, pumped limiters, diverter plates, rf antennae structure, and diagnostic probes. Peak heat fluxes are 15 - 30 MW/m2 for diverter plates, which will operate for 100 - 1000 seconds in future tokamaks. Disruption heat fluxes can approach 100,000 MW/m2 for 0.1 ms. Diverter plates are water-cooled heat sinks with armor tiles brazed on to the plasma facing side. Heat sink materials include OFHC, GlidcopTM, TZM, Mo-41Re, and niobium alloys. Armor tile materials include: carbon fiber composites, beryllium, silicon carbide, tungsten, and molybdenum. Tile thickness range from 2 - 10 mm, and heat sinks are 1 - 3 mm. A twisted tape insert is used to enhance heat transfer and increase the burnout safety margin from critical heat flux limits to 50 - 60 MW/m2 with water at 10 m/s and 4 MPa. Tests using rastered electron beams have shown thermal fatigue failures from cracks at the brazed interface between tiles and the heat sink after only 1000 cycles at 10 - 15 MW/m2. These fatigue lifetimes need to be increased an order of magnitude to meet future requirements. Other critical issues for plasma facing components include: surface erosion from sputtering and disruption erosion, eddy current forces and runaway electron impact from disruptions, neutron damage, tritium retention and release, remote maintenance of radioactive components, corrosion-erosion, and loss-of-coolant accidents.

  20. Structured Functional Principal Component Analysis

    PubMed Central

    Shou, Haochang; Zipunnikov, Vadim; Crainiceanu, Ciprian M.; Greven, Sonja

    2015-01-01

    Summary Motivated by modern observational studies, we introduce a class of functional models that expand nested and crossed designs. These models account for the natural inheritance of the correlation structures from sampling designs in studies where the fundamental unit is a function or image. Inference is based on functional quadratics and their relationship with the underlying covariance structure of the latent processes. A computationally fast and scalable estimation procedure is developed for high-dimensional data. Methods are used in applications including high-frequency accelerometer data for daily activity, pitch linguistic data for phonetic analysis, and EEG data for studying electrical brain activity during sleep. PMID:25327216

  1. SHM in complex structural components

    NASA Astrophysics Data System (ADS)

    Croxford, Anthony J.; Wilcox, Paul D.; Courtney, Charles R. P.; Drinkwater, Bruce W.

    2009-03-01

    The use of permanently attached arrays of sensors has made it clear that guided waves can be used for the SHM of structures. The approaches developed have relied on the use of reference signal subtraction to indicate changes to the state of the structure, such as the appearance of damage. The limit of performance of any system is defined by the post subtraction noise. In order to confirm the basic principles at work the majority of this work has been carried out on simple metallic plates. While important to confirm the levels of understanding, this is not sufficient for practical use. This paper looks at the application of SHM techniques in more complex structures, more typical of those any system would be used on in practise. A rib from a BaE 146 aircraft is used to demonstrate the practical difficulties of applying guided wave SHM methods to densely featured structures. A model system comprising a plate with a single stringer is used to demonstrate a method for normalizing signals to give responses directly related to the scattering properties of the change in the system, mitigating the effect of the position of the change, and a method is proposed to generalize the approach to complex systems. Preliminary tests in the region of the stringer are used to identify the experimental challenges to realizing the calibration on complex systems.

  2. Compression Strength of Composite Primary Structural Components

    NASA Technical Reports Server (NTRS)

    Johnson, Eric R.; Starnes, James H., Jr. (Technical Monitor)

    2000-01-01

    The focus of research activities under NASA Grant NAG-1-2035 was the response and failure of thin-walled structural components. The research is applicable to the primary load carrying structure of flight vehicles, with particular emphasis on fuselage and wing'structure. Analyses and tests were performed that are applicable to the following structural components an aft pressure bulkhead, or a composite pressure dome, pressure cabin damage containment, and fuselage frames subject to crash-type loads.

  3. Generalized Structured Component Analysis with Latent Interactions

    ERIC Educational Resources Information Center

    Hwang, Heungsun; Ho, Moon-Ho Ringo; Lee, Jonathan

    2010-01-01

    Generalized structured component analysis (GSCA) is a component-based approach to structural equation modeling. In practice, researchers may often be interested in examining the interaction effects of latent variables. However, GSCA has been geared only for the specification and testing of the main effects of variables. Thus, an extension of GSCA…

  4. Generalized Structured Component Analysis with Latent Interactions

    ERIC Educational Resources Information Center

    Hwang, Heungsun; Ho, Moon-Ho Ringo; Lee, Jonathan

    2010-01-01

    Generalized structured component analysis (GSCA) is a component-based approach to structural equation modeling. In practice, researchers may often be interested in examining the interaction effects of latent variables. However, GSCA has been geared only for the specification and testing of the main effects of variables. Thus, an extension of GSCA…

  5. U. S. fast reactor materials and structures program

    SciTech Connect

    Harms, W.O.; Purdy, C.M.

    1984-01-01

    The U.S. DOE has sponsored a vigorous breeder reactor materials and structures program for 15 years. Important contributions have resulted from this effort in the areas of design (inelastic rules, verified methods, seismic criteria, mechanical properties data); resolution of licensing issues (technical witnessing, confirmatory testing); construction (fabrication/welding procedures, nondestructive testing techniques); and operation (sodium purification, instrumentation and chemical analysis, radioactivity control, and in-service inspection. The national LMFBR program currently is being restructured. The Materials and Structures Program will focus its efforts in the following areas: (1) removal of anticipated licensing impediments through confirmation of the adequacy of structural design methods and criteria for components containing welds and geometric discontinuities, the generation of mechanical properties for stainless steel castings and weldments, and the evaluation of irradiation effects; (2) qualification of modified 9 Cr-1 Mo steel and tribological coatings for design flexibility; (3) development of improved inelastic design guidelines and procedures; (4) reform of design codes and standards and engineering practices, leading to simpler, less conservative rules and to simplified design analysis methods; and (5) incorporation of information from foreign program.

  6. Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

    NASA Astrophysics Data System (ADS)

    Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

    2017-02-01

    Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

  7. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    SciTech Connect

    Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  8. Potential Application of Electrical Signature Analysis Methods for Monitoring Small Modular Reactor Components

    SciTech Connect

    Damiano, Brian; Tucker Jr, Raymond W; Haynes, Howard D

    2010-01-01

    This paper will describe the technical basis behind ESA and why we consider it a viable SMR condition monitoring technology. Concepts are presented of how ESA could be applied to monitor two candidate small modular reactor components: the main coolant pumps and the control rod drives. We believe the general health of these two components can be monitored and trended over time, using ESA methods. Our optimism is based on over two decades of ESA development and testing on a wide variety of components and systems, many of which have similar operational features to the main coolant pumps and control rod drives.

  9. Requirements for Prognostic Health Management of Passive Components in Advanced Small Modular Reactors

    SciTech Connect

    Meyer, Ryan M.; Coble, Jamie B.; Ramuhalli, Pradeep

    2013-08-01

    Advanced small modular reactors (aSMRs), which are based on modularization of advanced reactor concepts, may provide a longer-term alternative to traditional light-water reactors and near term small modular reactors (SMRs), which are based on integral pressurized water reactor (iPWR) concepts. aSMRs are conceived for applications in remote locations and for diverse missions that include providing process or district heating, water desalination, and hydrogen production. Several challenges exist with respect to cost-effective operations and maintenance (O&M) of aSMRs, including the impacts of aggressive operating environments and modularity, and limiting these costs and staffing needs will be essential to ensuring the economic feasibility of aSMR deployment. In this regard, prognostic health management (PHM) systems have the potential to play a vital role in supporting the deployment of aSMR systems. This paper identifies requirements and technical gaps associated with implementation of PHM systems for passive aSMR components.

  10. Embrittlement and Flow Localization in Reactor Structural Materials

    SciTech Connect

    Xianglin Wu; Xiao Pan; James Stubbins

    2006-10-06

    Many reactor components and structural members are made from metal alloys due, in large part, to their strength and ability to resist brittle fracture by plastic deformation. However, brittle fracture can occur when structural material cannot undergo extensive, or even limited, plastic deformation due to irradiation exposure. Certain irradiation conditions lead to the development of a damage microstructure where plastic flow is limited to very small volumes or regions of material, as opposed to the general plastic flow in unexposed materials. This process is referred to as flow localization or plastic instability. The true stress at the onset of necking is a constant regardless of the irradiation level. It is called 'critical stress' and this critical stress has strong temperature dependence. Interrupted tensile testes of 316L SS have been performed to investigate the microstructure evolution and competing mechanism between mechanic twinning and planar slip which are believed to be the controlling mechanism for flow localization. Deformation twinning is the major contribution of strain hardening and good ductility for low temperatures, and the activation of twinning system is determined by the critical twinning stress. Phases transform and texture analyses are also discussed in this study. Finite element analysis is carried out to complement the microstructural analysis and for the prediction of materaials performance with and without stress concentration and irradiation.

  11. Irradiation of electronic components and circuits at the Portuguese Research Reactor: Lessons learned

    SciTech Connect

    Marques, J.G.; Ramos, A.R.; Fernandes, A.C.; Santos, J.P.

    2015-07-01

    The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, since the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor

  12. Response of structures to energetic events for the Savannah River Site production reactors probabilistic risk assessment

    SciTech Connect

    Santa Cruz, S.M.; Smith, D.C.; Yau, W.F.

    1992-10-01

    The response of structures to energetic events postulated to arise in a probabilistic risk assessment (PRA) of a Savannah River Site (SRS) production reactor is addressed. Energetic events that arise in PRAs can damage structures and therefore have a significant influence on subsequent accident progression. Consequently, the structural response is important to the calculated risk of operating a plant. Difficulties are encountered, however, in the analysis of structural response of components to energetic loadings. First, the analysis of energetic events often does not provide well-defined static or dynamic loads acting on the structures. Secondly, risk assessments, by their nature, address a wide range of events that are not necessarily precisely defined. This paper describes an approach taken to develop the structural analysis required to support the PRA of the SRS production reactor, that overcomes these difficulties.

  13. Response of structures to energetic events for the Savannah River Site production reactors probabilistic risk assessment

    SciTech Connect

    Santa Cruz, S.M.; Smith, D.C. ); Yau, W.F. )

    1992-01-01

    The response of structures to energetic events postulated to arise in a probabilistic risk assessment (PRA) of a Savannah River Site (SRS) production reactor is addressed. Energetic events that arise in PRAs can damage structures and therefore have a significant influence on subsequent accident progression. Consequently, the structural response is important to the calculated risk of operating a plant. Difficulties are encountered, however, in the analysis of structural response of components to energetic loadings. First, the analysis of energetic events often does not provide well-defined static or dynamic loads acting on the structures. Secondly, risk assessments, by their nature, address a wide range of events that are not necessarily precisely defined. This paper describes an approach taken to develop the structural analysis required to support the PRA of the SRS production reactor, that overcomes these difficulties.

  14. RELIABILITY MODELS OF AGING PASSIVE COMPONENTS INFORMED BY MATERIALS DEGRADATION METRICS TO SUPPORT LONG-TERM REACTOR OPERATIONS

    SciTech Connect

    Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y.

    2012-05-01

    Paper describes a methodology for the synthesis of nuclear power plant service data with expert-elicited materials degradation information to estimate the future failure rates of passive components. This method should be an important resource to long-term plant operations and reactor life extension. Conventional probabilistic risk assessments (PRAs) are not well suited to addressing long-term reactor operations. Since passive structures and components are among those for which replacement can be least practical, they might be expected to contribute increasingly to risk in an aging plant; yet, passives receive limited treatment in PRAs. Furthermore, PRAs produce only snapshots of risk based on the assumption of time-independent component failure rates. This assumption is unlikely to be valid in aging systems. The treatment of aging passive components in PRA presents challenges. Service data to quantify component reliability models are sparse, and this is exacerbated by the greater data demands of age-dependent reliability models. Another factor is that there can be numerous potential degradation mechanisms associated with the materials and operating environment of a given component. This deepens the data problem since risk-informed management of component aging will demand an understanding of the long-term risk significance of individual degradation mechanisms. In this paper we describe a Bayesian methodology that integrates metrics of materials degradation susceptibility with available plant service data to estimate age-dependent passive component reliabilities. Integration of these models into conventional PRA will provide a basis for materials degradation management informed by predicted long-term operational risk.

  15. Fuzzy Clusterwise Generalized Structured Component Analysis

    ERIC Educational Resources Information Center

    Hwang, Heungsun; Desarbo, Wayne S.; Takane, Yoshio

    2007-01-01

    Generalized Structured Component Analysis (GSCA) was recently introduced by Hwang and Takane (2004) as a component-based approach to path analysis with latent variables. The parameters of GSCA are estimated by pooling data across respondents under the implicit assumption that they all come from a single, homogenous group. However, as has been…

  16. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    SciTech Connect

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

  17. Development of Underwater Laser Cladding and Underwater Laser Seal Welding Techniques for Reactor Components

    NASA Astrophysics Data System (ADS)

    Hino, Takehisa; Tamura, Masataka; Tanaka, Yoshimi; Kouno, Wataru; Makino, Yoshinobu; Kawano, Shohei; Matsunaga, Keiji

    Stress corrosion cracking (SCC) has been reported at the aged components in many nuclear power plants. Toshiba has been developing the underwater laser welding. This welding technique can be conducted without draining the water in the reactor vessel. It is beneficial for workers not to exposure the radiation. The welding speed can be attaining twice as fast as that of Gas Tungsten Arc Welding (GTAW). The susceptibility of SCC can also be lower than the Alloy 600 base metal.

  18. Thermal-hydraulic limitations on water-cooled fusion reactor components

    SciTech Connect

    Cha, Y.S.; Misra, B.

    1986-01-01

    An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex geometries, especially under asymmetrically heated conditions. Since the data presented herein are concerned primarily with thermal protection of the reactor components, the final design should consider other factors such as thermal stresses, temperature limits, and fatigue.

  19. Swelling in light water reactor internal components: Insights from computational modeling

    SciTech Connect

    Stoller, Roger E.; Barashev, Alexander V.; Golubov, Stanislav I.

    2015-08-01

    A modern cluster dynamics model has been used to investigate the materials and irradiation parameters that control microstructural evolution under the relatively low-temperature exposure conditions that are representative of the operating environment for in-core light water reactor components. The focus is on components fabricated from austenitic stainless steel. The model accounts for the synergistic interaction between radiation-produced vacancies and the helium that is produced by nuclear transmutation reactions. Cavity nucleation rates are shown to be relatively high in this temperature regime (275 to 325°C), but are sensitive to assumptions about the fine scale microstructure produced under low-temperature irradiation. The cavity nucleation rates observed run counter to the expectation that void swelling would not occur under these conditions. This expectation was based on previous research on void swelling in austenitic steels in fast reactors. This misleading impression arose primarily from an absence of relevant data. The results of the computational modeling are generally consistent with recent data obtained by examining ex-service components. However, it has been shown that the sensitivity of the model s predictions of low-temperature swelling behavior to assumptions about the primary damage source term and specification of the mean-field sink strengths is somewhat greater that that observed at higher temperatures. Further assessment of the mathematical model is underway to meet the long-term objective of this research, which is to provide a predictive model of void swelling at relevant lifetime exposures to support extended reactor operations.

  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 pressure vessel structural integrity research

    SciTech Connect

    Pennell, W.E.; Corwin, W.R.

    1995-04-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallows surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT{sub NDT}) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on a shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.

  2. Control of activation levels to simplify waste management of fusion reactor ferritic steel components

    SciTech Connect

    Wiffen, F.W.; Santoro, R. T.

    1983-01-01

    Activation characteristics of a material for service in the neutron flux of a fusion reactor first wall fall into three areas: waste management, reactor maintenance and repair, and safety. Of these, the waste management area is the most likely to impact the public acceptance of fusion reactors for power generation. The decay of the activity in steels within tens of years could lead to simplified waste disposal or possibly even to materials recycle. Whether or not these can be achieved will be controlled by (1) selection of alloying elements, (2) control of critical impurity elements, and (3) control of cross contamination from other reactor components. Several criteria can be used to judge the acceptability of potential alloying elements in iron, and to define the limits on content of critical impurity elements. One approach is to select and limit alloying additions on the basis of the activity. If material recycle is a goal, N, Al, Ni, Cu, Nb, and Mo must be excluded. If simplified waste storage by shallow land burial is the goal, regulations limit the concentration of only a few isotopes. For first-wall material that will be exposed to 9 MW-y/m/sup 2/ service, allowable initial concentration limits include (in at. ppM) Ni < 20,000; Mo < 3650; N < 3650, Cu < 2400; and Nb < 1.0. The other constituent elements of ferritic steels will not be limited. Possible substitutes for the molybdenum normally used to strengthen the steels include W, Ta, Ti, and V.

  3. 10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL ...

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

    10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL CROSS SECTION. Giffals & Vallet, Inc., L. Rosetti, Associated Architects and Engineers, Detroit, Michigan; and U.S. Army Engineer Division, New England Corps of Engineers, Boston, Massachusetts. Drawing Number 35-84-04. (Original: AMTL Engineering Division, Watertown). - Watertown Arsenal, Building No. 100, Wooley Avenue, Watertown, Middlesex County, MA

  4. DEGRADATION SUSCEPTIBILITY METRICS AS THE BASES FOR BAYESIAN RELIABILITY MODELS OF AGING PASSIVE COMPONENTS AND LONG-TERM REACTOR RISK

    SciTech Connect

    Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y.; Ford, Benjamin E.

    2011-07-17

    Conventional probabilistic risk assessments (PRAs) are not well-suited to addressing long-term reactor operations. Since passive structures, systems and components are among those for which refurbishment or replacement can be least practical, they might be expected to contribute increasingly to risk in an aging plant. Yet, passives receive limited treatment in PRAs. Furthermore, PRAs produce only snapshots of risk based on the assumption of time-independent component failure rates. This assumption is unlikely to be valid in aging systems. The treatment of aging passive components in PRA does present challenges. First, service data required to quantify component reliability models are sparse, and this problem is exacerbated by the greater data demands of age-dependent reliability models. A compounding factor is that there can be numerous potential degradation mechanisms associated with the materials, design, and operating environment of a given component. This deepens the data problem since the risk-informed management of materials degradation and component aging will demand an understanding of the long-term risk significance of individual degradation mechanisms. In this paper we describe a Bayesian methodology that integrates the metrics of materials degradation susceptibility being developed under the Nuclear Regulatory Commission's Proactive Management of Materials of Degradation Program with available plant service data to estimate age-dependent passive component reliabilities. Integration of these models into conventional PRA will provide a basis for materials degradation management informed by the predicted long-term operational risk.

  5. EVALUATION OF ACTIVATION PRODUCTS IN REMAINING IN REMAINING K-, L- AND C-REACTOR STRUCTURES

    SciTech Connect

    Vinson, D.; Webb, R.

    2010-09-30

    An analytic model and calculational methodology was previously developed for P-reactor and R-reactor to quantify the radioisotopes present in Savannah River Site (SRS) reactor tanks and the surrounding structural materials as a result of neutron activation of the materials during reactor operation. That methodology has been extended to K-reactor, L-reactor, and C-reactor. The analysis was performed to provide a best-estimate source term input to the Performance Assessment for an in-situ disposition strategy by Site Decommissioning and Demolition (SDD). The reactor structure model developed earlier for the P-reactor and R-reactor analyses was also used for the K-reactor and L-reactor. The model was suitably modified to handle the larger Creactor tank and associated structures. For all reactors, the structure model consisted of 3 annular zones, homogenized by the amount of structural materials in the zone, and 5 horizontal layers. The curie content on an individual radioisotope basis and total basis for each of the regions was determined. A summary of these results are provided herein. The efficacy of this methodology to accurately predict the radioisotopic content of the reactor systems in question has been demonstrated and is documented in Reference 1. As noted in that report, results for one reactor facility cannot be directly extrapolated to other SRS reactors.

  6. COMPONENT DEGRADATION SUSCEPTIBILITIES AS THE BASES FOR MODELING REACTOR AGING RISK

    SciTech Connect

    Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y.

    2010-07-18

    The extension of nuclear power plant operating licenses beyond 60 years in the United States will be necessary if we are to meet national energy needs while addressing the issues of carbon and climate. Characterizing the operating risks associated with aging reactors is problematic because the principal tool for risk-informed decision-making, Probabilistic Risk Assessment (PRA), is not ideally-suited to addressing aging systems. The components most likely to drive risk in an aging reactor - the passives - receive limited treatment in PRA, and furthermore, standard PRA methods are based on the assumption of stationary failure rates: a condition unlikely to be met in an aging system. A critical barrier to modeling passives aging on the wide scale required for a PRA is that there is seldom sufficient field data to populate parametric failure models, and nor is there the availability of practical physics models to predict out-year component reliability. The methodology described here circumvents some of these data and modeling needs by using materials degradation metrics, integrated with conventional PRA models, to produce risk importance measures for specific aging mechanisms and component types. We suggest that these measures have multiple applications, from the risk-screening of components to the prioritization of materials research.

  7. Mechanical properties of thermally aged cast stainless steels from shippingport reactor components.

    SciTech Connect

    Chopra, O. K.; Shack, W. J.; Energy Technology

    1995-06-07

    Thermal embrittlement of static-cast CF-8 stainless steel components from the decommissioned Shippingport reactor has been characterized. Cast stainless steel materials were obtained from four cold-leg check valves, three hot-leg main shutoff valves, and two pump volutes. The actual time-at-temperature for the materials was {approx}13 y at {approx}281 C (538 F) for the hot-leg components and {approx}264 C (507 F) for the cold-leg components. Baseline mechanical properties for as-cast material were determined from tests on either recovery-annealed material, i.e., annealed for 1 h at 550 C and then water quenched, or material from the cooler region of the component. The Shippingport materials show modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength because of relatively low service temperatures and ferrite content of the steel. The procedure and correlations developed at Argonne National Laboratory for estimating mechanical properties of cast stainless steels predict accurate or slightly lower values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and JIC of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predicted the mechanical properties of the Ringhals 2 reactor hot- and crossover-leg elbows (CF-8M steel) after service of {approx}15 y and the KRB reactor pump cover plate (CF-8) after {approx}8 y of service.

  8. Mechanical properties of thermally aged cast stainless steels from Shippingport reactor components

    SciTech Connect

    Chopra, O.K.; Shack, W.J.

    1995-04-01

    Thermal embrittlement of static-cast CF-8 stainless steel components from the decommissioned Shippingport reactor has been characterized. Cast stainless steel materials were obtained from four cold-leg check valves, three hot-leg main shutoff valves, and two pump volutes. The actual time-at-temperature for the materials was {approximately}13 y at {approximately}281 C (538 F) for the hot-leg components and {approximately}264 C (507 F) for the cold-leg components. Baseline mechanical properties for as-cast material were determined from tests on either recovery-annealed material, i.e., annealed for 1 h at 550 C and then water quenched, or material from the cooler region of the component. The Shippingport materials show modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength because of relatively low service temperatures and ferrite content of the steel. The procedure and correlations developed at Argonne National Laboratory for estimating mechanical properties of cast stainless steels predict accurate or slightly lower values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J{sub IC} of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predicted the mechanical properties of the Ringhals 2 reactor hot and crossover-leg elbows (CF-8M steel) after service of {approximately} 15 y and the KRB reactor pump cover plate (CF-8) after {approximately} 8 y of service.

  9. Application of amorphous filler metals in production of fusion reactor high heat flux components

    SciTech Connect

    Kalin, B.A.; Fedotov, V.T.; Grigoriev, A.E.

    1994-12-31

    The technology of Al-Si, Zr-Ti-Be and Ti-Zr-Cu-Ni amorphous filler metals for Be and graphite brazing with Cu, Mo and V was developed. The fusion reactor high heat flux components from Cu-Be, Cu-graphite, Mo-Be, Mo-graphite, V-Re and V-graphite materials were produced by brazing. Every component represents metallic base, to which Be or graphite plates are brazed. The distance between plates was equal 0.2 times the plate height. These components were irradiated by hydrogen plasma with 5 x 10{sup 6} W/m{sup 2} power. The microstructure and the element distribution in the brazed zone were investigated before and after heat plasma irradiation. Topography graphite plate surfaces and topography of metal surfaces between plates were also investigated after heat plasma irradiation. The results of microstructure investigation and material erosion are discussed.

  10. Compression Strength of Composite Primary Structural Components

    NASA Technical Reports Server (NTRS)

    Johnson, Eric R.

    1998-01-01

    Research conducted under NASA Grant NAG-1-537 focussed on the response and failure of advanced composite material structures for application to aircraft. Both experimental and analytical methods were utilized to study the fundamental mechanics of the response and failure of selected structural components subjected to quasi-static loads. Most of the structural components studied were thin-walled elements subject to compression, such that they exhibited buckling and postbuckling responses prior to catastrophic failure. Consequently, the analyses were geometrically nonlinear. Structural components studied were dropped-ply laminated plates, stiffener crippling, pressure pillowing of orthogonally stiffened cylindrical shells, axisymmetric response of pressure domes, and the static crush of semi-circular frames. Failure of these components motivated analytical studies on an interlaminar stress postprocessor for plate and shell finite element computer codes, and global/local modeling strategies in finite element modeling. These activities are summarized in the following section. References to literature published under the grant are listed on pages 5 to 10 by a letter followed by a number under the categories of journal publications, conference publications, presentations, and reports. These references are indicated in the text by their letter and number as a superscript.

  11. Development of Fast Reactor Structural Integrity Monitoring Technology Using Optical Fiber Sensors

    NASA Astrophysics Data System (ADS)

    Matsuba, Ken-Ichi; Ito, Chikara; Kawahara, Hirotaka; Aoyama, Takafumi

    Significant thermal stresses are loaded onto the structures of sodium-cooled fast reactor (SFR) due to high temperature and large temperature gradients associated with employing sodium coolant with its high thermal conductivity and low heat capacity. Therefore, it is important to monitor the temperature variation, related stress and displacement, and vibration in the cooling system piping and components in order to assure structural integrity while the reactor plant is in-service. SFR structural integrity monitoring can be enhanced by an optical fiber sensor, which is capable of continuous or dispersed distribution measurements of various properties such as radiation dose, temperature, strain, displacement and acceleration. In the experimental fast reactor Joyo, displacement and vibration measurements of the primary cooling system have been carried out using Fiber Bragg Grating (FBG) sensors to evaluate the durability and measurement accuracy of FBG sensors in a high gamma-ray environment. The data were successfully obtained with no significant signal loss up to an accumulated gamma-ray dose of approximately 4×104 Gy corresponding to 120 EFPDs (effective full power days) operation. Measured displacement of the piping support was nearly equal to the calculated thermal displacement. Measured vibration power spectra of the piping support were similar to those measured with a reference acceleration sensor. The measured results indicate that the FBG sensor is suitable for monitoring the displacement and vibration aspects of fast reactor cooling system integrity in a high gamma-ray environment.

  12. Packaging, Transportation, and Disposal Logistics for Large Radioactively Contaminated Reactor Decommissioning Components

    SciTech Connect

    Lewis, Mark S.

    2008-01-15

    The packaging, transportation and disposal of large, retired reactor components from operating or decommissioning nuclear plants pose unique challenges from a technical as well as regulatory compliance standpoint. In addition to the routine considerations associated with any radioactive waste disposition activity, such as characterization, ALARA, and manifesting, the technical challenges for large radioactively contaminated components, such as access, segmentation, removal, packaging, rigging, lifting, mode of transportation, conveyance compatibility, and load securing require significant planning and execution. In addition, the current regulatory framework, domestically in Titles 49 and 10 and internationally in TS-R-1, does not lend itself to the transport of these large radioactively contaminated components, such as reactor vessels, steam generators, reactor pressure vessel heads, and pressurizers, without application for a special permit or arrangement. This paper addresses the methods of overcoming the technical and regulatory challenges. The challenges and disposition decisions do differ during decommissioning versus component replacement during an outage at an operating plant. During decommissioning, there is less concern about critical path for restart and more concern about volume reduction and waste minimization. Segmentation on-site is an available option during decommissioning, since labor and equipment will be readily available and decontamination activities are routine. The reactor building removal path is also of less concern and there are more rigging/lifting options available. Radionuclide assessment is necessary for transportation and disposal characterization. Characterization will dictate the packaging methodology, transportation mode, need for intermediate processing, and the disposal location or availability. Characterization will also assist in determining if the large component can be transported in full compliance with the transportation

  13. Effect of azimuthally asymmetric reactor components on a parallel plate capacitively coupled plasma

    SciTech Connect

    Kenney, Jason A.; Rauf, Shahid; Collins, Ken

    2009-11-15

    A three-dimensional fluid plasma model is used to investigate the impact of azimuthally asymmetric reactor components on spatial characteristics of parallel plate capacitively coupled plasmas. We consider three scenarios: high frequency (13.56 MHz) argon discharges with, separately, an off-axis circular plate surrounding the bottom electrode and an access port opening in the reactor sidewall, and a very high frequency (162 MHz) argon discharge with nonparallel electrodes. For the reactor with off-axis plate, both the Ar{sup +} density and flux are strongly perturbed toward the direction of maximum grounded surface area, with azimuthal variation in ion flux up to 10%. Perturbations in Ar{sup +} density due to the access port opening are localized to the region near the access port, and the impact on ion flux in the interelectrode region is minimal. Finally, the nonparallel electrodes result in a significant change in the location and shape of the Ar{sup +} density profile, going from a center-peaked discharge with parallel electrodes to a flattened off-center profile when tilted less than 1 deg. with a nominal 5 cm gap.

  14. Use of principal components analysis for reactor accident consequence evaluation and a comparison with other techniques

    SciTech Connect

    Gudiksen, P.H.; Walton, J.J.; Alpert, D.J.; Johnson, J.D.

    1981-04-01

    The consequences of a potential reactor accident are normally characterized in terms of frequency distributions for exceeding specified surface air concentrations and deposition levels since these may be directly related to individual or population radiation exposures. Since an accidental release of radioactivity could occur at any time, the frequency distributions are determined by performing a large number of calculations that include a variety of possible release characteristics and meteorological situations. Performing such a large number of calculations is generally only feasible with relatively simple analytical models that utilize only the meteorological observations from the reactor site to describe the transport and dispersion of the radioactive material out to distances of about 100 km from the reactor. The purpose of this work was to investigate the possibility of utilizing three-dimensional models for consequence analysis, since these are capable of including meteorological data from multiple sites and the effects of topography on the transport and dispersion of airborne radioactivity over the region of concern. The approach to this problem was to investigate the feasibility of using the principal components analysis (PCA) technique for identifying wind patterns and their frequencies and temporal variations.

  15. A review of the US joining technologies for plasma facing components in the ITER fusion reactor

    SciTech Connect

    Odegard, B.C. Jr.; Cadden, C.H.; Watson, R.D.; Slattery, K.T.

    1998-02-01

    This paper is a review of the current joining technologies for plasma facing components in the US for the International Thermonuclear Experimental Reactor (ITER) project. Many facilities are involved in this project. Many unique and innovative joining techniques are being considered in the quest to join two candidate armor plate materials (beryllium and tungsten) to a copper base alloy heat sink (CuNiBe, OD copper, CuCrZr). These techniques include brazing and diffusion bonding, compliant layers at the bond interface, and the use of diffusion barrier coatings and diffusion enhancing coatings at the bond interfaces. The development and status of these joining techniques will be detailed in this report.

  16. Method for fabricating wrought components for high-temperature gas-cooled reactors and product

    DOEpatents

    Thompson, Larry D.; Johnson, Jr., William R.

    1985-01-01

    A method and alloys for fabricating wrought components of a high-temperature gas-cooled reactor are disclosed. These wrought, nickel-based alloys, which exhibit strength and excellent resistance to carburization at elevated temperatures, include aluminum and titanium in amounts and ratios to promote the growth of carburization resistant films while preserving the wrought character of the alloys. These alloys also include substantial amounts of molybdenum and/or tungsten as solid-solution strengtheners. Chromium may be included in concentrations less than 10% to assist in fabrication. Minor amounts of carbon and one or more carbide-forming metals also contribute to high-temperature strength.

  17. Components of ice nucleation structures of bacteria.

    PubMed Central

    Turner, M A; Arellano, F; Kozloff, L M

    1991-01-01

    Nonprotein components attached to the known protein product of the inaZ gene of Pseudomonas syringae have been identified and shown to be necessary for the most efficient ice nucleation of supercooled H2O. Previous studies have shown that cultures of Ina+ bacteria have cells with three major classes of ice-nucleating structures with readily differentiated activities. Further, some cells in the culture have nucleating activities intermediate between those of the different classes and presumably have structures that are biosynthetic intermediates between those of the different classes. Since these structures cannot be readily isolated and analyzed, their components have been identified by the use of specific enzymes or chemical probes, by direct incorporation of labeled precursors, and by stimulation of the formation of specific classes of freezing structures by selective additions to the growth medium. From these preliminary studies it appears that the most active ice nucleation structure (class A) contains the ice nucleation protein linked to phosphatidylinositol and mannose, probably as a complex mannan, and possibly glucosamine. These nonprotein components are characteristic of those used to anchor external proteins to cell membranes of eucaryotic cells and suggest that a similar but not identical anchoring mechanism is required for efficient ice nucleation structure. The class B structure has been found to contain protein presumably linked to the mannan and glucosamine moieties but definitely not to the phosphatidylinositol. The class C structure, which has the poorest ice nucleation activity, appears to be the ice nucleation protein linked to a few mannose residues and to be partially imbedded in the outer cell membrane. Images FIG. 1 FIG. 2 FIG. 5 FIG. 9 FIG. 15 FIG. 16 PMID:1917876

  18. 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)

  19. 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.

  20. Reactor plasma facing component designs based on liquid metal concepts supported in porous systems

    NASA Astrophysics Data System (ADS)

    Tabarés, F. L.; Oyarzabal, E.; Martin-Rojo, A. B.; Tafalla, D.; de Castro, A.; Soleto, A.

    2017-01-01

    The use of liquid metals (LMs) as plasma facing components in fusion devices was proposed as early as 1970 for a field reversed concept and inertial fusion reactors. The idea was extensively developed during the APEX Project, at the turn of the century, and it is the subject at present of the biennial International Symposium on Lithium Applications (ISLA), whose fourth meeting took place in Granada, Spain at the end of September 2015. While liquid metal flowing concepts were specially addressed in USA research projects, the idea of embedding the metal in a capillary porous system (CPS) was put forwards by Russian teams in the 1990s, thus opening the possibility of static concepts. Since then, many ideas and accompanying experimental tests in fusion devices and laboratories have been produced, involving a large fraction of countries within the international fusion community. Within the EUROFusion Roadmap, these activities are encompassed into the working programs of the plasma facing components (PFC) and divertor tokamak test (DTT) packages. In this paper, a review of the state of the art in concepts based on the CPS set-up for a fusion reactor divertor target, aimed at preventing the ejection of the liquid metal by electro-magnetic (EM) forces generated under plasma operation, is described and required R+D activities on the topic, including ongoing work at CIEMAT specifically oriented to filling the remaining gaps, are stressed.

  1. Review of leakage-flow-induced vibrations of reactor components. [LMFBR

    SciTech Connect

    Mulcahy, T.M.

    1983-05-01

    The primary-coolant flow paths of a reactor system are usually subject to close scrutiny in a design review to identify potential flow-induced vibration sources. However, secondary-flow paths through narrow gaps in component supports, which parallel the primary-flow path, occasionally are the excitation source for significant vibrations even though the secondary-flow rates are orders of magnitude smaller than the primary-flow rate. These so-called leakage flow problems are reviewed here to identify design features and excitation sources that should be avoided. Also, design rules of thumb are formulated that can be employed to guide a design, but quantitative prediction of component response is found to require scale-model testing.

  2. Modeling structural dynamic behavior of SSME components

    NASA Technical Reports Server (NTRS)

    Kiefling, Larry A.; Saxon, J. B.; Prickett, T. L.

    1991-01-01

    FEM studies are presented of the nozzle and the low-pressure fuel-pump inducer designs for the Space Shuttle Main Engine (SSME) to analyze the effects of structural vibrations. FEM preprocessing software based on a CAD system is employed to develop a model of the component's sophisticated geometry. The nozzle geometry is also defined by means of the preprocessing technique and subsequently analyzed with respect to time-transient loading. The analysis is conducted with a Cray supercomputer using the SPAR/EAL FEM program. The investigation of the nozzle demonstrates the advantageous use of symmetry in the determination of nozzle response to SSME start-up transients. Plots of time vs strain are developed for gages on the nozzle wall and steerhorn tubing. The results of the inducer modeling are found to be adequate for investigating the component's principle modes, and the nozzle results indicate the suitability of the FEM techniques for optimizing the design of engine components.

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

    SciTech Connect

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

    2007-06-30

    -flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower

  4. System for inspecting large size structural components

    DOEpatents

    Birks, Albert S.; Skorpik, James R.

    1990-01-01

    The present invention relates to a system for inspecting large scale structural components such as concrete walls or the like. The system includes a mobile gamma radiation source and a mobile gamma radiation detector. The source and detector are constructed and arranged for simultaneous movement along parallel paths in alignment with one another on opposite sides of a structural component being inspected. A control system provides signals which coordinate the movements of the source and detector and receives and records the radiation level data developed by the detector as a function of source and detector positions. The radiation level data is then analyzed to identify areas containing defects corresponding to unexpected variations in the radiation levels detected.

  5. 54. ARAII. Structural steel framing for bottom SL1 reactor building. ...

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

    54. ARA-II. Structural steel framing for bottom SL-1 reactor building. October 16, 1957. Ineel photo no. 57-5186. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  6. Greenstone belts: Their components and structure

    NASA Technical Reports Server (NTRS)

    Vearncombe, J. R.; Barton, J. M., Jr.; Vanreenen, D. D.; Phillips, G. N.; Wilson, A. H.

    1986-01-01

    Greenstone sucessions are defined as the nongranitoid component of granitoid-greenstone terrain and are linear to irregular in shape and where linear are termed belts. The chemical composition of greenstones is described. Also discussed are the continental environments of greenstone successions. The effects of contact with granitoids, geophysical properties, recumbent folds and late formation structures upon greenstones are examined. Large stratigraphy thicknesses are explained.

  7. Relationships between chemical oxygen demand (COD) components and toxicity in a sequential anaerobic baffled reactor/aerobic completely stirred reactor system treating Kemicetine.

    PubMed

    Sponza, Delia Teresa; Demirden, Pinar

    2010-04-15

    In this study the interactions between toxicity removals and Kemicetine, COD removals, intermediate products of Kemicetine and COD components (CODs originating from slowly degradable organics, readily degradable organics, inert microbial products and from the inert compounds) were investigated in a sequential anaerobic baffled reactor (ABR)/aerobic completely stirred tank reactor (CSTR) system with a real pharmaceutical wastewater. The total COD and Kemicetine removal efficiencies were 98% and 100%, respectively, in the sequential ABR/CSTR systems. 2-Amino-1 (p-nitrophenil)-1,3 propanediol, l-p-amino phenyl, p-amino phenol and phenol were detected in the ABR as the main readily degradable inter-metabolites. In the anaerobic ABR reactor, the Kemicetin was converted to corresponding inter-metabolites and a substantial part of the COD was removed. In the aerobic CSTR reactor the inter-metabolites produced in the anaerobic reactor were completely removed and the COD remaining from the anerobic reactor was biodegraded. It was found that the COD originating from the readily degradable organics did not limit the anaerobic degradation process, while the CODs originating from the slowly degradable organics and from the inert microbial products significantly decreased the anaerobic ABR reactor performance. The acute toxicity test results indicated that the toxicity decreased from the influent to the effluent of the aerobic CSTR reactor. The ANOVA test statistics showed that there was a strong linear correlation between acute toxicity, CODs originating from the slowly degradable organics and inert microbial products. A weak correlation between acute toxicity and CODs originating from the inert compounds was detected.

  8. Method for producing components with internal architectures, such as micro-channel reactors, via diffusion bonding sheets

    DOEpatents

    Alman, David E.; Wilson, Rick D.; Davis, Daniel L.

    2011-03-08

    This invention relates to a method for producing components with internal architectures, and more particularly, this invention relates to a method for producing structures with microchannels via the use of diffusion bonding of stacked laminates. Specifically, the method involves weakly bonding a stack of laminates forming internal voids and channels with a first generally low uniaxial pressure and first temperature such that bonding at least between the asperites of opposing laminates occurs and pores are isolated in interfacial contact areas, followed by a second generally higher isostatic pressure and second temperature for final bonding. The method thereby allows fabrication of micro-channel devices such as heat exchangers, recuperators, heat-pumps, chemical separators, chemical reactors, fuel processing units, and combustors without limitation on the fin aspect ratio.

  9. Catalytic reactor

    SciTech Connect

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

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

  10. 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.

  11. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

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

  12. Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

    SciTech Connect

    Owen, M.B. Smith, R.; McNeil, J.

    1997-04-01

    Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs.

  13. Structural reliability analysis of laminated CMC components

    NASA Technical Reports Server (NTRS)

    Duffy, Stephen F.; Palko, Joseph L.; Gyekenyesi, John P.

    1991-01-01

    For laminated ceramic matrix composite (CMC) materials to realize their full potential in aerospace applications, design methods and protocols are a necessity. The time independent failure response of these materials is focussed on and a reliability analysis is presented associated with the initiation of matrix cracking. A public domain computer algorithm is highlighted that was coupled with the laminate analysis of a finite element code and which serves as a design aid to analyze structural components made from laminated CMC materials. Issues relevant to the effect of the size of the component are discussed, and a parameter estimation procedure is presented. The estimation procedure allows three parameters to be calculated from a failure population that has an underlying Weibull distribution.

  14. FDC, rapid fabrication of structural components

    SciTech Connect

    Agarwala, M.K.; Bandyopadhyay, A.; Weeren, R. van; Safari, A.; Danforth, S.C.; Langrana, N.A.; Jamalabad, V.R.; Whalen, P.J.

    1996-11-01

    Solid freeform fabrication (SFF) is used to make 3-D components directly from computer-aided design (CAD) files. Many SFF techniques have been developed to fabricate parts and prototypes from CAD without hard tooling, dies or molds. Most of these techniques have been commercialized for fabrication of polymer and plastic parts for design verification and form and fit. Other SFF techniques are being developed for production of ceramic components with functional properties. One such technique, called fused deposition of ceramics (FDC), has been developed and demonstrated for structural ceramics. FDC is based on existing fused deposition modeling (FDM{trademark}) technology, commercialized by Stratasys Inc. (Eden Prairie, Minn.), for processing of polymers and waxes. High-green-density, simple- and complex-shaped silicon nitride parts have been formed by fused deposition of ceramics.

  15. Design considerations for ITER (International Thermonuclear Experimental Reactor) plasma facing components

    SciTech Connect

    McGrath, R.T.; Koski, J.A.; Watson, R.D.; Causey, R.A.; Croessmann, C.D.; Dempsey, J.F.; Hosking, M.; Neimer, K.A.; Russo, A.J.; Salmonson, J.C.; Stephens, J.; Smith, M.F.; Watkins, J.G.; Whitley, J.B.

    1989-07-01

    The International Thermonuclear Experimental Reactor (ITER) is a joint design and R D project involving the USA, the Soviet Union, Japan and the European Community. These international partners are working together on the design of a fusion tokamak reactor that will operate in the D-T ignition regime. This report compiles the contributions to ITER made by Sandia National Laboratories in the area of design and R D for plasma facing components, such as the first wall and divertor. The following topics are discussed: divertor fabrication issues, divertor thermal-hydraulic analysis, separatrix sweeping effects, divertor tile 2-D stress analysis, electromechanical disruption effects, runaway electron and intense energy deposition analyses, lifetime analysis and tritium retention in plasma facing materials. Material properties for pyrolytic graphite and beryllium are presented. Use of pyrolytic graphite as the plasma facing material allows for operation with thicker graphite armor at the design heat flux level of 10 MW/m/sup 2/. The design of a divertor coated with plasma sprayed beryllium is presented as an attractive alternative to pyrolytic graphite armor tiles. Finally, the Sandia research and development plan for ITER is discussed. 82 figs.

  16. Mechanical-property degradation of cast stainless steel components from the Shippingport reactor

    SciTech Connect

    Chopra, O.K.

    1991-10-01

    The mechanical properties of cast stainless steels from the Shippingport reactor have been characterized. Baseline properties for unaged materials were obtained from tests on either recovery-annealed material or material from a cooler region of the component. The materials exhibited modest decrease in impact energy and fracture toughness and a small increase in tensile strength. The fracture toughness J-R curve, J{sub IC} value, tensile flow stress, and Charpy-impact energy of the materials showed very good agreement with estimations based on accelerated laboratory aging studies. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy that would be achieved after long-term aging, were established from materials that were aged further in the laboratory at temperatures between 320 and 400{degrees}C. The results showed very good agreement with estimates; the activation energies ranged from 125 to 250 kJ/mole and the minimum room temperature impact energy was <75 J/cm{sup 2}. The estimated impact energy and fracture toughness J-R curve for materials from the Ringhals reactor hot and crossover-leg elbows are also presented.

  17. Structural Integrity of Water Reactor Pressure Boundary Components.

    DTIC Science & Technology

    1979-12-31

    the SEM micrograph in Fig. 38. Further research suggested that large amounts of segregation particles and manganese-sulfide inclusions in the A508...interface. A large number of segregation particles is present in the miorostructure. 66 - ‘ V :s...inferior quality) A508 forging specimen, indicates that overt segregation and inclusion content could lead to hydrogen embrittlement mechanisms much

  18. A procedure for evaluating residual life of major components in light water reactors

    SciTech Connect

    Uchida, S.; Fujimori, H.; Ibe, E.; Kuniya, J.; Hayashi, M.; Fuse, M.; Yamauchi, K.

    1995-12-31

    A computer program for evaluating residual life of major components in boiling water reactors is proposed. It divides the stress corrosion cracking process into two stages; a probabilistic crack generation stage and a deterministic crack propagation one. The minimum period of the crack generation stage is evaluated assuming an exponential distribution of the stage. The crack propagation rate is calculated by the slip-dissolution/film-rupture model. The neutron flux and fluence dependence of the neutron radiation effects on material properties was evaluated by using theoretical models of radiation damage. The computer program works on an engineering work station. Evaluated results are displayed as a map of the residual life, or as graphs of crack length evolution.

  19. Lightweight Thermoformed Structural Components and Optics

    NASA Technical Reports Server (NTRS)

    Zeiders, Glenn W.; Bradford, Larry J.

    2004-01-01

    A technique that involves the use of thermoformed plastics has been developed to enable the design and fabrication of ultra-lightweight structural components and mirrors for use in outer space. The technique could also be used to produce items for special terrestrial uses in which minimization of weight is a primary design consideration. Although the inherent strengths of thermoplastics are clearly inferior to those of metals and composite materials, thermoplastics offer a distinct advantage in that they can be shaped, at elevated temperatures, to replicate surfaces (e.g., prescribed mirror surfaces) precisely. Furthermore, multiple elements can be bonded into structures of homogeneous design that display minimal thermal deformation aside from simple expansion. The design aspect of the present technique is based on the principle that the deflection of a plate that has internal structure depends far more on the overall thickness than on the internal details; thus, a very stiff, light structure can be made from thin plastic that is heatformed to produce a sufficiently high moment of inertia. General examples of such structures include I beams and eggcrates.

  20. Grain boundary engineering for structure materials of nuclear reactors

    SciTech Connect

    Tan, Lizhen; Allen, Todd R.; Busby, Jeremy T.

    2013-03-29

    Grain boundary engineering (GBE), primarily implemented by thermomechanical processing, is an effective and economical method of enhancing the properties of polycrystalline materials. Among the factors affecting grain boundary character distribution, literature data showed definitive effect of grain size and texture. GBE is more effective for austenitic stainless steels and Ni-base alloys compared to other structural materials of nuclear reactors, such as refractory metals, ferritic and ferritic–martensitic steels, and Zr alloys. Furthermore, GBE has shown beneficial effects on improving the strength, creep strength, and resistance to stress corrosion cracking and oxidation of austenitic stainless steels and Ni-base alloys.

  1. Nuclear reactor containment structure with continuous ring tunnel at grade

    DOEpatents

    Seidensticker, Ralph W.; Knawa, Robert L.; Cerutti, Bernard C.; Snyder, Charles R.; Husen, William C.; Coyer, Robert G.

    1977-01-01

    A nuclear reactor containment structure which includes a reinforced concrete shell, a hemispherical top dome, a steel liner, and a reinforced-concrete base slab supporting the concrete shell is constructed with a substantial proportion thereof below grade in an excavation made in solid rock with the concrete poured in contact with the rock and also includes a continuous, hollow, reinforced-concrete ring tunnel surrounding the concrete shell with its top at grade level, with one wall integral with the reinforced concrete shell, and with at least the base of the ring tunnel poured in contact with the rock.

  2. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    SciTech Connect

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177{degrees}C (350{degrees}F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program.

  3. NUHOWS - Storage and Transportation of Irradiated Reactor Components in Large Packages - 13439

    SciTech Connect

    Rae, Glen A.

    2013-07-01

    Most irradiated reactor components (hardware such as Control Rod Blades, Fuel Channels, Poison Curtains, etc.) generated at reactors previously required significant processing for size reduction due to the available transportation casks not being physically capable of containing unprocessed material. As of July 1, 2008, disposal for this typical waste class (B and C) became inaccessible (for the major part of the nation) due to the Barnwell, SC disposal facility being closed to all but its three compact states (CT, NJ and SC). Currently in the United States, most facilities are storing their irradiated hardware on-site in the spent fuel pools. Until recently with the opening of the Waste Control Specialists' Texas disposal facility, utilities faced the challenges of spent fuel pool space and capacity management. However, even with WCS's disposal availability, the site currently has annual Curie limitations for disposal, which will continue to promote interim on-site storage until such time as disposal is available. In response, Transnuclear Inc., (TN) an AREVA company, proceeded with designing a new large Radioactive Waste Container (RWC) that can be used to package irradiated hardware without the need for significant processing. The design features of the RWC allows for intermittent loadings of the hardware for better packaging efficiency, higher packaging density, space savings and reduced cost. This RWC is also compatible with TN's on-site modular vault storage system. Once completely loaded, the RWC can be transported to an on-site storage facility, an off-site storage facility and/or an available disposal facility. To accommodate the transportation, TN has designed a large transportation cask, the MP197HB. As the original design was for transporting fuel, it contains the necessary shielding to allow for the transport of unprocessed irradiated reactor components, while significantly reducing the amount of irradiated hardware shipments required with the use of

  4. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor... NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor... nuclear power or research reactor in the United States: Austria Belgium Bulgaria Canada Czech...

  5. 77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-04-19

    ... Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft interim staff guidance; Request for public... Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-4029..., Division of License Renewal, Office of Nuclear Reactor Regulation. BILLING CODE 7590-01-P...

  6. Online stress corrosion crack and fatigue usages factor monitoring and prognostics in light water reactor components: Probabilistic modeling, system identification and data fusion based big data analytics approach

    SciTech Connect

    Mohanty, Subhasish M.; Jagielo, Bryan J.; Iverson, William I.; Bhan, Chi Bum; Soppet, William S.; Majumdar, Saurin M.; Natesan, Ken N.

    2014-12-10

    Nuclear reactors in the United States account for roughly 20% of the nation's total electric energy generation, and maintaining their safety in regards to key component structural integrity is critical not only for long term use of such plants but also for the safety of personnel and the public living around the plant. Early detection of damage signature such as of stress corrosion cracking, thermal-mechanical loading related material degradation in safety-critical components is a necessary requirement for long-term and safe operation of nuclear power plant systems.

  7. Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components

    SciTech Connect

    Meyer, Ryan M.; Coble, Jamie B.; Hirt, Evelyn H.; Ramuhalli, Pradeep; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-05-17

    This report identifies a number of requirements for prognostics health management of passive systems in AdvSMRs, documents technical gaps in establishing a prototypical prognostic methodology for this purpose, and describes a preliminary research plan for addressing these technical gaps. AdvSMRs span multiple concepts; therefore a technology- and design-neutral approach is taken, with the focus being on characteristics that are likely to be common to all or several AdvSMR concepts. An evaluation of available literature is used to identify proposed concepts for AdvSMRs along with likely operational characteristics. Available operating experience of advanced reactors is used in identifying passive components that may be subject to degradation, materials likely to be used for these components, and potential modes of degradation of these components. This information helps in assessing measurement needs for PHM systems, as well as defining functional requirements of PHM systems. An assessment of current state-of-the-art approaches to measurements, sensors and instrumentation, diagnostics and prognostics is also documented. This state-of-the-art evaluation, combined with the requirements, may be used to identify technical gaps and research needs in the development, evaluation, and deployment of PHM systems for AdvSMRs. A preliminary research plan to address high-priority research needs for the deployment of PHM systems to AdvSMRs is described, with the objective being the demonstration of prototypic prognostics technology for passive components in AdvSMRs. Greater efficiency in achieving this objective can be gained through judicious selection of materials and degradation modes that are relevant to proposed AdvSMR concepts, and for which significant knowledge already exists. These selections were made based on multiple constraints including the analysis performed in this document, ready access to laboratory-scale facilities for materials testing and measurement, and

  8. Microstructural examination of fatigue accumulation in critical LWR (light water reactor) components: Final report

    SciTech Connect

    Allen, A.J.; Buttle, D.J.; Coleman, C.F.; Smith, F.A.; Smith, R.L.

    1988-01-01

    This report describes a morphological study of the feasibility of measuring the fatigue damage accumulation state of critical light water reactor (LWR) components by microstructural examination. The changes in microstructure associated with fatigue processes are first discussed so that relevant NDE measurement parameters can be identified. (The creep regime is not considered in this report). The candidate NDE techniques are then reviewed in detail under the following headings: positron annihilation, x-ray diffraction, magnetic techniques, the magnetic Barkhausen effect, the magneto acoustic technique, acoustic emission, ultrasonic techniques and finally other miscellaneous techniques applicable to fatigue damage assessment. All the feasible techniques are summarised and rated in a set of comparison tables. A possible programme for the immediate development of the positron annihilation lineshape technique is proposed. It is concluded that the most successful method of measuring the fatigue accumulation in LWR critical components in a way which relates to the intent of the ASME pressure vessel codes, is likely to be the use of several techniques together and the cross-relation of the results obtained by each. Five techniques are highlighted for immediate possible development: 'etching and surface replication', 'positron annihilation lineshapes', 'x-ray diffraction residual stress', 'acoustic emission' and 'ultrasonic surface acoustic waves'.

  9. Biofiltration of paint solvent mixtures in two reactor types: overloading by hydrophobic components.

    PubMed

    Paca, Jan; Halecky, Martin; Misiaczek, Ondrej; Jones, Kim; Kozliak, Evguenii; Sobotka, Miroslav

    2010-12-01

    Steady-state performance characteristics of a trickle bed reactor (TBR) and a biofilter (BF) in loading experiments with increasing toluene/xylenes inlet concentrations while maintaining a constant loading rate of hydrophilic components (methyl ethyl and methyl isobutyl ketones, acetone, and n-butyl acetate) of 4 g m⁻³ h⁻¹ were evaluated and compared, along with the systems' dynamic responses. At the same combined substrate loading of 55 g m⁻³ h⁻¹ for both reactors, the TBR achieved more than 1.5 times higher overall removal efficiency (RE(W)) than the BF. Increasing the loading rate of aromatics resulted in a gradual decrease of their REs. The degradation rates of acetone and n-butyl acetate were also inhibited at higher loads of aromatics, thus revealing a competition in cell catabolism. A step-drop in loading of aromatics resulted in an immediate increase of RE(W) with variations in the TBR, while the new steady-state value in the BF took 6-7 h to achieve. The TBR consistently showed a greater performance than BF in removing toluene and xylenes. Increasing the loading rate of aromatics resulted in a gradual decrease of their REs. The degradation rates of acetone and n-butyl acetate were also lower at higher OL(AROM), revealing a competition in the cell catabolism. The results obtained are consistent with the proposed hypothesis of greater toxic effects under low water content, i.e., in the biofilter, caused by aromatic hydrocarbons in the presence of polar ketones and esters, which may improve the hydrocarbon partitioning into the aqueous phase.

  10. System Engineering Program Applicability for the High Temperature Gas-Cooled Reactor (HTGR) Component Test Capability (CTC)

    SciTech Connect

    Jeffrey Bryan

    2009-06-01

    This white paper identifies where the technical management and systems engineering processes and activities to be used in establishing the High Temperature Gas-cooled Reactor (HTGR) Component Test Capability (CTC) should be addressed and presents specific considerations for these activities under each CTC alternative

  11. Biofiltration of paint solvent mixtures in two reactor types: overloading by polar components.

    PubMed

    Paca, Jan; Halecky, Martin; Misiaczek, Ondrej; Kozliak, Evguenii I; Jones, Kim

    2012-01-01

    Steady-state performances of a trickle bed reactor (TBR) and a biofilter (BF) in loading experiments with increasing inlet concentrations of polar solvents, acetone, methyl ethyl ketone, methyl isobutyl ketone and n-butyl acetate, were investigated, along with the system's dynamic responses. Throughout the entire experimentation time, a constant loading rate of aromatic components of 4 g(c)·m(-3)·h(-1) was maintained to observe the interactions between the polar substrates and aromatic hydrocarbons. Under low combined substrate loadings, the BF outperformed TBR not only in the removal of aromatic hydrocarbons but also in the removal of polar substrates. However, increasing the loading rate of polar components above the threshold value of 31-36 g(c)·m(-3)·h(-1) resulted in a steep and significant drop in the removal efficiencies of both polar (except for butyl acetate) and hydrophobic components, which was more pronounced in the BF; so the relative TBR/BF efficiency became reversed under such overloading conditions. A step-drop of the overall OL(POLAR) (combined loading by polar air pollutants) from overloading values to 7 g(c)·m(-3)·h(-1) resulted in an increase of all pollutant removal efficiencies, although in TBR the recovery was preceded by lag periods lasting between 5 min (methyl ethyl ketone) to 3.7 h (acetone). The occurrence of lag periods in the TBR recovery was, in part, due to the saturation of mineral medium with water-soluble polar solvents, particularly, acetone. The observed bioreactor behavior was consistent with the biological steps being rate-limiting.

  12. Progress in the Reliable Inspection of Cast Stainless Steel Reactor Piping Components

    SciTech Connect

    Doctor, Steven R.; Anderson, Michael T.; Diaz, Aaron A.; Cumblidge, Stephen E.

    2005-12-31

    Studies conducted at the Pacific N¬orthwest National Laboratory (PNNL) in Richland, Washington, have focused on assessing the effectiveness and reliability of novel NDE approaches for the inspection of coarse-grained, cast stainless steel reactor components. The primary objective of this work is to provide information to the United States Nuclear Regulatory Commission (US NRC) on the utility, effec¬tiveness and reliability of ultrasonic testing (UT) and eddy current testing (ET) inspection techniques as related to the inservice ultrasonic inspec¬tion of primary piping components in pressurized water reactors (PWRs). This paper describes progress, recent developments and results from assessments of three different NDE approaches including ultrasonic phased array inspection techniques, eddy current testing for surface-breaking flaws, and a low-frequency ultrasonic inspection methodology coupled with a synthetic aperture focusing technique (SAFT). Westinghouse Owner’s Group (WOG) cast stainless steel pipe segments with thermal and mechanical fatigue cracks, PNNL samples containing thermal fatigue cracks and several blank spool pieces were used for assessing the inspection methods. Eddy current studies were conducted on the inner diameter (ID) surface of piping specimens while the ultrasonic inspection methods were applied from the outer diameter (OD) surface of the specimens. The eddy current technique employed a Zetec MIZ-27SI Eddy Current instrument and a Zetec Z0000857-1 cross point spot probe with an operating frequency of 250 kHz. In order to reduce noise effects, degaussing of a subset of the samples resulted in noticeable improvements. The phased array approach was implemented using an R/D Tech Tomoscan III system operating at 1 MHz, providing composite volumetric images of the samples. The low-frequency ultrasonic method employs a zone-focused, multi-incident angle inspection protocol (operating at 250-500 kHz) coupled with SAFT for improved signal

  13. 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.

  14. Neutron Dosimetry on the Full-Core First Generation VVER-440 Aimed at Reactor Support Structure Load Evaluation

    NASA Astrophysics Data System (ADS)

    Borodkin, P.; Borodkin, G.; Khrennikov, N.; Konheiser, J.; Noack, K.

    2009-08-01

    Reactor support structures (RSS), especially the ferritic steel wall of the water tank, of first-generation VVER-440 are non-restorable reactor equipment, and their lifetime may restrict plant-life. All operated Russian first generation VVER-440 have a reduced core with dummy assemblies except Unit 4 of Novovoronezh nuclear power plant (NPP). In comparison with other reactors, the full-core loading scheme of this reactor provides the highest neutron fluence on the reactor pressure vessel (RPV) and RSS accumulated over design service-life and its prolongation. The radiation load parameters on the RPV and RSS that have resulted from this core loading scheme should be evaluated by means of precise calculations and validated by ex-vessel neutron dosimetry to provide the reliable assessment of embrittlement parameters of these reactor components. The results of different types of calculations and their comparison with measured data have been analyzed in this paper. The calculational analysis of RSS fluence rate variation in dependence on the core loading scheme, including the standard and low leakage core as well as the introduction of dummy assemblies, is presented in this paper.

  15. 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.

  16. 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.

  17. Nondestructive characterization of structural ceramic components

    SciTech Connect

    Ellingson, W.A.; Steckenrider, J.S.; Sivers, E.A.; Ling, J.R.

    1994-06-01

    Advanced structural ceramic components under development for heat-engine applications include both monolithic and continuous fiber composites (CFC). Nondestructive characterization (NDC) methods being developed differ for each material system. For monolithic materials, characterization during processing steps is important. For many CFC, only post process characterization is possible. Many different NDC systems have been designed and built A 3D x-ray micro computed tomographic (3DXCT) imaging system has been shown to be able to map density variations to better than 3% in pressure slip cast Si{sub 3}N{sub 4} monolithic materials. In addition, 3DXCT coupled to image processing has been shown to be able to map through-thickness fiber orientations in 2D lay-ups of 0{degrees}/45{degrees}, 0{degrees}/75{degrees}, 0{degrees}/90{degrees}, in SiC/SiC CVI CFC. Fourier optics based laser scatter systems have been shown to be able to detect surface and subsurface defects (as well as microstructural variations) in monolithic Si{sub 3}N{sub 4} bearing balls. Infrared methods using photothermal excitation have been shown to be able to detect and measure thermal diffusivity differences on SiC/SiC 2D laminated CFC which have been subjected to different thermal treatments including thermal shock and oxidizing environments. These NDC methods and their applications help provide information to allow reliable usage of ceramics in advanced heat engine applications.

  18. Software for Testing Electroactive Structural Components

    NASA Technical Reports Server (NTRS)

    Moses, Robert W.; Fox, Robert L.; Dimery, Archie D.; Bryant, Robert G.; Shams, Qamar

    2003-01-01

    A computer program generates a graphical user interface that, in combination with its other features, facilitates the acquisition and preprocessing of experimental data on the strain response, hysteresis, and power consumption of a multilayer composite-material structural component containing one or more built-in sensor(s) and/or actuator(s) based on piezoelectric materials. This program runs in conjunction with Lab-VIEW software in a computer-controlled instrumentation system. For a test, a specimen is instrumented with appliedvoltage and current sensors and with strain gauges. Once the computational connection to the test setup has been made via the LabVIEW software, this program causes the test instrumentation to step through specified configurations. If the user is satisfied with the test results as displayed by the software, the user activates an icon on a front-panel display, causing the raw current, voltage, and strain data to be digitized and saved. The data are also put into a spreadsheet and can be plotted on a graph. Graphical displays are saved in an image file for future reference. The program also computes and displays the power and the phase angle between voltage and current.

  19. Power reactivity decrement components of a homogeneous UPu10Zr-fueled 900-MW(thermal) liquid metal reactor

    SciTech Connect

    Meneghetti, D.; Kucera, D.A.

    1989-01-01

    Linear and Doppler feedback components of the power reactivity decrement (PRD) for a 900-MW(thermal) homogeneous UPu10Zr-fueled sodium-cooled reactor have been calculated. (The PRD is the negative of the reactivity required to bring the reactor from a zero-power hot-critical condition to a given power level.) The components are further separated into power-dependent and power-to-flow-dependent parts. These delineations enhance understanding of the contributions of the components to the feedback process. The delineation also enables the PRDs for other values of coolant flows to be estimated. The linear and Doppler components of the PRD are obtained using the EBRPOCO code, which calculates detailed axially delineated contributions of the components for every subassembly of a loading configuration. Separation of the components into power and power-to-flow parts is made by calculations of the components, assuming infinite thermal conductivities to obtain the power-to-flow values. Subtractions of these from the corresponding PRD quantities give the power-dependent parts. The values of the various feedback components are compared with corresponding quantities reported for an analogous U10Zr-fueled case.

  20. Structure and creep of Russian reactor steels with a BCC structure

    NASA Astrophysics Data System (ADS)

    Sagaradze, V. V.; Kochetkova, T. N.; Kataeva, N. V.; Kozlov, K. A.; Zavalishin, V. A.; Vil'danova, N. F.; Ageev, V. S.; Leont'eva-Smirnova, M. V.; Nikitina, A. A.

    2017-05-01

    The structural phase transformations have been revealed and the characteristics of the creep and long-term strength at 650, 670, and 700°C and 60-140 MPa have been determined in six Russian reactor steels with a bcc structure after quenching and high-temperature tempering. Creep tests were carried out using specially designed longitudinal and transverse microsamples, which were fabricated from the shells of the fuel elements used in the BN-600 fast neutron reactor. It has been found that the creep rate of the reactor bcc steels is determined by the stability of the lath martensitic and ferritic structures in relation to the diffusion processes of recovery and recrystallization. The highest-temperature oxide-free steel contains the maximum amount of the refractory elements and carbides. The steel strengthened by the thermally stable Y-Ti nanooxides has a record high-temperature strength. The creep rate at 700°C and 100 MPa in the samples of this steel is lower by an order of magnitude and the time to fracture is 100 times greater than that in the oxide-free reactor steels.

  1. Thermalhydraulic calculation for boiling water reactor and its natural circulation component

    SciTech Connect

    Trianti, Nuri Nurjanah,; Su’ud, Zaki; Arif, Idam; Permana, Sidik

    2015-09-30

    Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density and inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.

  2. Final Report: Safety of Plasma Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

    SciTech Connect

    Bourham, Mohamed A.; Gilligan, John G.

    1999-08-14

    Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.

  3. 76 FR 68514 - Request for a License To Export Reactor Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-11-04

    ... reactor 12 Perform seismic China. LLC, August 18, 2011, October control rod system testing necessary 6, 2011, XR174, 11005963. and associated for qualification equipment. of AP1000 (design) nuclear...

  4. Development of a Versatile Ultrasonic Internal Pipe/Vessel Component Monitor for In-Service Inspection of Nuclear Reactor Components

    SciTech Connect

    Searfass, Clifford T.; Malinowski, Owen M.; Van Velsor, Jason K.

    2015-03-22

    The stated goal of this work was to develop a versatile system which could accurately measure vessel and valve internal vibrations and cavitation formation under in-service conditions in nuclear power plants, ultrasonically. The developed technology will benefit the nuclear power generation industry by allowing plant operators to monitor valve and vessel internals during operation. This will help reduce planned outages and plant component failures. During the course of this work, Structural Integrity Associates, Inc. gathered information from industry experts that target vibration amplitudes to be detected should be in the range of 0.001-in to 0.005-in (0.025-mm to 0.127-mm) and target vibration frequency ranges which should be detected were found to be between 0-Hz and 300-Hz. During the performed work, an ultrasonic measuring system was developed which utilized ultrasonic pulse-echo time-of-flight measurements to measure vibration frequency and amplitude. The developed system has been shown to be able to measure vibration amplitudes as low as 0.0008-in (0.020-mm) with vibration frequencies in the range of 17-Hz to 1000-Hz. Therefore, the developed system was able to meet the industry needs for vibration measurement. The developed ultrasonic system was also to be able to measure cavitation formation by monitoring the received ultrasonic time- and frequency-domain signals. This work also demonstrated the survivability of commercially available probes at temperatures up to 300-F for several weeks.

  5. Improved performance of parallel surface/packed-bed discharge reactor for indoor VOCs decomposition: optimization of the reactor structure

    NASA Astrophysics Data System (ADS)

    Jiang, Nan; Hui, Chun-Xue; Li, Jie; Lu, Na; Shang, Ke-Feng; Wu, Yan; Mizuno, Akira

    2015-10-01

    The purpose of this paper is to develop a high-efficiency air-cleaning system for volatile organic compounds (VOCs) existing in the workshop of a chemical factory. A novel parallel surface/packed-bed discharge (PSPBD) reactor, which utilized a combination of surface discharge (SD) plasma with packed-bed discharge (PBD) plasma, was designed and employed for VOCs removal in a closed vessel. In order to optimize the structure of the PSPBD reactor, the discharge characteristic, benzene removal efficiency, and energy yield were compared for different discharge lengths, quartz tube diameters, shapes of external high-voltage electrode, packed-bed discharge gaps, and packing pellet sizes, respectively. In the circulation test, 52.8% of benzene was removed and the energy yield achieved 0.79 mg kJ-1 after a 210 min discharge treatment in the PSPBD reactor, which was 10.3% and 0.18 mg kJ-1 higher, respectively, than in the SD reactor, 21.8% and 0.34 mg kJ-1 higher, respectively, than in the PBD reactor at 53 J l-1. The improved performance in benzene removal and energy yield can be attributed to the plasma chemistry effect of the sequential processing in the PSPBD reactor. The VOCs mineralization and organic intermediates generated during discharge treatment were followed by CO x selectivity and FT-IR analyses. The experimental results indicate that the PSPBD plasma process is an effective and energy-efficient approach for VOCs removal in an indoor environment.

  6. Modelling of advanced structural materials for GEN IV reactors

    NASA Astrophysics Data System (ADS)

    Samaras, M.; Hoffelner, W.; Victoria, M.

    2007-09-01

    The choice of suitable materials and the assessment of long-term materials damage are key issues that need to be addressed for the safe and reliable performance of nuclear power plants. Operating conditions such as high temperatures, irradiation and a corrosive environment degrade materials properties, posing the risk of very expensive or even catastrophic plant damage. Materials scientists are faced with the scientific challenge to determine the long-term damage evolution of materials under service exposure in advanced plants. A higher confidence in life-time assessments of these materials requires an understanding of the related physical phenomena on a range of scales from the microscopic level of single defect damage effects all the way up to macroscopic effects. To overcome lengthy and expensive trial-and-error experiments, the multiscale modelling of materials behaviour is a promising tool, bringing new insights into the fundamental understanding of basic mechanisms. This paper presents the multiscale modelling methodology which is taking root internationally to address the issues of advanced structural materials for Gen IV reactors.

  7. Spectral structure of electron antineutrinos from nuclear reactors.

    PubMed

    Dwyer, D A; Langford, T J

    2015-01-09

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principles calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructures in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of these substructures can elucidate the nuclear processes occurring within reactors. These substructures can be a systematic issue for measurements utilizing the detailed spectral shape.

  8. Preliminary study of degradation from neutron effects of core-structural materials of Thai Research Reactor TRR-1/M1

    NASA Astrophysics Data System (ADS)

    Ampornrat, P.; Boonsuwan, P.; Sangkaew, S.; Angwongtrakool, T.

    2017-06-01

    Thai research reactor went first critical in 1962. The reactor was converted in 1977 from an MTR-type with high-enriched uranium fuel to a TRIGA-MARK III type using low-enriched uranium fuel, called TRR-1/M1. Since the TRR-1/M1 has been operated for almost 40 years, degradation of reactor structural materials is expected. In this preliminary study, the potential degradation from neutron effects of core-structural materials, e.g., fuel clad (SS304) and core components (Al6061) were studied. Assessment included calculation of neutron energy, flux and fluence in the reactor core to evaluate displacement rate (dpa) and irradiation effects on the material properties. Results showed maximum displacement rates on SS304 was 5.24×10-8 per cm3·sec and on Al6061 was 1.14×10-8 per cm3·sec. The corresponding maximum displacement levels were ∼17 dpa for SS304, and ∼4 dpa for Al6061. At these levels of displacement, it is possible for the materials to result in tensile strength increasing and ductility reduction. Further inspection on the core-structural materials needs to be conducted to validate the assessment results from this study.

  9. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    NASA Astrophysics Data System (ADS)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  10. Changing concepts of geologic structure and the problem of siting nuclear reactors: Examples from Washington State

    NASA Astrophysics Data System (ADS)

    Tabor, R. W.

    1986-09-01

    The conflict between regulation and healthy evolution of geological science has contributed to the difficulties of siting nuclear reactors. On the Columbia Plateau in Washington, but for conservative design of the Hanford reactor facility, the recognition of the little-understood Olympic-Wallowa lineament as a major, possibly still active structural alinement might have jeopardized the acceptability of the site for nuclear reactors. On the Olympic Peninsula, evolving concepts of compressive structures and their possible recent activity and the current recognition of a subducting Juan de Fuca plate and its potential for generating great earthquakes—both concepts little-considered during initial site selection—may delay final acceptance of the Satsop site. Conflicts of this sort are inevitable but can be accommodated if they are anticipated in the reactor-licensing process. More important, society should be increasing its store of geologic knowledge now, during the current recess in nuclear reactor siting.

  11. Polymer matrix nanocomposites for automotive structural components.

    PubMed

    Naskar, Amit K; Keum, Jong K; Boeman, Raymond G

    2016-12-06

    Over the past several decades, the automotive industry has expended significant effort to develop lightweight parts from new easy-to-process polymeric nanocomposites. These materials have been particularly attractive because they can increase fuel efficiency and reduce greenhouse gas emissions. However, attempts to reinforce soft matrices by nanoscale reinforcing agents at commercially deployable scales have been only sporadically successful to date. This situation is due primarily to the lack of fundamental understanding of how multiscale interfacial interactions and the resultant structures affect the properties of polymer nanocomposites. In this Perspective, we critically evaluate the state of the art in the field and propose a possible path that may help to overcome these barriers. Only once we achieve a deeper understanding of the structure-properties relationship of polymer matrix nanocomposites will we be able to develop novel structural nanocomposites with enhanced mechanical properties for automotive applications.

  12. Polymer matrix nanocomposites for automotive structural components

    NASA Astrophysics Data System (ADS)

    Naskar, Amit K.; Keum, Jong K.; Boeman, Raymond G.

    2016-12-01

    Over the past several decades, the automotive industry has expended significant effort to develop lightweight parts from new easy-to-process polymeric nanocomposites. These materials have been particularly attractive because they can increase fuel efficiency and reduce greenhouse gas emissions. However, attempts to reinforce soft matrices by nanoscale reinforcing agents at commercially deployable scales have been only sporadically successful to date. This situation is due primarily to the lack of fundamental understanding of how multiscale interfacial interactions and the resultant structures affect the properties of polymer nanocomposites. In this Perspective, we critically evaluate the state of the art in the field and propose a possible path that may help to overcome these barriers. Only once we achieve a deeper understanding of the structure-properties relationship of polymer matrix nanocomposites will we be able to develop novel structural nanocomposites with enhanced mechanical properties for automotive applications.

  13. Light Water Reactor Sustainability Program Advanced Seismic Soil Structure Modeling

    SciTech Connect

    Bolisetti, Chandrakanth; Coleman, Justin Leigh

    2015-06-01

    Risk calculations should focus on providing best estimate results, and associated insights, for evaluation and decision-making. Specifically, seismic probabilistic risk assessments (SPRAs) are intended to provide best estimates of the various combinations of structural and equipment failures that can lead to a seismic induced core damage event. However, in some instances the current SPRA approach has large uncertainties, and potentially masks other important events (for instance, it was not the seismic motions that caused the Fukushima core melt events, but the tsunami ingress into the facility). SPRA’s are performed by convolving the seismic hazard (this is the estimate of all likely damaging earthquakes at the site of interest) with the seismic fragility (the conditional probability of failure of a structure, system, or component given the occurrence of earthquake ground motion). In this calculation, there are three main pieces to seismic risk quantification, 1) seismic hazard and nuclear power plants (NPPs) response to the hazard, 2) fragility or capacity of structures, systems and components (SSC), and 3) systems analysis. Two areas where NLSSI effects may be important in SPRA calculations are, 1) when calculating in-structure response at the area of interest, and 2) calculation of seismic fragilities (current fragility calculations assume a lognormal distribution for probability of failure of components). Some important effects when using NLSSI in the SPRA calculation process include, 1) gapping and sliding, 2) inclined seismic waves coupled with gapping and sliding of foundations atop soil, 3) inclined seismic waves coupled with gapping and sliding of deeply embedded structures, 4) soil dilatancy, 5) soil liquefaction, 6) surface waves, 7) buoyancy, 8) concrete cracking and 9) seismic isolation The focus of the research task presented here-in is on implementation of NLSSI into the SPRA calculation process when calculating in-structure response at the area

  14. Components of microtubular structures in Saccharomyces cerevisiae.

    PubMed Central

    Pillus, L; Solomon, F

    1986-01-01

    Most studies of cytoskeletal organelles have concentrated on molecular analyses of abundant and biochemically accessible structures. In many of the classical cases, however, the nature of the system chosen has precluded a concurrent genetic analysis. The mitotic spindle of the yeast Saccharomyces cerevisiae is one example of an organelle that can be studied by both classical and molecular genetics. We show here that this microtubule structure also can be examined biochemically. The spindle can be isolated by selective extractions of yeast cells by using adaptations of methods successfully applied to animal cells. In this way, microtubule-associated proteins of the yeast spindle are identified. Images PMID:3517870

  15. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    SciTech Connect

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  16. Structured Observation Component. Secondary Teacher Education Program.

    ERIC Educational Resources Information Center

    Berger, Michael L.; Keen, Phyllis A.

    A format is presented for use of student teachers in structuring their classroom observation techniques. Fifteen classroom and school activities are listed with a comprehensive questionnaire accompanying each. These questionnaires guide the student on what behaviors to observe and suggest objective and subjective responses to these behaviors to be…

  17. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    SciTech Connect

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

  18. Analysis of truss, beam, frame, and membrane components. [composite structures

    NASA Technical Reports Server (NTRS)

    Knoell, A. C.; Robinson, E. Y.

    1975-01-01

    Truss components are considered, taking into account composite truss structures, truss analysis, column members, and truss joints. Beam components are discussed, giving attention to composite beams, laminated beams, and sandwich beams. Composite frame components and composite membrane components are examined. A description is given of examples of flat membrane components and examples of curved membrane elements. It is pointed out that composite structural design and analysis is a highly interactive, iterative procedure which does not lend itself readily to characterization by design or analysis function only.-

  19. Polymer matrix nanocomposites for automotive structural components

    DOE PAGES

    Naskar, Amit K.; Keum, Jong K.; Boeman, Raymond G.

    2016-12-06

    Over the past several decades, the automotive industry has expended significant effort to develop lightweight parts from new easy-to-process polymeric nanocomposites. These materials have been particularly attractive because they can increase fuel efficiency and reduce greenhouse gas emissions. However, attempts to reinforce soft matrices by nanoscale reinforcing agents at commercially deployable scales have been only sporadically successful to date. This situation is due primarily to the lack of fundamental understanding of how multiscale interfacial interactions and the resultant structures affect the properties of polymer nanocomposites. In this paper, we critically evaluate the state of the art in the field andmore » propose a possible path that may help to overcome these barriers. Finally, only once we achieve a deeper understanding of the structure–properties relationship of polymer matrix nanocomposites will we be able to develop novel structural nanocomposites with enhanced mechanical properties for automotive applications.« less

  20. Polymer matrix nanocomposites for automotive structural components

    SciTech Connect

    Naskar, Amit K.; Keum, Jong K.; Boeman, Raymond G.

    2016-12-06

    Over the past several decades, the automotive industry has expended significant effort to develop lightweight parts from new easy-to-process polymeric nanocomposites. These materials have been particularly attractive because they can increase fuel efficiency and reduce greenhouse gas emissions. However, attempts to reinforce soft matrices by nanoscale reinforcing agents at commercially deployable scales have been only sporadically successful to date. This situation is due primarily to the lack of fundamental understanding of how multiscale interfacial interactions and the resultant structures affect the properties of polymer nanocomposites. In this paper, we critically evaluate the state of the art in the field and propose a possible path that may help to overcome these barriers. Finally, only once we achieve a deeper understanding of the structure–properties relationship of polymer matrix nanocomposites will we be able to develop novel structural nanocomposites with enhanced mechanical properties for automotive applications.

  1. Assessment of current structural design methodology for high-temperature reactors based on failure tests

    SciTech Connect

    Corum, J.M.; Sartory, W.K.

    1985-01-01

    A mature design methodology, consisting of inelastic analysis methods, provided in Department of Energy guidelines, and failure criteria, contained in ASME Code Case N-47, exists in the United States for high-temperature reactor components. The objective of this paper is to assess the adequacy of this overall methodology by comparing predicted inelastic deformations and lifetimes with observed results from structural failure tests and from an actual service failure. Comparisons are presented for three types of structural situations: (1) nozzle-to-spherical shell specimens, where stresses at structural discontinuities lead to cracking, (2) welded structures, where metallurgical discontinuities play a key role in failures, and (3) thermal shock loadings of cylinders and pipes, where thermal discontinuities can lead to failure. The comparison between predicted and measured inelastic responses are generally reasonalbly good; quantities are sometimes overpredicted somewhat, and, sometimes underpredicted. However, even seemingly small discrepancies can have a significant effect on structural life, and lifetimes are not always as closely predicted. For a few cases, the lifetimes are substantially overpredicted, which raises questions regarding the adequacy of existing design margins.

  2. Structural evaluation of the Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads

    SciTech Connect

    Fischer, L.E.; Chou, C.K.; Lo, T.; Schwartz, M.W.

    1988-06-01

    A structural evaluation of Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads under the normal and hypothetical accident conditions of 10 CFR 71 was performed. Component performance criteria for the Shippingport package and the corresponding structural acceptance criteria for these components were developed based on a review of the package geometry, the planned transport environment, and the external radiation standards and dispersal limits of 10 CFR 71. The evaluation was performed using structural analysis methods. A demonstration combining simplified model tests and nonlinear finite element analyses was made to substantiate the structural analysis methods used to evaluate the Shippingport package. The package was analyzed and the results indicate that the package meets external radiation standards and release limits of 10 CFR 71. 13 refs., 50 figs., 19 tabs.

  3. Flow Components in a NaK Test Loop Designed to Simulate Conditions in a Nuclear Surface Power Reactor

    SciTech Connect

    Polzin, Kurt A.; Godfroy, Thomas J.

    2008-01-21

    A test loop using NaK as the working fluid is presently in use to study material compatibility effects on various components that comprise a possible nuclear reactor design for use on the lunar surface. A DC electromagnetic (EM) pump has been designed and implemented as a means of actively controlling the NaK flow rate through the system and an EM flow sensor is employed to monitor the developed flow rate. These components allow for the matching of the flow rate conditions in test loops with those that would be found in a full-scale surface-power reactor. The design and operating characteristics of the EM pump and flow sensor are presented. In the EM pump, current is applied to a set of electrodes to produce a Lorentz body force in the fluid. A measurement of the induced voltage (back-EMF) in the flow sensor provides the means of monitoring flow rate. Both components are compact, employing high magnetic field strength neodymium magnets thermally coupled to a water-cooled housing. A vacuum gap limits the heat transferred from the high temperature NaK tube to the magnets and a magnetically-permeable material completes the magnetic circuit. The pump is designed to produce a pressure rise of 34.5 kPa, and the flow sensor's predicted output is roughly 20 mV at the loop's nominal flow rate of 0.114 m{sup 3}/hr.

  4. Flow Components in a NaK Test Loop Designed to Simulate Conditions in a Nuclear Surface Power Reactor

    NASA Technical Reports Server (NTRS)

    Polzin, Kurt A.; Godfroy, Thomas J.

    2008-01-01

    A test loop using NaK as the working fluid is presently in use to study material compatibility effects on various components that comprise a possible nuclear reactor design for use on the lunar surface. A DC electromagnetic (EM) pump has been designed and implemented as a means of actively controlling the NaK flow rate through the system and an EM flow sensor is employed to monitor the developed flow rate. These components allow for the matching of the flow rate conditions in test loops with those that would be found in a full-scale surface-power reactor. The design and operating characteristics of the EM pump and flow sensor are presented. In the EM pump, current is applied to a set of electrodes to produce a Lorentz body force in the fluid. A measurement of the induced voltage (back-EMF) in the flow sensor provides the means of monitoring flow rate. Both components are compact, employing high magnetic field strength neodymium magnets thermally coupled to a water-cooled housing. A vacuum gap limits the heat transferred from the high temperature NaK tube to the magnets and a magnetically-permeable material completes the magnetic circuit. The pump is designed to produce a pressure rise of 5 psi, and the flow sensor's predicted output is roughly 20 mV at the loop's nominal flow rate of 0.5 GPM.

  5. Structure of two-component clusters

    NASA Astrophysics Data System (ADS)

    Clarke, A. S.; Kapral, R.; Patey, G. N.

    1994-08-01

    Phase separation in binary liquid Lennard-Jones clusters is investigated employing computer simulation methods. Clusters ranging in size from 50 to 240 particles are considered with special emphasis on systems with equal numbers of A and B particles. Cluster morphology is systematically explored by varying the ratios, α=ɛAB/ɛAA, β=ɛBB/ɛAA, Γ=σAB/σAA, and Δ=σBB/σAA, where σ and ɛ are the Lennard-Jones size and energy parameters. A detailed α, β ``phase diagram'' is presented for the case Γ=Δ=1. Stable phase separated clusters are shown to fall into two general classes: elongated clusters of cylindrical or dumbbell shape, the ends of which are A-rich and B-rich phases, and spherical coated clusters consisting of a core of one species coated by the other. More quantitative structural information is given in the form of interfacial density profiles. We also propose two theoretical models for phase separation in binary clusters. One is a simple macroscopiclike droplet approach and the other is a mean field lattice model. Both simple models capture many of the important physical features observed in the computer simulations. Together they provide insight into the nature of phase separation in small systems.

  6. EVALUATION OF THE DURABILITY OF THE STRUCTURAL CONCRETE OF REACTOR BUILDINGS AT SRS

    SciTech Connect

    Duncan, A.; Reigel, M.

    2011-02-28

    The Department of Energy (DOE) intends to close 100-150 facilities in the DOE complex using an in situ decommissioning (ISD) strategy that calls for grouting the below-grade interior volume of the structure and leaving the above-grade interior open or demolishing it and disposing of it in the slit trenches in E Area. These closures are expected to persist and remain stable for centuries, but there are neither facility-specific monitoring approaches nor studies on the rate of deterioration of the materials used in the original construction or on the ISD components added during closure (caps, sloped roofs, etc). This report will focus on the evaluation of the actual aging/degradation of the materials of construction used in the ISD structures at Savannah River Site (SRS) above grade, specifically P & R reactor buildings. Concrete blocks (six 2 to 5 ton blocks) removed from the outer wall of the P Reactor Building were turned over to SRNL as the first source for concrete cores. Larger cores were received as a result of grouting activities in P and R reactor facilities. The cores were sectioned and evaluated using microscopy, x-ray diffraction (XRD), ion chromatography (IC) and thermal analysis. Scanning electron microscopy shows that the aggregate and cement phases present in the concrete are consistent with the mix design and no degradation mechanisms are evident at the aggregate-cement interfaces. Samples of the cores were digested and analyzed for chloride ingress as well as sulfate attack. The concentrations of chloride and sulfate ions did not exceed the limits of the mix design and there is no indication of any degradation due to these mechanisms. Thermal analysis on samples taken along the longitudinal axis of the cores show that there is a 1 inch carbonation layer (i.e., no portlandite) present in the interior wall of the reactor building and a negligible carbonation layer in the exterior wall. A mixed layer of carbonate and portlandite extends deeper into the

  7. Evaluation of induced radioactivity in structural material of Toshiba Training Reactor 'TTR1'.

    PubMed

    Uematsu, Mikio; Kurosawa, Masahiko; Haruguchi, Yoshiko

    2005-01-01

    A decommissioning programme for the Toshiba Training Reactor (TTR1), a swimming pool type reactor used for reactor physics experiments and material irradiation, was started in August 2001. As a part of the programme, induced radioactivity in structural material was evaluated using neutron flux data obtained with the three-dimensional Sn code TORT. Induced activity was calculated with the isotope generation code ORIGEN-79 using activation cross section data created from multi-group library based on JENDL-3. The obtained results for radioactivities such as 60Co, 65Zn, 54Mn and 152Eu were compared with measured ones, and the present calculational method was confirmed to have enough accuracy.

  8. Experimental and analytical studies of high heat flux components for fusion experimental reactor

    NASA Astrophysics Data System (ADS)

    Araki, Masanori

    1993-03-01

    In this report, the experimental and analytical results concerning the development of plasma facing components of ITER are described. With respect to developing high heat removal structures for the divertor plates, an externally-finned swirl tube was developed based on the results of critical heat flux (CHF) experiments on various tube structures. As the result, the burnout heat flux, which also indicates incident CHF, of 41 (+/-) 1 MW/sq m was achieved in the externally-finned swirl tube. The applicability of existing CHF correlations based on uniform heating conditions was evaluated by comparing the CHF experimental data with the smooth and the externally-finned tubes under one-sided heating condition. As the results, experimentally determined CHF data for straight tube show good agreement, for the externally-finned tube, no existing correlations are available for prediction of the CHF. With respect to the evaluation of the bonds between carbon-based material and heat sink metal, results of brazing tests were compared with the analytical results by three dimensional model with temperature-dependent thermal and mechanical properties. Analytical results showed that residual stresses from brazing can be estimated by the analytical three directional stress values instead of the equivalent stress value applied. In the analytical study on the separatrix sweeping for effectively reducing surface heat fluxes on the divertor plate, thermal response of the divertor plate was analyzed under ITER relevant heat flux conditions and has been tested. As the result, it has been demonstrated that application of the sweeping technique is very effective for improvement in the power handling capability of the divertor plate and that the divertor mock-up has withstood a large number of additional cyclic heat loads.

  9. Structural materials for ITER in-vessel component design

    NASA Astrophysics Data System (ADS)

    Kalinin, G.; Gauster, W.; Matera, R.; Tavassoli, A.-A. F.; Rowcliffe, A.; Fabritsiev, S.; Kawamura, H.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m 2 in the basic performance phase (BPP)) within a temperature range from 20 to 300°C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350°C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. Estimates of radiation damage at the locations for re-welding show that the dose will not exceed 0.05 dpa (with He generation of 1 appm) for the manifold and 0.01 dpa (with He generation 0.1 appm) for the back plate for the BPP of ITER operation. Existing experimental data show that these levels will not result in property changes for SS; however, neutron irradiation and He generation promote crack formation in the heat affected zone during welding. Cu based alloys, DS-Cu (Glidcop A125) and PHCu CuCrZr bronze) are proposed as a structural materials for high heat flux

  10. Experimental Study of Thermal Crisis in Connection with Tokamak Reactor High Heat Flux Components

    SciTech Connect

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-12-31

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology.

  11. Optimization of composite wood structural components : processing and design choices

    Treesearch

    Theodore L. Laufenberg

    1985-01-01

    Decreasing size and quality of the world's forest resources are responsible for interest in producing composite wood structural components. Process and design optimization methods are offered in this paper. Processing concepts for wood composite structural products are reviewed to illustrate manufacturing boundaries and areas of high potential. Structural...

  12. Acoustic Emission and Guided Ultrasonic Waves for Detection and Continuous Monitoring of Cracks in Light Water Reactor Components

    SciTech Connect

    Meyer, Ryan M.; Coble, Jamie B.; Ramuhalli, Pradeep; Watson, Bruce E.; Cumblidge, Stephen E.; Doctor, Steven R.; Bond, Leonard J.

    2012-06-28

    Acoustic emission (AE) and guided ultrasonic waves (GUW) are considered for continuous monitoring and detection of cracks in Light Water Reactor (LWR) components. In this effort, both techniques are applied to the detection and monitoring of fatigue crack growth in a full scale pipe component. AE results indicated crack initiation and rapid growth in the pipe, and significant GUW responses were observed in response to the growth of the fatigue crack. After initiation, the crack growth was detectable with AE for approximately 20,000 cycles. Signals associated with initiation and rapid growth where distinguished based on total rate of activity and differences observed in the centroid frequency of hits. An intermediate stage between initiation and rapid growth was associated with significant energy emissions, though few hits. GUW exhibit a nearly monotonic trend with crack length with an exception of measurements obtained at 41 mm and 46 mm.

  13. Prioritization of reactor control components susceptible to fire damage as a consequence of aging

    SciTech Connect

    Lowry, W.; Vigil, R.; Nowlen, S.

    1994-01-01

    The Fire Vulnerability of Aged Electrical Components Test Program is to identify and assess issues of plant aging that could lead to an increase in nuclear power plant risk because of fires. Historical component data and prior analyses are used to prioritize a list of components with respect to aging and fire vulnerability and the consequences of their failure on plant safety systems. The component list emphasizes safety system control components, but excludes cables, large equipment, and devices encompassed in the Equipment Qualification (EQ) program. The test program selected components identified in a utility survey and developed test and fire conditions necessary to maximize the effectiveness of the test program. Fire damage considerations were limited to purely thermal effects.

  14. Nuclear reactor melt-retention structure to mitigate direct containment heating

    DOEpatents

    Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

    1991-01-01

    A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

  15. 10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

  16. 10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

  17. 10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

  18. Closeup view of Flume Bridge #4 showing structural components. Looking ...

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

    Close-up view of Flume Bridge #4 showing structural components. Looking northeast - Childs-Irving Hydroelectric Project, Childs System, Flume Bridge No. 4, Forest Service Road 708/502, Camp Verde, Yavapai County, AZ

  19. Impact of inocula and operating conditions on the microbial community structure of two anammox reactors.

    PubMed

    Costa, Maria Cristina Monteiro S; Carvalho, Luciana; Leal, Cintia Dutra; Dias, Marcela França; Martins, Karoline L; Garcia, Guilherme Brugger; Mancuelo, Isabella Daldegan; Hipólito, Thais; Conell, Erika F Abreu Mac; Okada, Dagoberto; Etchebehere, Claudia; Chernicharo, Carlos Augusto L; Araujo, Juliana Calabria

    2014-08-01

    The microbial community structure of the biomass selected in two distinctly inoculated anaerobic oxidation of ammonium (anammox) reactors was investigated and compared with the help of data obtained from 454-pyrosequencing analyses. The anammox reactors were operated for 550 days and seeded with different sludges: sediment from a constructed wetland (reactor I) and biomass from an aerated lagoon part of the oil-refinery wastewater treatment plant (reactor II). The anammox diversity in the inocula was evaluated by 16S rRNA gene-cloning analysis. The diversity of anammox bacteria was greater in the sludge from the oil-refinery (three of the five known genera of anammox were detected) than in the wetland sludge, in which only Candidatus Brocadia was observed. Pyrosequencing analysis demonstrated that the community enriched in both reactors had differing compositions despite the nearly similar operational conditions applied. The dominant phyla detected in both reactors were Proteobacteria, Chloroflexi, Planctomycetes, and Acidobacteria. The phylum Bacteroidetes, which is frequently observed in anammox reactors, was not detected. However, Acidobacteria and GN04 phyla were observed for the first time, suggesting their importance for this process. Our results suggest that, under similar operational conditions, anammox populations (Ca. Brocadia sinica and Ca. Brocadia sp. 40) were selected in both reactors despite the differences between the two initial inocula. Taken together, these results indicated that the type of inoculum and the culture conditions are key determinants of the general microbial composition of the biomass produced in the reactors. Operational conditions alone might play an important role in anammox selection.

  20. Apparatus, components and operating methods for circulating fluidized bed transport gasifiers and reactors

    DOEpatents

    Vimalchand, Pannalal; Liu, Guohai; Peng, Wan Wang

    2015-02-24

    The improvements proposed in this invention provide a reliable apparatus and method to gasify low rank coals in a class of pressurized circulating fluidized bed reactors termed "transport gasifier." The embodiments overcome a number of operability and reliability problems with existing gasifiers. The systems and methods address issues related to distribution of gasification agent without the use of internals, management of heat release to avoid any agglomeration and clinker formation, specific design of bends to withstand the highly erosive environment due to high solid particles circulation rates, design of a standpipe cyclone to withstand high temperature gasification environment, compact design of seal-leg that can handle high mass solids flux, design of nozzles that eliminate plugging, uniform aeration of large diameter Standpipe, oxidant injection at the cyclone exits to effectively modulate gasifier exit temperature and reduction in overall height of the gasifier with a modified non-mechanical valve.

  1. Engine structures analysis software: Component Specific Modeling (COSMO)

    NASA Astrophysics Data System (ADS)

    McKnight, R. L.; Maffeo, R. J.; Schwartz, S.

    1994-08-01

    A component specific modeling software program has been developed for propulsion systems. This expert program is capable of formulating the component geometry as finite element meshes for structural analysis which, in the future, can be spun off as NURB geometry for manufacturing. COSMO currently has geometry recipes for combustors, turbine blades, vanes, and disks. Component geometry recipes for nozzles, inlets, frames, shafts, and ducts are being added. COSMO uses component recipes that work through neutral files with the Technology Benefit Estimator (T/BEST) program which provides the necessary base parameters and loadings. This report contains the users manual for combustors, turbine blades, vanes, and disks.

  2. Engine Structures Analysis Software: Component Specific Modeling (COSMO)

    NASA Technical Reports Server (NTRS)

    Mcknight, R. L.; Maffeo, R. J.; Schwartz, S.

    1994-01-01

    A component specific modeling software program has been developed for propulsion systems. This expert program is capable of formulating the component geometry as finite element meshes for structural analysis which, in the future, can be spun off as NURB geometry for manufacturing. COSMO currently has geometry recipes for combustors, turbine blades, vanes, and disks. Component geometry recipes for nozzles, inlets, frames, shafts, and ducts are being added. COSMO uses component recipes that work through neutral files with the Technology Benefit Estimator (T/BEST) program which provides the necessary base parameters and loadings. This report contains the users manual for combustors, turbine blades, vanes, and disks.

  3. Simplified design procedures for fiber composite structural components/joints

    NASA Technical Reports Server (NTRS)

    Murthy, P. L. N.; Chamis, Christos C.

    1990-01-01

    Simplified step-by-step design procedures are summarized, which are suitable for the preliminary design of composite structural components such as panels (laminates) and composite built-up structures (box beams). Similar procedures are also summarized for the preliminary design of composite bolted and adhesively bonded joints. The summary is presented in terms of sample design cases complemented with typical results. Guidelines are provided which can be used in the design selection process of composite structural components/joints. Also, procedures to account for cyclic loads, hygrothermal effects and lamination residual stresses are included.

  4. Weight minimization of structural components for launch in space shuttle

    NASA Technical Reports Server (NTRS)

    Patnaik, Surya N.; Gendy, Atef S.; Hopkins, Dale A.; Berke, Laszlo

    1994-01-01

    Minimizing the weight of structural components of the space station launched into orbit in a space shuttle can save cost, reduce the number of space shuttle missions, and facilitate on-orbit fabrication. Traditional manual design of such components, although feasible, cannot represent a minimum weight condition. At NASA Lewis Research Center, a design capability called CometBoards (Comparative Evaluation Test Bed of Optimization and Analysis Routines for the Design of Structures) has been developed especially for the design optimization of such flight components. Two components of the space station - a spacer structure and a support system - illustrate the capability of CometBoards. These components are designed for loads and behavior constraints that arise from a variety of flight accelerations and maneuvers. The optimization process using CometBoards reduced the weights of the components by one third from those obtained with traditional manual design. This paper presents a brief overview of the design code CometBoards and a description of the space station components, their design environments, behavior limitations, and attributes of their optimum designs.

  5. Structural components of nuclear integrity with gene regulatory potential.

    PubMed

    Fenelon, Kelli D; Hopyan, Sevan

    2017-10-01

    The nucleus is a mechanosensitive and load-bearing structure. Structural components of the nucleus interact to maintain nuclear integrity and have become subjects of exciting research that is relevant to cell and developmental biology. Here we outline the boundaries of what is known about key architectural elements within the nucleus and highlight their potential structural and transcriptional regulatory functions. Copyright © 2017. Published by Elsevier Ltd.

  6. Insights for aging management of light water reactor components: Metal containments. Volume 5

    SciTech Connect

    Shah, V.N.; Sinha, U.P.; Smith, S.K.

    1994-03-01

    This report evaluates the available technical information and field experience related to management of aging damage to light water reactor metal containments. A generic aging management approach is suggested for the effective and comprehensive aging management of metal containments to ensure their safe operation. The major concern is corrosion of the embedded portion of the containment vessel and detection of this damage. The electromagnetic acoustic transducer and half-cell potential measurement are potential techniques to detect corrosion damage in the embedded portion of the containment vessel. Other corrosion-related concerns include inspection of corrosion damage on the inaccessible side of BWR Mark I and Mark II containment vessels and corrosion of the BWR Mark I torus and emergency core cooling system piping that penetrates the torus, and transgranular stress corrosion cracking of the penetration bellows. Fatigue-related concerns include reduction in the fatigue life (a) of a vessel caused by roughness of the corroded vessel surface and (b) of bellows because of any physical damage. Maintenance of surface coatings and sealant at the metal-concrete interface is the best protection against corrosion of the vessel.

  7. Micronutrient component changes in the biogas slurry treated by a pilot solar-heated anaerobic reactor

    NASA Astrophysics Data System (ADS)

    Yang, Z. Y.; Xu, Y. B.; Li, P. F.; Wang, Y. J.; Sun, J.; Zhang, Y. P.

    2017-06-01

    A solar-heated anaerobic reactor system was applied to decompose livestock wastewater, in which cattle manure and chopped straw were mixed (CODCr 15,000∼25,000 mg·l-1), the commercial microorganisms were added to ambient acidification (about 32°C) and the acclimated sludge was inoculated. Then, the experiments were carried out on wastewater anaerobic degradation and biogas production at 40∼42°C, as fed every 10 days till stable running. The results showed that NH3-N and PO4 3- of the biogas slurry were 441 mg·l-1 and 65.0 mg·l-1 on the 35th day, respectively. The concentration of K was up to 350 mg·l-1 in the biogas slurry, rather higher than that of Mg and Fe, which indicated that the available K could contribute more in the agricultural irrigation. Total amino acids were up to 23.7 mg·l-1 after anaerobic digestion, in which Lys, Thr, Ala and Arg were prominent in the biogas slurry. These amino acids could be beneficial to seed soaking, feed adding and apply as foliar fertilizer. The major volatile organic compounds were detected in the biogas slurry, including toluene, m-cresol (up to 0.036% in the process of ambient acidification) and triethylsilane, which could be reduced to scarcely influence on agricultural application after anaerobic digestion.

  8. Continuous electrochemical treatment of simulated industrial textile wastewater from industrial components in a tubular reactor.

    PubMed

    Körbahti, Bahadir K; Tanyolaç, Abdurrahman

    2009-10-30

    The continuous electrochemical treatment of industrial textile wastewater in a tubular reactor was investigated. The synthetic wastewater was based on the real process information of pretreatment and dyeing stages of the industrial mercerized and non-mercerized cotton and viscon production. The effects of residence time on chemical oxygen demand (COD), color and turbidity removals and pH change were studied under response surface optimized conditions of 30 degrees C, 25 g/L electrolyte concentration and 3505 mg/L COD feed concentration with 123.97 mA/cm(2) current density. Increasing residence time resulted in steady profiles of COD and color removals with higher treatment performances. The best column performance was realized at 3h of residence time as 53.5% and 99.3% for COD and color removals, respectively, at the expense of 193.1 kWh/kg COD with a mass transfer coefficient of 9.47 x 10(-6) m/s.

  9. Fluid flow structure around the mixer in a reactor with mechanical mixing

    SciTech Connect

    Lecheva, A.; Zheleva, I.

    2015-10-28

    Fluid flow structure around the mixer in a cylindrical reactor with mechanical mixing is studied and numerical results are presented in this article. The model area is complex because of the presence of convex corners of the mixer in the fluid flow. Proper boundary conditions for the vorticity calculated on the base of the stream function values near solid boundaries of the examined area are presented. The boundary value problem of motion of swirling incompressible viscous fluid in a vertical tank reactor with a mixer is solved numerically. The calculations are made by a computer code, written in MATLAB. The complex structure of the flow around the mixing disk is described and commented.

  10. Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

    SciTech Connect

    Ballinger, Ronald G.; Wang, Chun Yun; Kadak, Andrew; Todreas, Neil; Mirick, Bradley; Demetri, Eli; Koronowski, Martin

    2004-08-30

    The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R&D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the

  11. Oscillating liquid flow ICF Reactor

    SciTech Connect

    Petzoldt, R.W.

    1990-12-14

    Oscillating liquid flow in a falling molten salt inertial confinement fusion reactor is predicted to rapidly clear driver beam paths of residual liquid droplets. Oscillating flow will also provide adequate neutron and x-ray protection for the reactor structure with a short (2-m) fall distance permitting an 8 Hz repetition rate. A reactor chamber configuration is presented with specific features to clear the entire heavy-ion beam path of splashed molten salt. The structural components, including the structure between beam ports, are shielded. 3 refs., 12 figs.

  12. Structural analysis of ultra-high speed aircraft structural components

    NASA Technical Reports Server (NTRS)

    Lenzen, K. H.; Siegel, W. H.

    1977-01-01

    The buckling characteristics of a hypersonic beaded skin panel were investigated under pure compression with boundary conditions similar to those found in a wing mounted condition. The primary phases of analysis reported include: (1) experimental testing of the panel to failure; (2) finite element structural analysis of the beaded panel with the computer program NASTRAN; and (3) summary of the semiclassical buckling equations for the beaded panel under purely compressive loads. A comparison of each of the analysis methods is also included.

  13. Structural analysis methods development for turbine hot section components

    NASA Technical Reports Server (NTRS)

    Thompson, R. L.

    1989-01-01

    The structural analysis technologies and activities of the NASA Lewis Research Center's gas turbine engine HOT Section Technoloogy (HOST) program are summarized. The technologies synergistically developed and validated include: time-varying thermal/mechanical load models; component-specific automated geometric modeling and solution strategy capabilities; advanced inelastic analysis methods; inelastic constitutive models; high-temperature experimental techniques and experiments; and nonlinear structural analysis codes. Features of the program that incorporate the new technologies and their application to hot section component analysis and design are described. Improved and, in some cases, first-time 3-D nonlinear structural analyses of hot section components of isotropic and anisotropic nickel-base superalloys are presented.

  14. Structural Analysis Methods Development for Turbine Hot Section Components

    NASA Technical Reports Server (NTRS)

    Thompson, Robert L.

    1988-01-01

    The structural analysis technologies and activities of the NASA Lewis Research Center's gas turbine engine Hot Section Technology (HOST) program are summarized. The technologies synergistically developed and validated include: time-varying thermal/mechanical load models; component-specific automated geometric modeling and solution strategy capabilities; advanced inelastic analysis methods; inelastic constitutive models; high-temperature experimental techniques and experiments; and nonlinear structural analysis codes. Features of the program that incorporate the new technologies and their application to hot section component analysis and design are described. Improved and, in some cases, first-time 3-D nonlinear structural analyses of hot section components of isotropic and anisotropic nickel-base superalloys are presented.

  15. Channel structures in aerobic biofilms of fixed-film reactors treating contaminated groundwater.

    PubMed Central

    Massol-Deyá, A A; Whallon, J; Hickey, R F; Tiedje, J M

    1995-01-01

    Scanning electron microscopy, confocal scanning laser microscopy, and fatty acid methyl ester profiles were used to study the development, organization, and structure of aerobic multispecies biofilm communities in granular activated-carbon (GAC) fluidized-bed reactors treating petroleum-contaminated groundwaters. The sequential development of biofilm structure was studied in a laboratory reactor fed toluene-amended groundwater and colonized by the indigenous aquifer populations. During the early stages of colonization, microcolonies were observed primarily in crevices and other regions sheltered from hydraulic shear forces. Eventually, these microcolonies grew over the entire surface of the GAC. This growth led to the development of discrete discontinuous multilayer biofilm structures. Cell-free channel-like structures of variable sizes were observed to interconnect the surface film with the deep inner layers. These interconnections appeared to increase the biological surface area per unit volume ratio, which may facilitate transport of substrates into and waste products out of deep regions of the biofilm at rates greater than possible by diffusion alone. These architectural features were also observed in biofilms from four field-scale GAC reactors that were in commercial operation treating petroleum-contaminated groundwaters. These shared features suggest that formation of cell-free channel structures and their maintenance may be a general microbial strategy to deal with the problem of limiting diffusive transport in thick biofilms typical of fluidized-bed reactors. PMID:7574613

  16. A structured approach to evaluating aging of the advanced test reactor

    SciTech Connect

    Dwight, J.E.

    1990-01-01

    An aging evaluation program has been developed for the United States Department of Energy's Advanced Test Reactor to support the current goal of operation through the year 2014 and beyond. The Aging Evaluation and Life Extension Program (AELEX) employs a three-phased approach. In Phases 1 and 2, now complete, components were identified, categorized and prioritized. Critical components were selected and aging mechanisms for the critical components identified. An initial evaluation of the critical components was performed and extended life operation for the plant appears to be both technically and economically feasible. Detailed evaluations of the critical components are now in progress in the early stages of Phase 3. Some results are available. Evaluations of many non-critical components and refinements to the program based on probabilistic risk assessment results will follow in later stages of Phase 3. 6 refs., 2 figs., 5 tabs.

  17. Reactor building

    SciTech Connect

    Hista, J. C.

    1984-09-18

    Reactor building comprising a vessel shaft anchored in a slab which is peripherally locked. This reactor building comprises a confinement enclosure within which are positioned internal structures constituted by an internal structure floor, a vessel shaft, a slab being positioned between the general floor and the internal structure floor, the vesse

  18. 77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-20

    ... Aging Lessons Learned (GALL) Report for the aging management of stainless steel structures and.... Background The NRC issues LR-ISGs to communicate insights and lessons learned and to address emergent...

  19. Jacking mechanism for upper internals structure of a liquid metal nuclear reactor

    DOEpatents

    Gillett, James E.; Wineman, Arthur L.

    1984-01-01

    A jacking mechanism for raising the upper internals structure of a liquid metal nuclear reactor which jacking mechanism uses a system of gears and drive shafts to transmit force from a single motor to four mechanically synchronized ball jacks to raise and lower support columns which support the upper internals structure. The support columns have a pin structure which rides up and down in a slot in a housing fixed to the reactor head. The pin has two locking plates which can be rotated around the pin to bring bolt holes through the locking plates into alignment with a set of bolt holes in the housing, there being a set of such housing bolt holes corresponding to both a raised and a lowered position of the support column. When the locking plate is so aligned, a surface of the locking plate mates with a surface in the housing such that the support column is then supported by the locking plate and not by the ball jacks. Since the locking plates are to be installed and bolted to the housing during periods of reactor operation, the ball jacks need not be sized to react the large forces which occur or potentially could occur on the upper internals structure of the reactor during operation. The locking plates react these loads. The ball jacks, used only during refueling, can be smaller, which enable conventionally available equipment to fulfill the precision requirements for the task within available space.

  20. Designing a Component-Based Architecture for the Modeling and Simulation of Nuclear Fuels and Reactors

    SciTech Connect

    Billings, Jay Jay; Elwasif, Wael R; Hively, Lee M; Bernholdt, David E; Hetrick III, John M; Bohn, Tim T

    2009-01-01

    Concerns over the environment and energy security have recently prompted renewed interest in the U.S. in nuclear energy. Recognizing this, the U.S. Dept. of Energy has launched an initiative to revamp and modernize the role that modeling and simulation plays in the development and operation of nuclear facilities. This Nuclear Energy Advanced Modeling and Simulation (NEAMS) program represents a major investment in the development of new software, with one or more large multi-scale multi-physics capabilities in each of four technical areas associated with the nuclear fuel cycle, as well as additional supporting developments. In conjunction with this, we are designing a software architecture, computational environment, and component framework to integrate the NEAMS technical capabilities and make them more accessible to users. In this report of work very much in progress, we lay out the 'problem' we are addressing, describe the model-driven system design approach we are using, and compare them with several large-scale technical software initiatives from the past. We discuss how component technology may be uniquely positioned to address the software integration challenges of the NEAMS program, outline the capabilities planned for the NEAMS computational environment and framework, and describe some initial prototyping activities.

  1. Acoustic emission and guided ultrasonic waves for detection and continuous monitoring of cracks in light water reactor components

    SciTech Connect

    Meyer, R. M.; Coble, J.; Ramuhalli, P.; Watson, B.; Cumblidge, S. E.; Doctor, S. R.; Bond, L. J.

    2012-07-01

    Acoustic emission (AE) and guided ultrasonic waves (GUW) are considered for continuous monitoring and detection of cracks in Light Water Reactor (LWR) components. In this effort, both techniques are applied to the detection and monitoring of fatigue crack growth in a full scale pipe component. AE results indicated crack initiation and rapid growth in the pipe, and significant GUW responses were observed in response to the growth of the fatigue crack. After initiation, the crack growth was detectable with AE for approximately 20,000 cycles. Signals associated with initiation and rapid growth were distinguished based on total rate of activity and differences observed in the centroid frequency of hits. An intermediate stage between initiation and rapid growth was associated with significant energy emissions, though few hits. GUW exhibit a nearly monotonic trend with crack length with an exception of measurements obtained at crack lengths of 41 mm and 46 mm. Coupling variability and shadowing by the electro-discharge machining (EDM) starter notch set the lower limit of detectability. (authors)

  2. The component content of active particles in a plasma-chemical reactor based on volume barrier discharge

    NASA Astrophysics Data System (ADS)

    Soloshenko, I. A.; Tsiolko, V. V.; Pogulay, S. S.; Terent'yeva, A. G.; Bazhenov, V. Yu; Shchedrin, A. I.; Ryabtsev, A. V.; Kuzmichev, A. I.

    2007-02-01

    In this paper the results of theoretical and experimental studies of the component content of active particles formed in a plasma-chemical reactor composed of a multiple-cell generator of active particles, based on volume barrier discharge, and a working chamber are presented. For calculation of the content of uncharged plasma components an approach is proposed which is based on averaging of the power introduced over the entire volume. Advantages of such an approach lie in an absence of fitting parameters, such as the dimensions of microdischarges, their surface density and rate of breakdown. The calculation and the experiment were accomplished with the use of dry air (20% relative humidity) as the plasma generating medium. Concentrations of O3, HNO3, HNO2, N2 O5 and NO3 were measured experimentally in the discharge volume and working chamber for the residence time of particles on a discharge of 0.3 s and more and discharge specific power of 1.5 W cm-3. It has been determined that the best agreement between the calculation and the experiment occurs at calculated gas medium temperatures in the discharge plasma of about 400-425 K, which correspond to the experimentally measured rotational temperature of nitrogen. In most cases the calculated concentrations of O3, HNO3, HNO2, N2O5 and NO3 for the barrier discharge and the working chamber are in fairly good agreement with the respective measured values.

  3. Synchronization of chemical noise-sustained structures in asymmetrically coupled differential-flow reactors.

    PubMed

    Izús, Gonzalo G; Sánchez, Alejandro D

    2013-12-01

    The differential-flow-induced chemical instability is investigated in the context of two coupled reactors with cubic autocatalytic kinetics (the Gray-Scott model). Previous results for master-slave arrangement [Izús, Deza, and Sánchez, J. Chem. Phys. 132, 234112 (2010)] are extended in this study to include bidirectional coupling between reactions. Numerical simulations in the convectively unstable regime show that synchronized noise-sustained structures are developed in both reactors due to the selective amplification of noise. A theoretical analysis shows that the nature of the synchronization and the stability of the synchronized manifold are related with the properties of the critical modes.

  4. Vertical distribution of structural components in corn stover

    USDA-ARS?s Scientific Manuscript database

    In much of the United States, corn (Zea mays L.) stover is the most abundant and widespread agricultural residue. Because of this abundance, stover has been targeted as feedstock for second generation fuel production and other bio-products. Ethanol yield is linked to sugars, while structural compone...

  5. Component modes damping assignment methodology for articulated, multiflexible body structures

    NASA Technical Reports Server (NTRS)

    Lee, Allan Y.

    1993-01-01

    To simulate the dynamical motion of articulated, multiflexible body structures, one can use multibody simulation packages such as DISCOS. To this end, one must supply appropriate reduced-order models for all of the flexible components involved. The component modes projection and assembly model reduction (COMPARE) methodology is one way to construct these reduced-order component models, which when reassembled capture important system input-to-output mapping of the full-order model at multiple system configurations of interest. In conjunction, we must also supply component damping matrices which when reassembled generate a system damping matrix that has certain desirable properties. The problem of determining the damping factors of components' modes to achieve a given system damping matrix is addressed here. To this end, we must establish from first principles a matrix-algebraic relation between the system's modal damping matrix and the components' modal damping matrices. An unconstrained/constrained optimization problem can then be formulated to determine the component modes' damping factors that best satisfy that matrix-algebraic relation. The effectiveness of the developed methodology, called ModeDamp, has been successfully demonstrated on a high-order, finite element model of the Galileo spacecraft.

  6. High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

    SciTech Connect

    Not Available

    1983-01-01

    The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented.

  7. Influence of operation conditions on structure and properties of 12% Cr steels as candidate structural materials for fusion reactor

    NASA Astrophysics Data System (ADS)

    Ioltukhovsky, A. G.; Leontyeva-Smirnova, M. V.; Kazennov, Y. I.; Medvedeva, E. A.; Tselishchev, A. V.; Shamardin, V. K.; Povstyanko, A. V.; Ostrovsky, S. E.; Dvoryashin, A. M.; Porollo, S. I.; Vorobyev, A. N.; Khabarov, V. S.

    1998-10-01

    The Russian experience in the development and operation of the nuclear core components in fast reactors with a sodium coolant demonstrates that 12% Cr steels may be successfully used at temperatures of 270-650°C and at high neutron damage dose (up to 100 dpa and above). The priority of the temperature but not of dose of the irradiation is noted for the steels at 270-350°C. In addition, the following may take place: a sharp decrease in the ductility of material, a change in the mechanism of fracture with which the ductile-brittle-transition-temperature (DBTT) shift is associated. With an increase in irradiation temperature to 350-500°C and the irradiation dose (up to 100 dpa) chromium steels are observed to strengthen; their ductility increased monotonously, and embrittlement does not show up. With the irradiation temperature increased above 500°C (up to 650-690°C), the material becomes plastic and some of its strength properties are decreased. The high level of the irradiation resistance of 12% Cr steels is a result of their structure and phase transformations. The properties of the welded joints of 12% Cr steels under the conditions of the neutron irradiation are slightly inferior to the properties of the base metal.

  8. TA-2 Water Boiler Reactor (Phase I) decommissioning: Resolving ground water interference during outdoor component removal and soil decontamination

    SciTech Connect

    Elder, J.C.

    1987-01-01

    Extensive ground water was encountered in excavations as structures, pipe, and soil contaminated with Cs-137 were removed during the outdoor phase (Phase I) of the TA-2 Water Boiler Reactor decommissioning. To avoid releasing contamination to a nearby surface stream, contaminated soil and water pumped or scooped into secondary pits above the level of ground water interference. These pits served as holding tanks/natural filters because high soil pH in the area enhanced attachment of Cs-137 to soil particles. Water seeped out of the secondary pits, leaving contaminated soil behind to easily scraped from the surfaces of the pit. Stabilization and revegetation of the area, complemented by a long-term environmental surveillance program, has resulted in successful closure, although total removal of all contamination was not possible. 3 refs., 5 figs.

  9. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, J.D.; Cassulo, J.C.; Pedersen, D.R.; Baker, L. Jr.

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and can be discharged from the reactor core. The invention provides a porous bed of sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  10. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker Jr., Louis

    1986-07-01

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  11. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker, Jr., Louis

    1986-01-01

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  12. Carbon source--a strong determinant of microbial community structure and performance of an anaerobic reactor.

    PubMed

    Kundu, K; Bergmann, I; Hahnke, S; Klocke, M; Sharma, S; Sreekrishnan, T R

    2013-12-01

    Industrial effluents differ in their organic composition thereby providing different carbon sources to the microbial communities involved in its treatment. This study aimed to investigate the correlation of microbial community structure with wastewater composition and reactor's performance. Self-immobilized granules were developed in simulated wastewater based on different carbon sources (glucose, sugarcane molasses, and milk) in three hybrid anaerobic reactors operated at 37°C. To study archaeal community structure, a polyphasic approach was used with both qualitative and quantitative analysis. While PCR-denaturing gradient gel electrophoresis of 16S rRNA gene did not reveal major shifts in diversity of archaea with change in substrate, quantification of different groups of methanogens and total bacteria by real-time PCR showed variations in relative abundances with the dominance of Methanosaetaceae and Methanobacteriales. These data were supported by differences in the ratio of total counts of archaea and bacteria analyzed by catalyzed reporter deposition - fluorescence in situ hybridization. During hydraulic and organic shocks, the molasses-based reactor showed the best performance followed by the milk- and the glucose-based reactor. The study indicates that carbon source shapes the microbial community structure more in terms of relative abundance with distinct metabolic capacities rather than its diversity itself.

  13. Evaluation of low activation vanadium alloys for structural material in a fusion reactor

    SciTech Connect

    Loomis, B.A.; Hull, A.B.; Smith, D.L.

    1989-10-23

    The V-7.2Cr-14.5Ti, V-9.2Cr-4.9Ti, V-9.9Cr-9.2Ti, V-13.5Cr-5.2Ti, V-4.1Cr-4.3Ti, Vanstar-7, V-4.6Ti, V-17.7Ti, and V-3.1Ti-(0.5-1.0)Si alloys were evaluated for use as structural material in a fusion reactor. The alloys were evaluated on the basis of their yield strength, swelling resistance, resistance to hydrogen and irradiation embrittlement, and compatibility with a lithium reactor coolant. On the basis of these evaluations, the V-7.2Cr-14.5Ti, V-9.2Cr-4.9Ti, V-9.9Cr-9.2Ti, V-13.5Cr-5.2Ti, Vanstar-7, and V-3.1Ti-(0.5-1.0)Si alloys are considered unacceptable for structural material in a fusion reactor, whereas the V-4.1Cr-4.3Ti, V-4.6Ti, and V-17.7Ti alloys are recommended for more intensive evaluation. The V-7Cr-5Ti alloy may have the optimum combination of strength, DBTT, swelling rate, and lithium dissolution rate for a structural material in a fusion reactor. 4 refs., 6 figs., 4 tabs.

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

    SciTech Connect

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

    1992-08-01

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

  15. Improved Joining of Metal Components to Composite Structures

    NASA Technical Reports Server (NTRS)

    Semmes, Edmund

    2009-01-01

    Systems requirements for complex spacecraft drive design requirements that lead to structures, components, and/or enclosures of a multi-material and multifunctional design. The varying physical properties of aluminum, tungsten, Invar, or other high-grade aerospace metals when utilized in conjunction with lightweight composites multiply system level solutions. These multi-material designs are largely dependent upon effective joining techAn improved method of joining metal components to matrix/fiber composite material structures has been invented. The method is particularly applicable to equipping such thin-wall polymer-matrix composite (PMC) structures as tanks with flanges, ceramic matrix composite (CMC) liners for high heat engine nozzles, and other metallic-to-composite attachments. The method is oriented toward new architectures and distributing mechanical loads as widely as possible in the vicinities of attachment locations to prevent excessive concentrations of stresses that could give rise to delaminations, debonds, leaks, and other failures. The method in its most basic form can be summarized as follows: A metal component is to be joined to a designated attachment area on a composite-material structure. In preparation for joining, the metal component is fabricated to include multiple studs projecting from the aforementioned face. Also in preparation for joining, holes just wide enough to accept the studs are molded into, drilled, or otherwise formed in the corresponding locations in the designated attachment area of the uncured ("wet') composite structure. The metal component is brought together with the uncured composite structure so that the studs become firmly seated in the holes, thereby causing the composite material to become intertwined with the metal component in the joining area. Alternately, it is proposed to utilize other mechanical attachment schemes whereby the uncured composite and metallic parts are joined with "z-direction" fasteners. The

  16. Crystal structure of the RNA component of bacterial ribonuclease P.

    PubMed

    Torres-Larios, Alfredo; Swinger, Kerren K; Krasilnikov, Andrey S; Pan, Tao; Mondragón, Alfonso

    2005-09-22

    Transfer RNA (tRNA) is produced as a precursor molecule that needs to be processed at its 3' and 5' ends. Ribonuclease P is the sole endonuclease responsible for processing the 5' end of tRNA by cleaving the precursor and leading to tRNA maturation. It was one of the first catalytic RNA molecules identified and consists of a single RNA component in all organisms and only one protein component in bacteria. It is a true multi-turnover ribozyme and one of only two ribozymes (the other being the ribosome) that are conserved in all kingdoms of life. Here we show the crystal structure at 3.85 A resolution of the RNA component of Thermotoga maritima ribonuclease P. The entire RNA catalytic component is revealed, as well as the arrangement of the two structural domains. The structure shows the general architecture of the RNA molecule, the inter- and intra-domain interactions, the location of the universally conserved regions, the regions involved in pre-tRNA recognition and the location of the active site. A model with bound tRNA is in agreement with all existing data and suggests the general basis for RNA-RNA recognition by this ribozyme.

  17. Structure of nitrilotriacetate monooxygenase component B from Mycobacterium thermoresistibile

    PubMed Central

    Zhang, Y.; Edwards, T. E.; Begley, D. W.; Abramov, A.; Thompkins, K. B.; Ferrell, M.; Guo, W. J.; Phan, I.; Olsen, C.; Napuli, A.; Sankaran, B.; Stacy, R.; Van Voorhis, W. C.; Stewart, L. J.; Myler, P. J.

    2011-01-01

    Mycobacterium tuberculosis belongs to a large family of soil bacteria which can degrade a remarkably broad range of organic compounds and utilize them as carbon, nitrogen and energy sources. It has been proposed that a variety of mycobacteria can subsist on alternative carbon sources during latency within an infected human host, with the help of enzymes such as nitrilotriacetate monooxygenase (NTA-Mo). NTA-Mo is a member of a class of enzymes which consist of two components: A and B. While component A has monooxygenase activity and is responsible for the oxidation of the substrate, component B consumes cofactor to generate reduced flavin mononucleotide, which is required for component A activity. NTA-MoB from M. thermoresistibile, a rare but infectious close relative of M. tuberculosis which can thrive at elevated temperatures, has been expressed, purified and crystallized. The 1.6 Å resolution crystal structure of component B of NTA-Mo presented here is one of the first crystal structures determined from the organism M. thermo­resistibile. The NTA-MoB crystal structure reveals a homodimer with the characteristic split-barrel motif typical of flavin reductases. Surprisingly, NTA-MoB from M. thermoresistibile contains a C-terminal tail that is highly conserved among myco­bacterial orthologs and resides in the active site of the other protomer. Based on the structure, the C-terminal tail may modulate NTA-MoB activity in mycobacteria by blocking the binding of flavins and NADH. PMID:21904057

  18. Mapping Flow Localization Processes in Deformation of Irradiated Reactor Structural Alloys

    SciTech Connect

    Farrell, K.

    2002-07-18

    Metals that can sustain plastic deformation homogeneously throughout their bulk tend to be tough and malleable. Often, however, if a metal has been hardened it will no longer deform uniformly. Instead, the deformation occurs in narrow bands on a microscopic scale wherein stresses and strains become concentrated in localized zones. This strain localization degrades the mechanical properties of the metal by causing premature plastic instability failure or by inducing the formation of cracks. Irradiation with neutrons hardens a metal and makes it more prone to deformation by strain localization. Although this has been known since the earliest days of radiation damage studies, a full measure of the connection between neutron irradiation hardening and strain localization is wanting, particularly in commercial alloys used in the construction of nuclear reactors. Therefore, the goal of this project is to systematically map the extent of involvement of strain localization processes in plastic deformation of three reactor alloys that have been neutron irradiated. The deformation processes are to be identified and related to changes in the tensile properties of the alloys as functions of neutron fluence (dose) and degree of plastic strain. The intent is to define the role of strain localization in radiation embrittlement phenomena. The three test materials are a tempered bainitic A533B steel, representing reactor pressure vessel steel, an annealed 316 stainless steel and annealed Zircaloy-4 representing reactor internal components.

  19. Multi-Scale Sizing of Lightweight Multifunctional Spacecraft Structural Components

    NASA Technical Reports Server (NTRS)

    Bednarcyk, Brett A.

    2005-01-01

    This document is the final report for the project entitled, "Multi-Scale Sizing of Lightweight Multifunctional Spacecraft Structural Components," funded under the NRA entitled "Cross-Enterprise Technology Development Program" issued by the NASA Office of Space Science in 2000. The project was funded in 2001, and spanned a four year period from March, 2001 to February, 2005. Through enhancements to and synthesis of unique, state of the art structural mechanics and micromechanics analysis software, a new multi-scale tool has been developed that enables design, analysis, and sizing of advance lightweight composite and smart materials and structures from the full vehicle, to the stiffened structure, to the micro (fiber and matrix) scales. The new software tool has broad, cross-cutting value to current and future NASA missions that will rely on advanced composite and smart materials and structures.

  20. Probabilistic structural analysis methods for select space propulsion system components

    NASA Technical Reports Server (NTRS)

    Millwater, H. R.; Cruse, T. A.

    1989-01-01

    The Probabilistic Structural Analysis Methods (PSAM) project developed at the Southwest Research Institute integrates state-of-the-art structural analysis techniques with probability theory for the design and analysis of complex large-scale engineering structures. An advanced efficient software system (NESSUS) capable of performing complex probabilistic analysis has been developed. NESSUS contains a number of software components to perform probabilistic analysis of structures. These components include: an expert system, a probabilistic finite element code, a probabilistic boundary element code and a fast probability integrator. The NESSUS software system is shown. An expert system is included to capture and utilize PSAM knowledge and experience. NESSUS/EXPERT is an interactive menu-driven expert system that provides information to assist in the use of the probabilistic finite element code NESSUS/FEM and the fast probability integrator (FPI). The expert system menu structure is summarized. The NESSUS system contains a state-of-the-art nonlinear probabilistic finite element code, NESSUS/FEM, to determine the structural response and sensitivities. A broad range of analysis capabilities and an extensive element library is present.

  1. Novel electrode structure in a DBD reactor applied to the degradation of phenol in aqueous solution

    NASA Astrophysics Data System (ADS)

    Mercado-Cabrera, Antonio; Peña-Eguiluz, Rosendo; López-Callejas, Régulo; Jaramillo-Sierra, Bethsabet; Valencia-Alvarado, Raúl; Rodríguez-Méndez, Benjamín; Muñoz-Castro, Arturo E.

    2017-07-01

    Phenol degradation experimental results are presented in a similar wastewater aqueous solution using a non-thermal plasma reactor in a coaxial dielectric barrier discharge. The novelty of the work is that one of the electrodes of the reactor has the shape of a hollow screw which shows an enhanced efficiency compared with a traditional smooth structure. The experimentation was carried out with gas mixtures of 90% Ar-10% O2, 80% Ar-20% O2 and 0% Ar-100% O2. After one hour of treatment the removal efficiency was 76%, 92%, and 97%, respectively, assessed with a gas chromatographic mass spectrometry technique. For both reactors used, the ozone concentration was measured. The screw electrode required less energy, for all gas mixtures, than the smooth electrode, to maintain the same ozone concentration. On the other hand, it was also observed that in both electrodes the electrical conductivity of the solution changed slightly from ˜0.0115 S m-1 up to ˜0.0430 S m-1 after one hour of treatment. The advantages of using the hollow screw electrode structure compared with the smooth electrode were: (1) lower typical power consumption, (2) the generation of a uniform plasma throughout the reactor benefiting the phenol degradation, (3) a relatively lower temperature of the aqueous solution during the process, and (4) the plasma generation length is larger.

  2. Evaluation of Halpern's "structural component" for improving critical thinking.

    PubMed

    Nieto, Ana Ma; Saiz, Carlos

    2008-05-01

    Halpern (1998) proposed a four-component model for promoting the transfer of critical thinking. One of them, the "structural component," focuses on how to organize teaching so that critical thinking skills can be generalized. Here, we assess the efficiency of that type of organization. Thus, one group of university students received instruction following the suggestions specified in that component and their performance was compared with that of other university students who received instruction in the same skills but using a different procedure, and with that of a control group. In comparison with the control group, the performance of both instructed groups was better after training. However, no significant differences were observed between either instruction group; both forms of instruction afforded very similar results.

  3. Fluid-Structure Interaction for Coolant Flow in Research-type Nuclear Reactors

    SciTech Connect

    Curtis, Franklin G; Ekici, Kivanc; Freels, James D

    2011-01-01

    The High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), is scheduled to undergo a conversion of the fuel used and this proposed change requires an extensive analysis of the flow through the reactor core. The core consists of 540 very thin and long fuel plates through which the coolant (water) flows at a very high rate. Therefore, the design and the flow conditions make the plates prone to dynamic and static deflections, which may result in flow blockage and structural failure which in turn may cause core damage. To investigate the coolant flow between fuel plates and associated structural deflections, the Fluid-Structure Interaction (FSI) module in COMSOL will be used. Flow induced flutter and static deflections will be examined. To verify the FSI module, a test case of a cylinder in crossflow, with vortex induced vibrations was performed and validated.

  4. Structural failure analysis of reactor vessels due to molten core debris

    SciTech Connect

    Pfeiffer, P.A.

    1993-08-01

    Maintaining structural integrity of the reactor vessel during a postulated core melt accident is an important safety consideration in the design of the vessel. This paper addresses the failure predictions of the vessel due to thermal and pressure loadings from the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on a wet or dry cavity and pressurization of the vessel based on operating pressure or atmospheric (pipe break). The analyses considered both short term (minutes) and long term (days) failure modes. Short term failure modes include creep at elevated temperatures and plastic instabilities of the structure. Long term failure modes are caused by creep rupture that lead to plastic instability of the structure. The analyses predict the reactor vessel will remain intact after the core melt has deposited on the lower vessel head.

  5. 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.

  6. Structural and wetting properties of fuel cell components

    NASA Astrophysics Data System (ADS)

    Volfkovich, Yu. M.; Sosenkin, V. E.; Bagotsky, V. S.

    The operation of proton exchange membrane (PEMFC) and direct methanol fuel cells (DMFC) is connected with the flow of different gaseous and liquid components in the cell's membrane-electrode assembly (MEA). The structural and wetting properties of different components of the MEA influence the rate and direction of these flows and hence the fuel cell's efficiency. For a better understanding of the mechanism of all processes influencing the fuel cell efficiency, for a mathematical modelling of these processes, and for a possibility of their optimization, a detailed knowledge of the geometrical structure and wetting properties of all MEA components is necessary. This review describes the results of such investigations performed mainly by using the method of standard contact porosimetry (MSCP). This method gives the possibility to receive information on multicomponent porous and powdered materials hitherto not accessible, viz. their wetting and swelling properties, pore corrugation, and also isotherms of capillary pressure and bond energy. Measurements of MEA components by this method can be performed under exactly the same conditions (temperature, compression degree, contact with water, etc.) as those existing in real fuel cells.

  7. Life assessment of structural components using inelastic finite element analyses

    NASA Astrophysics Data System (ADS)

    Arya, Vinod K.; Halford, Gary R.

    1993-10-01

    The need for enhanced and improved performance of structural components subject to severe cyclic thermal/mechanical loadings, such as in the aerospace industry, requires development of appropriate solution technologies involving time-dependent inelastic analyses. Such analyses are mandatory to predict local stress-strain response and to assess more accurately the cyclic life time of structural components. The NASA-Lewis Research Center is cognizant of this need. As a result of concerted efforts at Lewis during the last few years, several such finite element solution technologies (in conjunction with the finite element program MARC) were developed and successfully applied to numerous uniaxial and multiaxial problems. These solution technologies, although developed for use with MARC program, are general in nature and can easily be extended for adaptation with other finite element programs such as ABAQUS, ANSYS, etc. The description and results obtained from two such inelastic finite element solution technologies are presented. The first employs a classical (non-unified) creep-plasticity model. An application of this technology is presented for a hypersonic inlet cowl-lip problem. The second of these technologies uses a unified creep-plasticity model put forth by Freed. The structural component for which this finite element solution technology is illustrated, is a cylindrical rocket engine thrust chamber. The advantages of employing a viscoplastic model for nonlinear time-dependent structural analyses are demonstrated. The life analyses for cowl-lip and cylindrical thrust chambers are presented. These analyses are conducted by using the stress-strain response of these components obtained from the corresponding finite element analyses.

  8. A life prediction model for laminated composite structural components

    NASA Technical Reports Server (NTRS)

    Allen, David H.

    1990-01-01

    A life prediction methodology for laminated continuous fiber composites subjected to fatigue loading conditions was developed. A summary is presented of research completed. A phenomenological damage evolution law was formulated for matrix cracking which is independent of stacking sequence. Mechanistic and physical support was developed for the phenomenological evolution law proposed above. The damage evolution law proposed above was implemented to a finite element computer program. And preliminary predictions were obtained for a structural component undergoing fatigue loading induced damage.

  9. Life assessment of structural components using inelastic finite element analyses

    NASA Technical Reports Server (NTRS)

    Arya, Vinod K.; Halford, Gary R.

    1993-01-01

    The need for enhanced and improved performance of structural components subject to severe cyclic thermal/mechanical loadings, such as in the aerospace industry, requires development of appropriate solution technologies involving time-dependent inelastic analyses. Such analyses are mandatory to predict local stress-strain response and to assess more accurately the cyclic life time of structural components. The NASA-Lewis Research Center is cognizant of this need. As a result of concerted efforts at Lewis during the last few years, several such finite element solution technologies (in conjunction with the finite element program MARC) were developed and successfully applied to numerous uniaxial and multiaxial problems. These solution technologies, although developed for use with MARC program, are general in nature and can easily be extended for adaptation with other finite element programs such as ABAQUS, ANSYS, etc. The description and results obtained from two such inelastic finite element solution technologies are presented. The first employs a classical (non-unified) creep-plasticity model. An application of this technology is presented for a hypersonic inlet cowl-lip problem. The second of these technologies uses a unified creep-plasticity model put forth by Freed. The structural component for which this finite element solution technology is illustrated, is a cylindrical rocket engine thrust chamber. The advantages of employing a viscoplastic model for nonlinear time-dependent structural analyses are demonstrated. The life analyses for cowl-lip and cylindrical thrust chambers are presented. These analyses are conducted by using the stress-strain response of these components obtained from the corresponding finite element analyses.

  10. Structural evaluations and dynamic testing of solar electric propulsion components.

    NASA Technical Reports Server (NTRS)

    Womack, J. R.; Chen, J.-C.

    1972-01-01

    Results of an experimental and analytical study of selected Solar Electric Propulsion (SEP) thrust subsystem components are presented. A mass and center-of-mass mockup of an experimental power conditioning (PC) panel was used for the PC structure evaluation. Test environment was based on that specified for prooftesting the Viking spacecraft electronics. Modal frequencies were surveyed over the range of 0-2000 Hz using both conventional test equipment and the holographic interferometry technique. Results of the forced vibration tests point to the need for some additional stiffening to make the structure qualifiable.

  11. Incorporation of a hierarchical grid component structure into GRIDGEN

    NASA Technical Reports Server (NTRS)

    Steinbrenner, John P.; Chawner, John R.

    1993-01-01

    The underlying framework of the GRIDGEN multiple block grid generation system has been refined so that grid components are now stored within a hierarchical data structure. This restructuring has enhanced the usability of the software by allowing grids to be generated on a more intuitive level. This new framework also provides a means by which the multiple block system can be edited at most any level in the grid generation process. Editing tools are currently being added to GRIDGEN so that a change to the grid can be propagated backward and forward in the data hierarchy. The new data structure, the editing tools, and other recent GRIDGEN improvements are described in this paper.

  12. [Endothelial glycocalyx of blood circulation. I. Finding, components, structure organization].

    PubMed

    Maksimenko, A V; Turashev, A D

    2014-01-01

    In normal state, a complex multicomponent system called glycocalyx is present on the surface of endothelial vascular system. The structure of the glycocalyx is determined by a group ofproteoglycans, glycoproteins and glycosaminoglycans, originating from endothelial cells and blood flow. Due to its complexity and location on the border of the system of blood circulation, glycocalyx participates in a number of functions supporting the metabolism of the vascular wall. Complete or partial loss of this structure in pathologicalconditions leads to inconsistencies in the vascular wall and changes in its functions. The first part of this review considers the history of detection and determination of endothelial glycocalyx structure, utilized methods and approaches. The molecular composition of the glycocalyx, properties of its components and glycocalyx structure organization are described. The English version of the paper: Russian Journal of Bioorganic Chemistry, see also http://www.maik.ru.

  13. NEUTRONIC REACTOR

    DOEpatents

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  14. COBRA-NC: a thermal hydraulics code for transient analysis of nuclear reactor components. Volume 2. COBRA-NC numerical solution methods

    SciTech Connect

    Thurgood, M.J.; George, T.L.; Wheeler, C.L.

    1986-04-01

    The COBRA-NC computer program has been developed to predict the thermal-hydraulic response of nuclear reactor components to thermal-hydraulic transients. The code solves the multicomponent, compressible three-dimensional, two-fluid, three-field equations for two-phase flow. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. The code has been used to model flow and heat transfer within the reactor core, the reactor vessel, the steam generators, and in the nuclear containment. This volume describes the finite-volume equations and the numerical solution methods used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of the numerical methods used to obtain a solution to the hydrodynamic equations.

  15. COBRA-NC: a thermal-hydraulic code for transient analysis of nuclear reactor components. Equations and constitutive models. Volume 1

    SciTech Connect

    Wheeler, C.L.; Thurgood, M.J.; Guidotti, T.E.; DeBellis, D.E.

    1986-05-01

    COBRA-NC is a digital computer program written in FORTRAN IV that simulates the response of nuclear reactor components and systems to thermal-hydraulic transients. The code solves the multicomponent, compressible, three-dimensional, two-fluid, three-field equations for two-phase flow. The three velocity fields are the vapor/gas field, the continuous liquid field, and the liquid drop field. The code has been used to model flow and heat transfer within the reactor core, the reactor vessel, the steam generators, and in the nuclear containment. The conservation equations, equations of state, and physical models that are common to all applications are presented in this volume of the code documentation.

  16. Computer-aided design of antenna structures and components

    NASA Technical Reports Server (NTRS)

    Levy, R.

    1976-01-01

    This paper discusses computer-aided design procedures for antenna reflector structures and related components. The primary design aid is a computer program that establishes cross sectional sizes of the structural members by an optimality criterion. Alternative types of deflection-dependent objectives can be selected for designs subject to constraints on structure weight. The computer program has a special-purpose formulation to design structures of the type frequently used for antenna construction. These structures, in common with many in other areas of application, are represented by analytical models that employ only the three translational degrees of freedom at each node. The special-purpose construction of the program, however, permits coding and data management simplifications that provide advantages in problem size and execution speed. Size and speed are essentially governed by the requirements of structural analysis and are relatively unaffected by the added requirements of design. Computation times to execute several design/analysis cycles are comparable to the times required by general-purpose programs for a single analysis cycle. Examples in the paper illustrate effective design improvement for structures with several thousand degrees of freedom and within reasonable computing times.

  17. Fuel processing in integrated micro-structured heat-exchanger reactors

    NASA Astrophysics Data System (ADS)

    Kolb, G.; Schürer, J.; Tiemann, D.; Wichert, M.; Zapf, R.; Hessel, V.; Löwe, H.

    Micro-structured fuel processors are under development at IMM for different fuels such as methanol, ethanol, propane/butane (LPG), gasoline and diesel. The target application are mobile, portable and small scale stationary auxiliary power units (APU) based upon fuel cell technology. The key feature of the systems is an integrated plate heat-exchanger technology which allows for the thermal integration of several functions in a single device. Steam reforming may be coupled with catalytic combustion in separate flow paths of a heat-exchanger. Reactors and complete fuel processors are tested up to the size range of 5 kW power output of a corresponding fuel cell. On top of reactor and system prototyping and testing, catalyst coatings are under development at IMM for numerous reactions such as steam reforming of LPG, ethanol and methanol, catalytic combustion of LPG and methanol, and for CO clean-up reactions, namely water-gas shift, methanation and the preferential oxidation of carbon monoxide. These catalysts are investigated in specially developed testing reactors. In selected cases 1000 h stability testing is performed on catalyst coatings at weight hourly space velocities, which are sufficiently high to meet the demands of future fuel processing reactors.

  18. Structural dynamics test simulation and optimization for aerospace components

    SciTech Connect

    Klenke, S.E.; Baca, T.J.

    1996-06-01

    This paper initially describes an innovative approach to product realization called Knowledge Based Testing (KBT). This research program integrates test simulation and optimization software, rapid fabrication techniques and computational model validation to support a new experimentally-based design concept. This design concept implements well defined tests earlier in the design cycle enabling the realization of highly reliable aerospace components. A test simulation and optimization software environment provides engineers with an essential tool needed to support this KBT approach. This software environment, called the Virtual Environment for Test Optimization (VETO), integrates analysis and test based models to support optimal structural dynamic test design. A goal in developing this software tool is to provide test and analysis engineers with a capability of mathematically simulating the complete structural dynamics test environment within a computer. A developed computational model of an aerospace component can be combined with analytical and/or experimentally derived models of typical structural dynamic test instrumentation within the VETO to determine an optimal test design. The VETO provides the user with a unique analysis and visualization environment to evaluate new and existing test methods in addition to simulating specific experiments designed to maximize test based information needed to validate computational models. The results of both a modal and a vibration test design are presented for a reentry vehicle and a space truss structure.

  19. Structure of the vault, a ubiquitous celular component.

    PubMed

    Kong, L B; Siva, A C; Rome, L H; Stewart, P L

    1999-04-15

    The vault is a ubiquitous and highly conserved ribonucleoprotein particle of approximately 13 MDa. This particle has been shown to be upregulated in certain multidrug-resistant cancer cell lines and to share a protein component with the telomerase complex. Determination of the structure of the vault was undertaken to provide a first step towards understanding the role of this cellular component in normal metabolism and perhaps to shed some light on its role in mediating drug resistance. Over 1300 particle images were combined to calculate an approximately 31 A resolution structure of the vault. Rotational power spectra did not yield a clear symmetry peak, either because of the thin, smooth walls or inherent flexibility of the vault. Although cyclic eightfold (C8) symmetry was imposed, the resulting reconstruction may be partially cylindrically averaged about the eightfold axis. Our results reveal the vault to be a hollow, barrel-like structure with two protruding caps and an invaginated waist. Although the normal cellular function of the vault is as yet undetermined, the structure of the vault is consistent with either a role in subcellular transport, as previously suggested, or in sequestering macromolecular assemblies.

  20. Crystal structure of the γ-secretase component nicastrin

    PubMed Central

    Xie, Tian; Yan, Chuangye; Zhou, Rui; Zhao, Yanyu; Sun, Linfeng; Yang, Guanghui; Lu, Peilong; Ma, Dan; Shi, Yigong

    2014-01-01

    γ-Secretase is an intramembrane protease responsible for the generation of amyloid-β (Aβ) peptides. Aberrant accumulation of Aβ leads to the formation of amyloid plaques in the brain of patients with Alzheimer's disease. Nicastrin is the putative substrate-recruiting component of the γ-secretase complex. No atomic-resolution structure had been identified on γ-secretase or any of its four components, hindering mechanistic understanding of γ-secretase function. Here we report the crystal structure of nicastrin from Dictyostelium purpureum at 1.95-Å resolution. The extracellular domain of nicastrin contains a large lobe and a small lobe. The large lobe of nicastrin, thought to be responsible for substrate recognition, associates with the small lobe through a hydrophobic pivot at the center. The putative substrate-binding pocket is shielded from the small lobe by a lid, which blocks substrate entry. These structural features suggest a working model of nicastrin function. Analysis of nicastrin structure provides insights into the assembly and architecture of the γ-secretase complex. PMID:25197054

  1. Effect of Different Structural Materials on Neutronic Performance of a Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Übeyli, Mustafa; Tel, Eyyüp

    2003-06-01

    Selection of structural material for a fusion-fission (hybrid) reactor is very important by taking into account of neutronic performance of the blanket. Refractory metals and alloys have much higher operating temperatures and neutron wall load (NWL) capabilities than low activation materials (ferritic/martensitic steels, vanadium alloys and SiC/SiC composites) and austenitic stainless steels. In this study, effect of primary candidate refractory alloys, namely, W-5Re, T111, TZM and Nb-1Zr on neutronic performance of the hybrid reactor was investigated. Neutron transport calculations were conducted with the help of SCALE 4.3 System by solving the Boltzmann transport equation with code XSDRNPM. Among the investigated structural materials, tantalum had the worst performance due to the fact that it has higher neutron absorption cross section than others. And W-5Re and TZM having similar results showed the best performance.

  2. DETERMINING THE EFFECTS OF RADIATION ON AGING CONCRETE STRUCTURES OF NUCLEAR REACTORS

    SciTech Connect

    Serrato, M.

    2010-01-29

    The U.S. Department of Energy Office of Environmental Management (DOE-EM) is responsible for the Decontamination and Decommissioning (D&D) of nuclear facilities throughout the DOE Complex. Some of these facilities will be completely dismantled, while others will be partially dismantled and the remaining structure will be stabilized with cementitious fill materials. The latter is a process known as In-Situ Decommissioning (ISD). The ISD decision process requires a detailed understanding of the existing facility conditions, and operational history. System information and material properties are need for aged nuclear facilities. This literature review investigated the properties of aged concrete structures affected by radiation. In particular, this review addresses the Savannah River Site (SRS) isotope production nuclear reactors. The concrete in the reactors at SRS was not seriously damaged by the levels of radiation exposure. Loss of composite compressive strength was the most common effect of radiation induced damage documented at nuclear power plants.

  3. Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors

    SciTech Connect

    William Richins; Stephen Novascone; Cheryl O'Brien

    2009-08-01

    Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors William Richins1, Stephen Novascone1, and Cheryl O’Brien1 1Idaho National Laboratory, US Dept. of Energy, Idaho Falls, Idaho, USA, e-mail: William.Richins@inl.gov The Idaho National Laboratory (INL, USA) and IASMiRT sponsored an international forum Nov 5-6, 2008 in Porvoo, Finland for nuclear industry, academic, and regulatory representatives to identify structural issues in current and future advanced reactor design, especially for extreme conditions and external threats. The purpose of this Topical Workshop was to articulate research, engineering, and regulatory Code development needs. The topics addressed by the Workshop were selected to address critical industry needs specific to advanced reactor structures that have long lead times and can be the subject of future SMiRT technical sessions. The topics were; 1) structural/materials needs for extreme conditions and external threats in contemporary (Gen. III) and future (Gen. IV and NGNP) advanced reactors and 2) calibrating simulation software and methods that address topic 1 The workshop discussions and research needs identified are presented. The Workshop successfully produced interactive discussion on the two topics resulting in a list of research and technology needs. It is recommended that IASMiRT communicate the results of the discussion to industry and researchers to encourage new ideas and projects. In addition, opportunities exist to retrieve research reports and information that currently exists, and encourage more international cooperation and collaboration. It is recommended that IASMiRT continue with an off-year workshop series on select topics.

  4. Structural Design Challenges in Design Certification Applications for New Reactors

    SciTech Connect

    Miranda, M.; Braverman, J.; Wei, X.; Hofmayer, C.; Xu, J.

    2011-07-17

    The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are confined within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of structural design chal- lenges encountered in recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

  5. Preliminary Measurements Supporting Reactor Vessel and Large Component Inspection Using X-Ray Backscatter Radiography by Selective Detection

    SciTech Connect

    Shedlock, Daniel; Dugan, Edward T.; Jacobs, Alan M.; Houssay, Laurent

    2006-07-01

    X-ray backscatter radiography by selective detection (RSD) is a field tested and innovative approach to non-destructive evaluation (NDE). RSD is an enhanced single-side x-ray Compton backscatter imaging (CBI) technique which selectively detects scatter components to improve image contrast and quality. Scatter component selection is accomplished through a set of specially designed detectors with fixed and movable collimators. Experimental results have shown that this NDE technique can be used to detect boric acid deposition on a metallic plate through steel foil reflective insulation commonly covering reactor pressure vessels. The current system is capable of detecting boric acid deposits with sub-millimeter resolution, through such insulating materials. Industrial systems have been built for Lockheed Martin Space Co. and NASA. Currently the x-ray backscatter RSD scanning systems developed by the University of Florida are being used to inspect the spray-on foam insulation (SOFI) used on the external tank of the space shuttle. RSD inspection techniques have found subsurface cracking in the SOFI thought to be responsible for the foam debris which separated from the external tank during the last shuttle launch. These industrial scanning systems can be customized for many applications, and a smaller, lighter, more compact unit design is being developed. The smaller design is approximately four inches wide, three inches high, and about 12 inches in length. This smaller RSD system can be used for NDE of areas that cannot be reached with larger equipment. X-ray backscatter RSD is a proven technology that has been tested on a wide variety of materials and applications. Currently the system has been used to inspect materials such as aluminum, plastics, honeycomb laminates, reinforced carbon composites, steel, and titanium. The focus of RSD is for one-sided detection for applications where conventional non-destructive examination methods either will not work or give poor

  6. Contributions of each isotope in structural material on radiation damage in a hybrid reactor

    NASA Astrophysics Data System (ADS)

    Günay, Mehtap

    2016-11-01

    In this study, the fluids were used in the liquid first-wall, blanket and shield zones of the designed hybrid reactor system. In this study, salt-heavy metal mixtures consisting of 93-85% Li20Sn80 + 5% SFG-PuO2 and 2-10% UO2, 93-85% Li20Sn80 + 5% SFG-PuO2 and 2-10% NpO2, and 93-85% Li20Sn80 + 5% SFG-PuO2 and 2-10% UCO were used as fluids. In this study, the effect on the radiation damage of spent fuel-grade (SFG)-PuO2, UO2, NpO2 and UCO contents was investigated in the structural material of a designed fusion-fission hybrid reactor system. In the designed hybrid reactor system were investigated the effect on the radiation damage of the selected fluid according to each isotopes of structural material in the structural material for 30 full power years (FPYs). Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library.

  7. Nde of Bonded Aluminum Components on Aircraft Structures

    NASA Astrophysics Data System (ADS)

    Barnard, Daniel J.; Hsu, David K.; Foreman, Cory; Wendt, Scott; Kreitinger, Nicholas A.; Steffes, Gary J.

    2008-02-01

    Bonded aluminum structures have been commonly used on aircraft for many years, and many of these applications include flight control surfaces. These bonded structures can be made up of aluminum face sheets adhesively bonded to a central honeycomb core, or they could also be composed of machined components that are bonded in a tongue-in-groove type manner called Grid-Lock. Nondestructive Inspection (NDI) methods of bonded aluminum structures usually involve the detection of skin-to-core disbonds, core buckling and damage caused by impacts. In the case of Grid-Lock, NDI techniques are focused on the detection of failures in the tongue-in-groove adhesive joint. Three nondestructive inspection methods were applied to honeycomb sandwich structures and Grid-Lock panels. The three methods were computer aided tap test (CATT), air-coupled ultrasonic testing (ACUT), and mechanical impedance analysis (MIA). The honeycomb structures tested consisted of structural panels and flight control surfaces from various aircraft. The Grid-Lock samples tested are laboratory specimens that simulate various defects. Experimental results and comparisons from each of these methods and samples will be presented.

  8. Composite Load Spectra for Select Space Propulsion Structural Components

    NASA Technical Reports Server (NTRS)

    Ho, Hing W.; Newell, James F.

    1994-01-01

    Generic load models are described with multiple levels of progressive sophistication to simulate the composite (combined) load spectra (CLS) that are induced in space propulsion system components, representative of Space Shuttle Main Engines (SSME), such as transfer ducts, turbine blades and liquid oxygen (LOX) posts. These generic (coupled) models combine the deterministic models for composite load dynamic, acoustic, high-pressure and high rotational speed, etc., load simulation using statistically varying coefficients. These coefficients are then determined using advanced probabilistic simulation methods with and without strategically selected experimental data. The entire simulation process is included in a CLS computer code. Applications of the computer code to various components in conjunction with the PSAM (Probabilistic Structural Analysis Method) to perform probabilistic load evaluation and life prediction evaluations are also described to illustrate the effectiveness of the coupled model approach.

  9. HWMA/RCRA CLOSURE PLAN FOR THE MATERIALS TEST REACTOR WING (TRA-604) LABORATORY COMPONENTS VOLUNTARY CONSENT ORDER ACTION PLAN VCO-5.8 D REVISION2

    SciTech Connect

    KIRK WINTERHOLLER

    2008-02-25

    This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan was developed for the laboratory components of the Test Reactor Area Catch Tank System (TRA-630) that are located in the Materials Test Reactor Wing (TRA-604) at the Reactor Technology Complex, Idaho National Laboratory Site, to meet a further milestone established under Voluntary Consent Order Action Plan VCO-5.8.d. The TRA-604 laboratory components addressed in this closure plan were deferred from the TRA-630 Catch Tank System closure plan due to ongoing laboratory operations in the areas requiring closure actions. The TRA-604 laboratory components include the TRA-604 laboratory warm wastewater drain piping, undersink drains, subheaders, and the east TRA-604 laboratory drain header. Potentially contaminated surfaces located beneath the TRA-604 laboratory warm wastewater drain piping and beneath the island sinks located in Laboratories 126 and 128 (located in TRA-661) are also addressed in this closure plan. The TRA-604 laboratory components will be closed in accordance with the interim status requirements of the Hazardous Waste Management Act/Resource Conservation and Recovery Act as implemented by the Idaho Administrative Procedures Act 58.01.05.009 and 40 Code of Federal Regulations 265, Subparts G and J. This closure plan presents the closure performance standards and the methods for achieving those standards.

  10. Process Performance and Bacterial Community Structure Under Increasing Influent Disturbances in a Membrane-Aerated Biofilm Reactor.

    PubMed

    Tian, Hailong; Yan, Yingchun; Chen, Yuewen; Wu, Xiaolei; Li, Baoan

    2016-02-01

    The membrane-aerated biofilm reactor (MABR) is a promising municipal wastewater treatment process. In this study, two cross-flow MABRs were constructed to explore the carbon and nitrogen removal performance and bacterial succession, along with changes of influent loading shock comprising flow velocity, COD, and NH4-N concentrations. Redundancy analysis revealed that the function of high flow velocity was mainly embodied in facilitating contaminants diffusion and biosorption rather than the success of overall bacterial populations (p > 0.05). In contrast, the influent NH4-N concentration contributed most to the variance of reactor efficiency and community structure (p < 0.05). Pyrosequencing results showed that Anaerolineae, and Beta- and Alphaproteobacteria were the dominant groups in biofilms for COD and NH4-N removal. Among the identified genera, Nitrosomonas and Nitrospira were the main nitrifiers, and Hyphomicrobium, Hydrogenophaga, and Rhodobacter were the key denitrifiers. Meanwhile, principal component analysis indicated that bacterial shift in MABR was probably the combination of stochastic and deterministic processes.

  11. Habitat, topographical, and geographical components structuring shrubsteppe bird communities

    USGS Publications Warehouse

    Knick, S.T.; Rotenberry, J.T.; Leu, M.

    2008-01-01

    Landscapes available to birds to select for breeding locations are arrayed along multiple dimensions. Identifying the primary gradients structuring shrubsteppe bird communities in the western United States is important because widespread habitat loss and alteration are shifting the environmental template on which these birds depend. We integrated field habitat surveys, GIS coverages, and bird counts from 61 Breeding Bird Survey routes located in shrubsteppe habitats across a >800 000 km2 region to determine the gradients of habitat, topography, and geography underlying bird communities. A small set of habitat features dominated the primary environmental gradients in a canonical ordination; the 13 species in the shrubsteppe bird community were closely packed along the first two axes. Using hierarchical variance partitioning, we identified habitat as the most important pure (31% explained variation) or shared component. Topography (9%) and geography (4%) were minor components but each shared a larger contribution with habitat (habitat-topography 21%; habitat-geography 22%) in explaining the organization of the bird community. In a second tier partition of habitat structure, pure composition (% land cover) was more important (45%) than configuration (patch size and edge) (7%); the two components shared 27% of the explained variation in the bird community axes. Local (9%), community (14%), and landscape (10%) levels contributed equally. Adjacent organizational levels had a larger shared contribution (local-community 26%; community-landscape 27%) than more separated local-landscape levels (21%). Extensive conversion of shrubsteppe habitats to agriculture, exotic annual grasslands, or pinyon (Pinus spp.)-juniper (Juniperus spp.) woodlands is occurring along the primary axes of habitat structure. Because the shrubsteppe bird community was organized along short gradients dominated by habitat features, relatively small shifts in their available environment will exert a

  12. Residual Strength Analysis Methodology: Laboratory Coupons to Structural Components

    NASA Technical Reports Server (NTRS)

    Dawicke, D. S.; Newman, J. C., Jr.; Starnes, J. H., Jr.; Rose, C. A.; Young, R. D.; Seshadri, B. R.

    2000-01-01

    The NASA Aircraft Structural Integrity (NASIP) and Airframe Airworthiness Assurance/Aging Aircraft (AAA/AA) Programs have developed a residual strength prediction methodology for aircraft fuselage structures. This methodology has been experimentally verified for structures ranging from laboratory coupons up to full-scale structural components. The methodology uses the critical crack tip opening angle (CTOA) fracture criterion to characterize the fracture behavior and a material and a geometric nonlinear finite element shell analysis code to perform the structural analyses. The present paper presents the results of a study to evaluate the fracture behavior of 2024-T3 aluminum alloys with thickness of 0.04 inches to 0.09 inches. The critical CTOA and the corresponding plane strain core height necessary to simulate through-the-thickness effects at the crack tip in an otherwise plane stress analysis, were determined from small laboratory specimens. Using these parameters, the CTOA fracture criterion was used to predict the behavior of middle crack tension specimens that were up to 40 inches wide, flat panels with riveted stiffeners and multiple-site damage cracks, 18-inch diameter pressurized cylinders, and full scale curved stiffened panels subjected to internal pressure and mechanical loads.

  13. Effect of sodium diclofenac loads on mesophase components and structure.

    PubMed

    Efrat, Rivka; Shalev, Deborah E; Hoffman, Roy E; Aserin, Abraham; Garti, Nissim

    2008-07-15

    We studied the effect of a model electrolytic drug on intermolecular interactions, conformational changes, and phase transitions in structured discontinuous cubic QL lyotropic liquid crystals. These changes were due to competition with hydration of the lipid headgroups. Structural changes of the phase induced by solubilization loads of sodium diclofenac (Na-DFC) were investigated by directly observing the water, ethanol, and Na-DFC components of the resulting phases using 2H and 23Na NMR. Na-DFC interacted with the surfactant glycerol monoolein (GMO) at the interface while interfering with the mesophase curvature and also competed with hydration of the surfactant headgroups. Increasing quantities of solubilized Na-DFC promoted phase transitions from cubic phase (discontinuous (QL) and bicontinuous (Q)) into lamellar structures and subsequently into a disordered lamellar phase. Quadrupolar coupling of deuterated ethanol by 2H NMR showed that it is located near the headgroups of the lipid and apparently is hydrogen bonded to the GMO headgroups. A phase transition between two lamellar phases (L alpha to L alpha*) was seen by 23Na NMR of Na-DFC at a concentration where the characteristics of the drug change from kosmotropic to chaotropic. These findings show that loads of solubilized drug may affect the structure of its vehicle and, as a result, its transport across skin-blood barriers. The structural changes of the mesophase may also aid controlled drug delivery.

  14. Performance robustness of the UASB reactors treating saline phenolic wastewater and analysis of microbial community structure.

    PubMed

    Wang, Wei; Wu, Benteng; Pan, Shanglei; Yang, Kai; Hu, Zhenhu; Yuan, Shoujun

    2017-06-05

    Anaerobic digestion was an important way to remove phenols from saline wastewater; however the anaerobic microorganisms were adversely affected by high concentration of salts. In order to clarify the performance robustness and microbial community structure for anaerobic digestion of saline phenolic wastewater, the UASB reactors were compared to treat phenolic wastewater under saline and non-saline conditions. The saline reactors were operated stably with phenols concentration increasing from 100 to 500mgL(-1) at 10g Na(+) L(-1). The robustness of the saline reactors was weakened at 1000mg phenols L(-1) and 10g Na(+) L(-1). However, the substrate utilization rates (SURs) for phenol, catechol, resorcinol, hydroquinone, and the specific methanogenic activity (SMA) of sludge were decreased by 95%, 85%, 97%, 78%, and 68%, respectively with phenols concentration enhancing from 1000 to 2000mgL(-1). Moreover, the SURs for phenol, catechol, resorcinol, hydroquinone, and the SMA of sludge were reduced by 32%, 65%, 74%, 45%, and 59%, respectively with Na(+) concentration increasing from 10 to 20gL(-1), in comparison with the values obtained at 10g Na(+) L(-1) and 1000mg phenols L(-1). Finally, the analysis of microbial community structure demonstrated that phenols degraders were less tolerant to high concentrations of Na(+) and phenols than methanogens.

  15. Magnons in one-dimensional k-component Fibonacci structures

    NASA Astrophysics Data System (ADS)

    Costa, C. H.; Vasconcelos, M. S.

    2014-05-01

    We have studied the magnon transmission through of one-dimensional magnonic k-component Fibonacci structures, where k different materials are arranged in accordance with the following substitution rule: Sn(k)=Sn-1(k)Sn-k(k) (n ≥k=0,1,2,…), where Sn(k) is the nth stage of the sequence. The calculations were carried out in exchange dominated regime within the framework of the Heisenberg model and taking into account the RPA approximation. We have considered multilayers composed of simple cubic spin-S Heisenberg ferromagnets, and, by using the powerful transfer-matrix method, the spin wave transmission is obtained. It is demonstrated that the transmission coefficient has a rich and interesting magnonic pass- and stop-bands structures, which depends on the frequency of magnons and the k values.

  16. Magnons in one-dimensional k-component Fibonacci structures

    SciTech Connect

    Costa, C. H.; Vasconcelos, M. S.

    2014-05-07

    We have studied the magnon transmission through of one-dimensional magnonic k-component Fibonacci structures, where k different materials are arranged in accordance with the following substitution rule: S{sub n}{sup (k)}=S{sub n−1}{sup (k)}S{sub n−k}{sup (k)} (n≥k=0,1,2,…), where S{sub n}{sup (k)} is the nth stage of the sequence. The calculations were carried out in exchange dominated regime within the framework of the Heisenberg model and taking into account the RPA approximation. We have considered multilayers composed of simple cubic spin-S Heisenberg ferromagnets, and, by using the powerful transfer-matrix method, the spin wave transmission is obtained. It is demonstrated that the transmission coefficient has a rich and interesting magnonic pass- and stop-bands structures, which depends on the frequency of magnons and the k values.

  17. Structures and Components in Galaxy Clusters: Observations and Models

    NASA Astrophysics Data System (ADS)

    Bykov, A. M.; Churazov, E. M.; Ferrari, C.; Forman, W. R.; Kaastra, J. S.; Klein, U.; Markevitch, M.; de Plaa, J.

    2015-05-01

    Clusters of galaxies are the largest gravitationally bounded structures in the Universe dominated by dark matter. We review the observational appearance and physical models of plasma structures in clusters of galaxies. Bubbles of relativistic plasma which are inflated by supermassive black holes of AGNs, cooling and heating of the gas, large scale plasma shocks, cold fronts, non-thermal halos and relics are observed in clusters. These constituents are reflecting both the formation history and the dynamical properties of clusters of galaxies. We discuss X-ray spectroscopy as a tool to study the metal enrichment in clusters and fine spectroscopy of Fe X-ray lines as a powerful diagnostics of both the turbulent plasma motions and the energetics of the non-thermal electron populations. The knowledge of the complex dynamical and feedback processes is necessary to understand the energy and matter balance as well as to constrain the role of the non-thermal components of clusters.

  18. Structural ECM components in the premetastatic and metastatic niche.

    PubMed

    Høye, Anette M; Erler, Janine T

    2016-06-01

    The aim of this review is to give an overview of the extracellular matrix (ECM) components that are important for creating structural changes in the premetastatic and metastatic niche. The successful arrival and survival of cancer cells that have left the primary tumor and colonized distant sites depends on the new microenvironment they encounter. The primary tumor itself releases factors into the circulation that travel to distant organs and then initiate structural changes, both non-enzymatic and enzymatic, to create a favorable niche for the disseminating tumor cells. Therapeutic strategies aimed at targeting cell-ECM interactions may well be one of the best viable approaches to combat metastasis and thus improve patient care.

  19. Technical Letter Report, An Evaluation of Ultrasonic Phased Array Testing for Reactor Piping System Components Containing Dissimilar Metal Welds, JCN N6398, Task 2A

    SciTech Connect

    Diaz, Aaron A.; Cinson, Anthony D.; Crawford, Susan L.; Anderson, Michael T.

    2009-11-30

    Research is being conducted for the U.S. Nuclear Regulatory Commission at the Pacific Northwest National Laboratory to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light-water reactor components. The scope of this research encom¬passes primary system pressure boundary materials including dissimilar metal welds (DMWs), cast austenitic stainless steels (CASS), piping with corrosion-resistant cladding, weld overlays, inlays and onlays, and far-side examinations of austenitic piping welds. A primary objective of this work is to evaluate various NDE methods to assess their ability to detect, localize, and size cracks in steel components that challenge standard and/or conventional inspection methodologies. This interim technical letter report provides a summary of a technical evaluation aimed at assessing the capabilities of phased-array (PA) ultrasonic testing (UT) methods as applied to the inspection of small-bore DMW components that exist in the reactor coolant systems (RCS) of pressurized water reactors (PWRs). Operating experience and events such as the circumferential cracking in the reactor vessel nozzle-to-RCS hot leg pipe at V.C. Summer nuclear power station, identified in 2000, show that in PWRs where primary coolant water (or steam) are present under normal operation, Alloy 82/182 materials are susceptible to pressurized water stress corrosion cracking. The extent and number of occurrences of DMW cracking in nuclear power plants (domestically and internationally) indicate the necessity for reliable and effective inspection techniques. The work described herein was performed to provide insights for evaluating the utility of advanced NDE approaches for the inspection of DMW components such as a pressurizer surge nozzle DMW, a shutdown cooling pipe DMW, and a ferritic (low-alloy carbon steel)-to-CASS pipe DMW configuration.

  20. Composite load spectra for select space propulsion structural components

    NASA Technical Reports Server (NTRS)

    Newell, J. F.; Ho, H. W.; Kurth, R. E.

    1991-01-01

    The work performed to develop composite load spectra (CLS) for the Space Shuttle Main Engine (SSME) using probabilistic methods. The three methods were implemented to be the engine system influence model. RASCAL was chosen to be the principal method as most component load models were implemented with the method. Validation of RASCAL was performed. High accuracy comparable to the Monte Carlo method can be obtained if a large enough bin size is used. Generic probabilistic models were developed and implemented for load calculations using the probabilistic methods discussed above. Each engine mission, either a real fighter or a test, has three mission phases: the engine start transient phase, the steady state phase, and the engine cut off transient phase. Power level and engine operating inlet conditions change during a mission. The load calculation module provides the steady-state and quasi-steady state calculation procedures with duty-cycle-data option. The quasi-steady state procedure is for engine transient phase calculations. In addition, a few generic probabilistic load models were also developed for specific conditions. These include the fixed transient spike model, the poison arrival transient spike model, and the rare event model. These generic probabilistic load models provide sufficient latitude for simulating loads with specific conditions. For SSME components, turbine blades, transfer ducts, LOX post, and the high pressure oxidizer turbopump (HPOTP) discharge duct were selected for application of the CLS program. They include static pressure loads and dynamic pressure loads for all four components, centrifugal force for the turbine blade, temperatures of thermal loads for all four components, and structural vibration loads for the ducts and LOX posts.

  1. Structural Components of Synaptic Plasticity and Memory Consolidation

    PubMed Central

    Bailey, Craig H.; Kandel, Eric R.; Harris, Kristen M.

    2015-01-01

    Consolidation of implicit memory in the invertebrate Aplysia and explicit memory in the mammalian hippocampus are associated with remodeling and growth of preexisting synapses and the formation of new synapses. Here, we compare and contrast structural components of the synaptic plasticity that underlies these two distinct forms of memory. In both cases, the structural changes involve time-dependent processes. Thus, some modifications are transient and may contribute to early formative stages of long-term memory, whereas others are more stable, longer lasting, and likely to confer persistence to memory storage. In addition, we explore the possibility that trans-synaptic signaling mechanisms governing de novo synapse formation during development can be reused in the adult for the purposes of structural synaptic plasticity and memory storage. Finally, we discuss how these mechanisms set in motion structural rearrangements that prepare a synapse to strengthen the same memory and, perhaps, to allow it to take part in other memories as a basis for understanding how their anatomical representation results in the enhanced expression and storage of memories in the brain. PMID:26134321

  2. The 5-kwe reactor thermoelectric system summary

    NASA Technical Reports Server (NTRS)

    Vanosdol, J. H. (Editor)

    1973-01-01

    Design of the 5-kwe reactor thermoelectric system was initiated in February 1972 and extended through the conceptual design phase into the preliminary design phase. Design effort was terminated in January, 1973. This report documents the system and component requirements, design approaches, and performance and design characteristics for the 5-kwe system. Included is summary information on the reactor, radiation shields, power conversion systems, thermoelectric pump, radiator/structure, liquid metal components, and the control system.

  3. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  4. X-Aerogels for Structural Components and High Temperature Applications

    NASA Technical Reports Server (NTRS)

    2005-01-01

    Future NASA missions and space explorations rely on the use of materials that are strong ultra lightweight and able to withstand extreme temperatures. Aerogels are low density (0.01-0.5 g/cu cm) high porosity materials that contain a glass like structure formed through standard sol-gel chemistry. As a result of these structural properties, aerogels are excellent thermal insulators and are able to withstand temperatures in excess of l,000 C. The open structure of aerogels, however, renders these materials extremely fragile (fracturing at stress forces less than 0.5 N/sq cm). The goal of NASA Glenn Research Center is to increase the strength of these materials by templating polymers and metals onto the surface of an aerogel network facilitating the use of this material for practical applications such as structural components of space vehicles used in exploration. The work this past year focused on two areas; (1) the research and development of new templated aerogels materials and (2) process development for future manufacturing of structural components. Research and development occurred on the production and characterization of new templating materials onto the standard silica aerogel. Materials examined included polymers such as polyimides, fluorinated isocyanates and epoxies, and, metals such as silver, gold and platinum. The final properties indicated that the density of the material formed using an isocyanate is around 0.50 g/cc with a strength greater than that of steel and has low thermal conductivity. The process used to construct these materials is extremely time consuming and labor intensive. One aspect of the project involved investigating the feasibility of shortening the process time by preparing the aerogels in the templating solvent. Traditionally the polymerization used THF as the solvent and after several washes to remove any residual monomers and water, the solvent around the aerogels was changed to acetonitrile for the templating step. This process

  5. Structure of the basal components of a bacterial transporter

    SciTech Connect

    Meisner, Jeffrey; Maehigashi, Tatsuya; André, Ingemar; Dunham, Christine M.; Moran, Jr., Charles P.

    2012-12-10

    Proteins SpoIIQ and SpoIIIAH interact through two membranes to connect the forespore and the mother cell during endospore development in the bacterium Bacillus subtilis. SpoIIIAH consists of a transmembrane segment and an extracellular domain with similarity to YscJ proteins. YscJ proteins form large multimeric rings that are the structural scaffolds for the assembly of type III secretion systems in Gram-negative bacteria. The predicted ring-forming motif of SpoIIIAH and other evidence led to the model that SpoIIQ and SpoIIIAH form the core components of a channel or transporter through which the mother cell nurtures forespore development. Therefore, to understand the roles of SpoIIIAH and SpoIIQ in channel formation, it is critical to determine whether SpoIIIAH adopts a ring-forming structural motif, and whether interaction of SpoIIIAH with SpoIIQ would preclude ring formation. We report a 2.8-{angstrom} resolution structure of a complex of SpoIIQ and SpoIIIAH. SpoIIIAH folds into the ring-building structural motif, and modeling shows that the structure of the SpoIIQ-SpoIIIAH complex is compatible with forming a symmetrical oligomer that is similar to those in type III systems. The inner diameters of the two most likely ring models are large enough to accommodate several copies of other integral membrane proteins. SpoIIQ contains a LytM domain, which is found in metalloendopeptidases, but lacks residues important for metalloprotease activity. Other LytM domains appear to be involved in protein-protein interactions. We found that the LytM domain of SpoIIQ contains an accessory region that interacts with SpoIIIAH.

  6. Local atomic structures of single-component metallic glasses

    NASA Astrophysics Data System (ADS)

    Trady, Salma; Hasnaoui, Abdellatif; Mazroui, M.'hammed; Saadouni, Khalid

    2016-10-01

    In this study we examine the structural properties of single-component metallic glasses of aluminum. We use a molecular dynamics simulation based on semi-empirical many-body potential, derived from the embedded atom method (EAM). The radial distribution function (RDF), common neighbors analysis method (CNA), coordination number analysis (CN) and Voronoi tessellation are used to characterize the metal's local structure during the heating and cooling (quenching). The simulation results reveal that the melting temperature depends on the heating rate. In addition, atomic visualization shows that the structure of aluminum after fast quenching is in a glassy state, confirmed quantitatively by the splitting of the second peak of the radial distribution function, and by the appearance of icosahedral clusters observed via CNA technique. On the other hand, the Wendt-Abraham parameters are calculated to determine the glass transition temperature (Tg), which depends strongly on the cooling rate; it increases while the cooling rate increases. On the basis of CN analysis and Voronoi tessellation, we demonstrate that the transition from the Al liquid to glassy state is mainly due to the formation of distorted and perfect icosahedral clusters.

  7. Novel durable bio-photocatalyst purifiers, a non-heterogeneous mechanism: accelerated entrapped dye degradation into structural polysiloxane-shield nano-reactors.

    PubMed

    Dastjerdi, Roya; Montazer, Majid; Shahsavan, Shadi; Böttcher, Horst; Moghadam, M B; Sarsour, Jamal

    2013-01-01

    This research has designed innovative Ag/TiO(2) polysiloxane-shield nano-reactors on the PET fabric to develop novel durable bio-photocatalyst purifiers. To create these very fine nano-reactors, oppositely surface charged multiple size nanoparticles have been applied accompanied with a crosslinkable amino-functionalized polysiloxane (XPs) emulsion. Investigation of photocatalytic dye decolorization efficiency revealed a non-heterogeneous mechanism including an accelerated degradation of entrapped dye molecules into the structural polysiloxane-shield nano-reactors. In fact, dye molecules can be adsorbed by both Ag and XPs due to their electrostatic interactions and/or even via forming a complex with them especially with silver NPs. The absorbed dye and active oxygen species generated by TiO(2) were entrapped by polysiloxane shelter and the presence of silver nanoparticles further attract the negative oxygen species closer to the adsorbed dye molecules. In this way, the dye molecules are in close contact with concentrated active oxygen species into the created nano-reactors. This provides an accelerated degradation of dye molecules. This non-heterogeneous mechanism has been detected on the sample containing all of the three components. Increasing the concentration of Ag and XPs accelerated the second step beginning with an enhanced rate. Further, the treated samples also showed an excellent antibacterial activity.

  8. Safety and environmental comparisons of stainless steel with alternative structural materials for fusion reactors

    SciTech Connect

    Kinzig, A.P.; Holdren, J.P.; Hibbard, P.J.

    1994-08-01

    Using the FuseDose II computer code, we calculated and compared several indices of safety and environmental (S&E) hazards for conceptual magnetic-fusion reactor designs based on a variety of structural materials-stainless steel, ferritic steel, vanadium-chromium-titanium alloy, and silicon-carbide-and, for comparison, the fuel of a liquid-metal fast breeder fission reactor. FuseDose II is a second-generation code derived from the Fuse-Dose code used in the U.S. Department of Energy`s Committee on Environmental, Safety, and Economic Aspects of Magnetic Fusion Energy (ESECOM) study in the late 1980s. The comparisons update and extend those of the ESECOM study by adding the stainless-steel case, some new indices, graphical representation of the results, and other refinements. The results of our analysis support earlier conclusions concerning the S&E liabilities of stainless steel: The use of stainless steel would significantly reduce the S&E advantages of fusion over fission that are implied by the indices we consider, compared with the advantages portrayed in the ESECOM results for lower-activation fusion materials. The dose potentials represented by the radioactive materials that conceivably could be mobilized in severe accidents are substantially higher for the stainless steel case than for the lower activation fusion designs analyzed by ESECOM, and the waste disposal burden imposed by a stainless steel fusion reactor, though significantly smaller than that associated with a fission reactor of the same output, is high enough to rule out the chance of qualification for shallow burial under current regulations (in contrast to some of the lower activation fusion cases). This work underscores the conclusion that research to demonstrate the viability of the low-activation materials is essential if fusion is to achieve its potential for large and easily demonstrated S&E advantages over fission. 37 refs., 17 figs., 8 tabs.

  9. Structure of glycopeptides isolated from bovine milk component PP3.

    PubMed

    Girardet, J M; Coddeville, B; Plancke, Y; Strecker, G; Campagna, S; Spik, G; Linden, G

    1995-12-15

    The heat-stable acid-soluble phosphoglycoprotein component PP3 was isolated from the bovine milk proteose peptone fraction by concanavalin A affinity chromatography. Glycopeptides were released by pronase digestion of the milk component PP3 and were subsequently separated by high-pH anion-exchange chromatography on CarboPac PA-1. The primary structures of the glycan and peptide moieties of eight N-glycopeptides have been established by combining methylation analysis, mass spectrometry, 400-MHz 1H-NMR spectroscopy, and peptide sequence analysis. All the analyzed fractions contained biantennary N-acetyllactosamine-type carbohydrate chains, some of them with a GalNAc(beta 1-4)GlcNAc or a NeuAc(alpha 2-6)GalNAc(beta 1-4)GlcNAc group. This particular sequence did or did not replace the Gal(beta 1-4)GlcNAc group usually found in most N-linked glycans. Moreover, the sialylated Gal and GalNAc residues were only found on the Man(alpha 1-3) antenna.

  10. Electrical Properties of Structural Components of the Crystalline Lens

    PubMed Central

    Mathias, R. T.; Rae, J. L.; Eisenberg, R. S.

    1979-01-01

    The electrical properties of the crystalline lens of the frog eye are measured with stochastic currents applied with a microelectrode near the center of the preparation and potential recorded just under the surface. The stochastic signals are decomposed by Fourier analysis into sinusoidal components, and the impedance is determined from the ratio of mean cross power to input power. The data are fit by an electrical model that includes two paths for current flow: one through the cytoplasm, gap junctions, and outer membrane; the other through inner membranes and the extracellular space between lens fibers. The electrical properties of the structures of the lens which appear as circuit components in the model are determined by the fit to the data. The resistivity of the extracellular space within the lens is comparable to the resistivity of Ringer. The outer membrane has a normal resistance of 5 kohm · cm2 but large capacitance of 10 μF/cm2, probably because it represents the properties of several layers of fibers. The inner membranes have properties reminiscent of artificial lipid bilayers: they have high membrane resistance, 2.2 megohm · cm2, and low specific capacitance, 0.8 μF/cm2. There is so much membrane within the lens, however, that the sum of the current flow across all the inner membranes is comparable to that across the outer surface. PMID:262384

  11. Induced radioactivity of LDEF materials and structural components

    NASA Technical Reports Server (NTRS)

    Harmon, B. A.; Laird, C. E.; Fishman, G. J.; Parnell, T. A.; Camp, D. C.; Frederick, C. E.; Hurley, D. L.; Lindstrom, D. J.; Moss, C. E.; Reedy, R. C.; hide

    1996-01-01

    We present an overview of the Long Duration Exposure Facility (LDEF) induced activation measurements. The LDEF, which was gravity-gradient stabilized, was exposed to the low Earth orbit (LEO) radiation environment over a 5.8 year period. Retrieved activation samples and structural components from the spacecraft were analyzed with low and ultra-low background HPGe gamma spectrometry at several national facilities. This allowed a very sensitive measurement of long-lived radionuclides produced by proton- and neutron-induced reactions in the time-dependent, non-isotropic LEO environment. A summary of major findings from this study is given that consists of directionally dependent activation, depth profiles, thermal neutron activation, and surface beryllium-7 deposition from the upper atmosphere. We also describe a database of these measurements that has been prepared for use in testing radiation environmental models and spacecraft design.

  12. Induced radioactivity of LDEF materials and structural components

    NASA Technical Reports Server (NTRS)

    Harmon, B. A.; Laird, C. E.; Fishman, G. J.; Parnell, T. A.; Camp, D. C.; Frederick, C. E.; Hurley, D. L.; Lindstrom, D. J.; Moss, C. E.; Reedy, R. C.; Reeves, J. H.; Smith, A. R.; Winn, W. G.; Benton, E. V.

    1996-01-01

    We present an overview of the Long Duration Exposure Facility (LDEF) induced activation measurements. The LDEF, which was gravity-gradient stabilized, was exposed to the low Earth orbit (LEO) radiation environment over a 5.8 year period. Retrieved activation samples and structural components from the spacecraft were analyzed with low and ultra-low background HPGe gamma spectrometry at several national facilities. This allowed a very sensitive measurement of long-lived radionuclides produced by proton- and neutron-induced reactions in the time-dependent, non-isotropic LEO environment. A summary of major findings from this study is given that consists of directionally dependent activation, depth profiles, thermal neutron activation, and surface beryllium-7 deposition from the upper atmosphere. We also describe a database of these measurements that has been prepared for use in testing radiation environmental models and spacecraft design.

  13. Performance categorization of structures, systems & components and related issues

    SciTech Connect

    Hossain, Q.A.

    1993-09-30

    Provisions of DOE-STD-1021-93 on performance categorization of structures, systems and components (SSCs) subjected to natural phenomena hazards (NPHs) are summarized. The interrelationship among safety classification of SSCs (per DOE 6430.1A and DOE 5480.30), facility hazard categorization/classification (per DOE 5481.1B and DOE 5480.23), and NPH performance categorization of SSCs (per DOE 5480.28 and DOE-STD-1021-93) is discussed. The compatibility between the safety goals in the Department of Energy Safety Policy, SEN-35-91, and the numerical NPH performance goals of DOE 5480.28, as presented in UCRL-ID-12612 (draft), is examined.

  14. 3D printed components with ultrasonically arranged microscale structure

    NASA Astrophysics Data System (ADS)

    Llewellyn-Jones, Thomas M.; Drinkwater, Bruce W.; Trask, Richard S.

    2016-02-01

    This paper shows the first application of in situ manipulation of discontinuous fibrous structure mid-print, within a 3D printed polymeric composite architecture. Currently, rapid prototyping methods (fused filament fabrication, stereolithography) are gaining increasing popularity within the engineering commnity to build structural components. Unfortunately, the full potential of these components is limited by the mechanical properties of the materials used. The aim of this study is to create and demonstrate a novel method to instantaneously orient micro-scale glass fibres within a selectively cured photocurable resin system, using ultrasonic forces to align the fibres in the desired 3D architecture. To achieve this we have mounted a switchable, focused laser module on the carriage of a three-axis 3D printing stage, above an in-house ultrasonic alignment rig containing a mixture of photocurable resin and discontinuous 14 μm diameter glass fibre reinforcement(50 μm length). In our study, a suitable print speed of 20 mm s-1 was used, which is comparable to conventional additive layer techniques. We show the ability to construct in-plane orthogonally aligned sections printed side by side, where the precise orientation of the configurations is controlled by switching the ultrasonic standing wave profile mid-print. This approach permits the realisation of complex fibrous architectures within a 3D printed landscape. The versatile nature of the ultrasonic manipulation technique also permits a wide range of particle types (diameters, aspect ratios and functions) and architectures (in-plane, and out-plane) to be patterned, leading to the creation of a new generation of fibrous reinforced composites for 3D printing.

  15. Microbial community structure and dynamics in anaerobic fluidized‐bed and granular sludge‐bed reactors: influence of operational temperature and reactor configuration

    PubMed Central

    Bialek, Katarzyna; Kumar, Amit; Mahony, Thérèse; Lens, Piet N. L.; O' Flaherty, Vincent

    2012-01-01

    Summary Methanogenic community structure and dynamics were investigated in two different, replicated anaerobic wastewater treatment reactor configurations [inverted fluidized bed (IFB) and expanded granular sludge bed (EGSB)] treating synthetic dairy wastewater, during operating temperature transitions from 37°C to 25°C, and from 25°C to 15°C, over a 430‐day trial. Non‐metric multidimensional scaling (NMS) and moving‐window analyses, based on quantitative real‐time PCR data, along with denaturing gradient gel electrophoresis (DGGE) profiling, demonstrated that the methanogenic communities developed in a different manner in these reactor configurations. A comparable level of performance was recorded for both systems at 37°C and 25°C, but a more dynamic and diverse microbial community in the IFB reactors supported better stability and adaptative capacity towards low temperature operation. The emergence and maintenance of particular bacterial genotypes (phylum Firmicutes and Bacteroidetes) was associated with efficient protein hydrolysis in the IFB, while protein hydrolysis was inefficient in the EGSB. A significant community shift from a Methanobacteriales and Methanosaetaceae towards a Methanomicrobiales‐predominated community was demonstrated during operation at 15°C in both reactor configurations. PMID:22967313

  16. Thermochemical storage for CSP via redox structured reactors/heat exchangers: The RESTRUCTURE project

    NASA Astrophysics Data System (ADS)

    Karagiannakis, George; Pagkoura, Chrysoula; Konstandopoulos, Athanasios G.; Tescari, Stefania; Singh, Abhishek; Roeb, Martin; Lange, Matthias; Marcher, Johnny; Jové, Aleix; Prieto, Cristina; Rattenbury, Michael; Chasiotis, Andreas

    2017-06-01

    The present work provides an overview of activities performed in the framework of the EU-funded collaborative project RESTRUCTURE, the main goal of which was to develop and validate a compact structured reactor/heat exchanger for thermochemical storage driven by 2-step high temperature redox metal oxide cycles. The starting point of development path included redox materials qualification via both theoretical and lab-scale experimental studies. Most favorable compositions were cobalt oxide/alumina composites. Preparation of small-scale structured bodies included various approaches, ranging from perforated pellets to more sophisticated honeycomb geometries, fabricated by extrusion and coating. Proof-of-concept of the proposed novel reactor/heat exchanger was successfully validated in small-scale structures and the next step included scaling up of redox honeycombs production. Significant challenges were identified for the case of extruded full-size bodies and the final qualified approach related to preparation of cordierite substrates coated with cobalt oxide. The successful experimental evaluation of the pilot reactor/heat exchanger system constructed motivated the preliminary techno-economic evaluation of the proposed novel thermochemical energy storage concept. Taking into account experimental results, available technologies and standard design aspects a model for a 70.5 MWe CSP plant was defined. Estimated LCOE costs were calculated to be in the range of reference values for Combined Cycle Power Plants operated by natural gas. One of main cost contributors was the storage system itself, partially due to relatively high cost of cobalt oxide. This highlighted the need to identify less costly and equally efficient to cobalt oxide redox materials.

  17. Bonding and structure in dense multi-component molecular mixtures

    DOE PAGES

    Meyer, Edmund R.; Ticknor, Christopher; Bethkenhagen, Mandy; ...

    2015-10-30

    We have performed finite-temperature density functional theory molecular dynamics simulations on dense methane, ammonia, and water mixtures (CH4:NH3:H2O) for various compositions and temperatures (2000 K ≤ T ≤ 10000 K) that span a set of possible conditions in the interiors of ice-giant exoplanets. The equation-of-state, pair distribution functions, and bond autocorrelation functions (BACF) were used to probe the structure and dynamics of these complex fluids. In particular, an improvement to the choice of the cutoff in the BACF was developed that allowed analysis refinements for density and temperature effects. We note the relative changes in the nature of these systemsmore » engendered by variations in the concentration ratios. As a result, a basic tenet emerges from all these comparisons that varying the relative amounts of the three heavy components (C,N,O) can effect considerable changes in the nature of the fluid and may in turn have ramifications for the structure and composition of various planetary layers.« less

  18. Bonding and structure in dense multi-component molecular mixtures

    SciTech Connect

    Meyer, Edmund R.; Ticknor, Christopher; Bethkenhagen, Mandy; Hamel, Sebastien; Redmer, Ronald; Kress, Joel D.; Collins, Lee A.

    2015-10-30

    We have performed finite-temperature density functional theory molecular dynamics simulations on dense methane, ammonia, and water mixtures (CH4:NH3:H2O) for various compositions and temperatures (2000 K ≤ T ≤ 10000 K) that span a set of possible conditions in the interiors of ice-giant exoplanets. The equation-of-state, pair distribution functions, and bond autocorrelation functions (BACF) were used to probe the structure and dynamics of these complex fluids. In particular, an improvement to the choice of the cutoff in the BACF was developed that allowed analysis refinements for density and temperature effects. We note the relative changes in the nature of these systems engendered by variations in the concentration ratios. As a result, a basic tenet emerges from all these comparisons that varying the relative amounts of the three heavy components (C,N,O) can effect considerable changes in the nature of the fluid and may in turn have ramifications for the structure and composition of various planetary layers.

  19. Development and fabrication of structural components for a scramjet engine

    NASA Technical Reports Server (NTRS)

    Buchmann, O. A.

    1990-01-01

    A program broadly directed toward design and development of long-life (100 hours and 1,000 cycles with a goal of 1,000 hours and 10,000 cycles) hydrogen-cooled structures for application to scramjets is presented. Previous phases of the program resulted in an overall engine design and analytical and experimental characterization of selected candidate materials and concepts. The latter efforts indicated that the basic life goals for the program can be reached with available means. The main objective of this effort was an integrated, experimental evaluation of the results of the previous program phases. The fuel injection strut was selected for this purpose, including fabrication development and fabrication of a full-scale strut. Testing of the completed strut was to be performed in a NASA-Langley wind tunnel. In addition, conceptual designs were formulated for a heat transfer test unit and a flat panel structural test unit. Tooling and fabrication procedures required to fabricate the strut were developed, and fabrication and delivery to NASA of all strut components, including major subassemblies, were completed.

  20. Structure and phase behavior in five-component microemulsions

    SciTech Connect

    Billman, J.F. ); Kaler, E.W. )

    1990-03-01

    Droplet-to-bicontinuous structure transitions in a family of five-component microemulsions formed with sodium 4-(1{prime}-heptylnonyl)benzenesulfonate, isobutyl alcohol, D{sub 2}O, sodium chloride, and alkanes with even carbon numbers from octane to hexadecane are probed by using small-angle neutron scattering, electrical conductivity, and NMR self-diffusion measurements. The phase behavior and structure of these microemulsions are intimately linked and depend on salinity and the chain length of the alkane. Both the range of salt concentration in which the three-phase region is observed and the range of microemulsion water volume fraction within the three-phase region decrease with decreasing alkane chain length. Further, the appearance of the three-phase region is preceded by droplet-to-bicontinuous transitions. Microemulsions not exhibiting three-phase regions become bicontinuous only when they contain equal amounts of oil and water. The coincidence of the so-called percolation thresholds as determined by using electrical conductivity and self-diffusion measurements shows that electrical conduction in a dispersion of water droplets occurs with the exchange of material between the droplets.

  1. Quantifying Ecosystem Structural Components with Highly Portable Lidar

    NASA Astrophysics Data System (ADS)

    Schaaf, C.; Paynter, I.; Peri, F.; Saenz, E. J.; Genest, D.; Strahler, A. H.; Li, Z.

    2015-12-01

    Terrestrial laser scanners (TLS), which utilize light detection and ranging (lidar) have demonstrated the ability to produce accurate reconstructions of ecosystems, including spatially complex systems such as forests. Reconstructions at the object or plot scale can be used to interpret or simulate satellite observations, particularly for lidar instruments such as those involved in the forthcoming GEDI and ICESat 2 missions. The Compact Biomass Lidar (CBL) is a TLS optimized for portability and scanning speed, developed and operated by University of Massachusetts Boston. This 905nm wavelength scanner achieves an angular resolution of 0.25 degrees at a rate of 33 seconds per scan. The rapid scanning of the CBL and similar highly portable TLS improve acquisition of 3D surfaces such as canopy height models and digital elevation models derived from point clouds. This is due to the ability to capture additional scanning points within the window of temporal stability for the ecosystem, mitigating the rapid loss of information density associated with distance and occlusion. Utilizing terrestrial lidar in tandem with airborne lidar profiles vertically distributed structural components of ecosystems, such as the canopy of forests. We will present 3D surfaces documenting the growth of vegetation species including the invasive Phragmites australis over the 2015 growing season at Plum Island Long Term Ecological Research sites, derived from CBL. Additionally we will show vertical structure profiles from voxelization analyses in tropical forest (La Selva, Costa Rica) and temperate forest (Harvard Forest, MA, USA). We will discuss and present results from emerging point cloud reconstruction methods, including the Quantitative Structure Model (QSM) for tree modeling, and their implications particularly for GEDI-related calibration and validation studies.

  2. Structural biology facilities at Brookhaven National Laboratory`s high flux beam reactor

    SciTech Connect

    Korszun, Z.R.; Saxena, A.M.; Schneider, D.K.

    1994-12-31

    The techniques for determining the structure of biological molecules and larger biological assemblies depend on the extent of order in the particular system. At the High Flux Beam Reactor at the Brookhaven National Laboratory, the Biology Department operates three beam lines dedicated to biological structure studies. These beam lines span the resolution range from approximately 700{Angstrom} to approximately 1.5{Angstrom} and are designed to perform structural studies on a wide range of biological systems. Beam line H3A is dedicated to single crystal diffraction studies of macromolecules, while beam line H3B is designed to study diffraction from partially ordered systems such as biological membranes. Beam line H9B is located on the cold source and is designed for small angle scattering experiments on oligomeric biological systems.

  3. Design and evaluation of experimental ceramic automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1974-01-01

    The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

  4. Exploratory evaluation of ceramics for automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1972-01-01

    An exploratory evaluation of ceramics for automobile thermal reactors was conducted. Potential ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance lasting over 800 hours in engine dynamometer tests and over 15,000 miles (24,200 km) of vehicle road tests. Reactors containing glass-ceramic components did not perform as well as silicon carbide. But the glass-ceramics still offer good potential for reactor use. The results of this study are considered to be a reasonable demonstration of the potential use of ceramics in thermal reactors.

  5. Design and evaluation of experimental ceramic automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1974-01-01

    The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

  6. Evaluation of radionuclide penetration of structural concrete surfaces in the Three Mile Island Unit 2 reactor building

    SciTech Connect

    Davis, C.M.

    1988-01-01

    The March 28, 1979 loss-of-coolant accident at Three Mile Island Unit 2 (TMI-2) resulted in the exposure of /approx/3000 m/sup 2/ of reactor building internal concrete surfaces to both liquid-and vapor-borne contaminants. The period of contact between the major structural concrete surfaces and the aqueous solutions of mixed fission products ranged from a few days to several years. Exclusive of the reactor building basement impingement walls above a height of 1.6 m, all concrete surfaces were protected with an epoxy-based coating. This coating provides a tough, easily decontaminated surface for the concrete during normal operation and maintenance cycles. At the completion of the gross decontamination of the accessible reactor building elevations in 1982, exposure rates remained elevated above expected levels as indicated by early decontamination factors. Exposure rate measurements and small-scale scarification samples of the reactor building surfaces demonstrated that the protective coatings and concrete in the reactor building had absorbed radionuclides, thereby creating a large fixed source. In September 1983, a concrete core sampling program was conducted in the TMI-2 reactor building to assess the depth of contaminant penetration into the coatings and concrete on elevations 93m and 106 m. Sampling of the reactor building basement concrete surfaces (elevation 86 m) was deferred until 1985 and 1986 to provide lead time for remote systems development.

  7. [New exploration on effect of characteristics of traditional Chinese medicine components structure on multi-ingredient/component pharmacokinetics].

    PubMed

    Gu, Jun-Fei; Feng, Liang; Zhang, Ming-Hua; Qin, Dong; Jia, Xiao-Bin

    2014-07-01

    The study on the pharmacokinetics of traditional Chinese medicines (TCMs) is a linking science during the modernization of TCMs, and plays an important role in the studies on the complex material base of TCMs, the in vivo process of ingredient/ component and the pharmacokinetics-pharmacodynamics correlation. However, because of the multi-ingredient/component system of TCMs, how to scientifically reveal the pharmacokinetics that is consistent with TCMs' characteristics has long been a hotspot and difficulty for the exploration. The optimal composition structure of the material basis of TCMs shows the best efficacy, while the difference between the multi-ingredient/component composition structures in the efficacy is closely related to their absorption, transport, metabolism and excretion in vivo. In this article, the authors systematically review the study methods for pharmacokinetics of TCMs and their compounds, and explore the pharmacokinetics of TCMs based on the "component structure theory". As a result, the method for integrating TCM component structure and the TCM pharmacokinetics was proposed to be adopted to intensively study the effect of the component structure on the in vivo TCM multi-ingredient/component pharmacokinetic characteristics, in order to promote the TCM modernization and innovation in China.

  8. Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

    NASA Astrophysics Data System (ADS)

    Natesan, K.; Li, Meimei; Chopra, O. K.; Majumdar, S.

    2009-07-01

    Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.

  9. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    SciTech Connect

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; Ghose, S.; Wells, P.; Stan, T.; Almirall, N.; Odette, G. R.; Ecker, L. E.

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitates that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.

  10. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    DOE PAGES

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; ...

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitatesmore » that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.« less

  11. Sandwich-structured enzyme membrane reactor for efficient conversion of maltose into isomaltooligosaccharides.

    PubMed

    Zhang, Lei; Su, Yanlei; Zheng, Yang; Jiang, Zhongyi; Shi, Jiafu; Zhu, Yuanyuan; Jiang, Yanjun

    2010-12-01

    A novel enzyme membrane reactor with sandwich structure has been developed by confining glucosidase between two sheets of ultrafiltration membranes to effectively convert maltose to isomaltooligosaccharides (IMOs). The hydrophilic ultrafiltration membranes, which were prepared by phase inversion method using PES as bulk polymer and Pluronic F127 as both surface modification and pore formation agent, exhibited the desirable enzyme adsorption-resistant property. The scanning electron microscopy (SEM) photographs showed that two sheets of PES/Pluronic F127 membranes were packed tightly and glucosidase was kept in a free state within a nanoscale space. When the weight ratio of Pluronic F127 to PES was 30%, glucosidase could be completely rejected by the membranes. Due to the sandwich structuring of the membrane reactor and the high hydrophilicity of the PES/Pluronic F127 membrane surface, maltose conversion and yield reached 100% and 58% under the optimum experimental conditions (pH 6.0, 50 degrees C), respectively. 2010 Elsevier Ltd. All rights reserved.

  12. 78 FR 19541 - Proposed Revision to Design of Structures, Components, Equipment and Systems

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-04-01

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Proposed Revision to Design of Structures, Components, Equipment and Systems AGENCY: Nuclear... comments on the proposed revisions in Chapter 3, ``Design of Structures, Components, Equipment, and...

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

    SciTech Connect

    Lexa, D.; Kropf, A.J.

    2006-07-01

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

  14. Structural dynamic and thermal stress analysis of nuclear reactor vessel support system

    NASA Technical Reports Server (NTRS)

    Chi-Diango, J.

    1972-01-01

    A nuclear reactor vessel is supported by a Z-ring and a box ring girder. The two proposed structural configurations to transmit the loads from the Z-ring and the box ring girder to the foundation are shown. The cantilever concrete ledge transmitting the load from the Z-ring and the box girder via the cavity wall to the foundation is shown, along with the loads being transmitted through one of the six steel columns. Both of these two supporting systems were analyzed by using rigid format 9 of NASTRAN for dynamic loads, and the thermal stresses were analyzed by AXISOL. The six column configuration was modeled by a combination of plate and bar elements, and the concrete cantilever ledge configuration was modeled by plate elements. Both configurations were found structurally satisfactory; however, nonstructural considerations favored the concrete cantilever ledge.

  15. Implicit finite element structural dynamic formulation for long-duration accidents in reactor piping systems

    SciTech Connect

    Wang, C.Y.

    1985-01-01

    This taper describes an implicit three-dimensional finite-element formulation for the structural analysis of reactor piping system. The numerical algorithm considers hoop, flexural, axial, and torsion modes of the piping structures. It is unconditionally stable and can be used for calculation of piping response under static or long duration dynamic loads. The method uses a predictor-corrector, successive iterative scheme which satisfies the equilibrium equations. A set of stiffness equations representing the discretized equations of motion are derived to predict the displacement increments. The calculated displacement increments are then used to correct the element nodal forces. The algorithm is fairly general, and is capable of treating large displacements and elastic-plastic materials with thermal and strain-rate effects. 7 refs., 7 figs.

  16. THE FINE STRUCTURE OF CORTICAL COMPONENTS OF PARAMECIUM MULTIMICRONUCLEATUM

    PubMed Central

    Sedar, Albert W.; Porter, Keith R.

    1955-01-01

    The electron microscope was used to study the structure and three dimensional relationships of the components of the body cortex in thin sections of Paramecium multimicronucleatum. Micrographs of sections show that the cortex is covered externally by two closely apposed membranes (together ∼250 A thick) constituting the pellicle. Beneath the pellicle the surface of the animal is molded into ridges that form a polygonal ridgework with depressed centers. It is these ridges that give the surface of the organism its characteristic configuration and correspond to the outer fibrillar system of the light microscope image. The outer ends of the trichocysts with their hood-shaped caps are located in the centers of the anterior and posterior ridges of each polygon. The cilia extend singly from the depressed centers of the surface polygons. Each cilium shows two axial filaments with 9 peripheral and parallel filaments embedded in a matrix and the whole surrouned by a thin ciliary membrane. The 9 peripheral filaments are double and these are evenly spaced in a circle around the central pair. The ciliary membrane is continuous with the outer member of the pellicular membrane, whereas the plasma membrane is continuous with the inner member of the pellicular membrane. At the level of the plasma membrane the proximal end of the cilium is continuous with its tube-shaped basal body or kinetosome. The peripheral filaments of the cilium, together with the material of cortical matrix which tends to condense around them, form the sheath of the basal body. The kinetodesma connecting the ciliary kinetosomes (inner fibrillar system of the light microscopist) is composed of a number of discrete fibrils which overlap in a shingle-like fashion. Each striated kinetosomal fibril originates from a ciliary kinetosome and runs parallel to other kinetosomal fibrils arising from posterior kinetosomes of a particular meridional array. Sections at the level of the ciliary kinetosomes reveal an

  17. CHEMICAL STRUCTURES IN COAL: GEOCHEMICAL EVIDENCE FOR THE PRESENCE OF MIXED STRUCTURAL COMPONENTS.

    USGS Publications Warehouse

    Hatcher, P.G.; Breger, I.A.; Maciel, G.E.; Szeverenyi, N.M.

    1983-01-01

    The purpose of this paper is to summarize work on the chemical structural components of coal, comparing them with their possible plant precursors in modern peat. Solid-state **1**3C nuclear magnetic resonance (NMR), infrared spectroscopy (IR), elemental analysis and, in some cases, individual compound analyses formed the bases for these comparisons.

  18. Two-phase integrated sludge thickening and digestion (TISTD) reactor microbial diversity and community structure succession rules.

    PubMed

    Qiang, He; Xingfu, Sun; Li, Gu; Hainan, Ai

    2014-12-01

    A two-phase integrated sludge thickening and digestion (TISTD) reactor composed of an inner and an outer reactor was developed. Acidification of natural organic material was the primary process in the outer reactor, whilst methane production was the dominant bioreaction occurring in the inner one. The special structure of TISTD thus enables the effective separation of the acid production phase and methane production phase during sludge processing. Molecular biological technology, including 16S rRNA gene and PCR-TGGE, was utilized to investigate the overall microbial community structure and diversity, as well as the processes of dynamic change. Analysis was also conducted on succinate dehydrogenase and coenzyme F420 change trends at each dosing ratio. The microbial community structure of the system exhibited disorder gradually and led to collapse when the dosing ratio increased above 30 %.

  19. 75 FR 78777 - Advisory Committee On Reactor Safeguards; Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-16

    ...: The Advisory Committee on Reactor Safeguards was established by Section 29 of the Atomic Energy Act... accident phenomena; design of nuclear power plant structures, systems and components; materials science...

  20. Review of Recent Aging-Related Degradation Occurrences of Structures and Passive Components in U.S. Nuclear Power Plants

    SciTech Connect

    Nie,J.; Braverman, J.; Hofmayer, C.; Choun, Y.-S.; Kim, M.K.; Choi, I.-K.

    2009-04-02

    The Korea Atomic Energy Research Institute (KAERI) and Brookhaven National Laboratory (BNL) are collaborating to develop seismic capability evaluation technology for degraded structures and passive components (SPCs) under a multi-year research agreement. To better understand the status and characteristics of degradation of SPCs in nuclear power plants (NPPs), the first step in this multi-year research effort was to identify and evaluate degradation occurrences of SPCs in U.S. NPPs. This was performed by reviewing recent publicly available information sources to identify and evaluate the characteristics of degradation occurrences and then comparing the information to the observations in the past. Ten categories of SPCs that are applicable to Korean NPPs were identified, comprising of anchorage, concrete, containment, exchanger, filter, piping system, reactor pressure vessel, structural steel, tank, and vessel. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.

  1. Effects of the antimicrobial tylosin on the microbial community structure of an anaerobic sequencing batch reactor.

    PubMed

    Shimada, Toshio; Li, Xu; Zilles, Julie L; Morgenroth, Eberhard; Raskin, Lutgarde

    2011-02-01

    The effects of the antimicrobial tylosin on a methanogenic microbial community were studied in a glucose-fed laboratory-scale anaerobic sequencing batch reactor (ASBR) exposed to stepwise increases of tylosin (0, 1.67, and 167 mg/L). The microbial community structure was determined using quantitative fluorescence in situ hybridization (FISH) and phylogenetic analyses of bacterial 16S ribosomal RNA (rRNA) gene clone libraries of biomass samples. During the periods without tylosin addition and with an influent tylosin concentration of 1.67 mg/L, 16S rRNA gene sequences related to Syntrophobacter were detected and the relative abundance of Methanosaeta species was high. During the highest tylosin dose of 167 mg/L, 16S rRNA gene sequences related to Syntrophobacter species were not detected and the relative abundance of Methanosaeta decreased considerably. Throughout the experimental period, Propionibacteriaceae and high GC Gram-positive bacteria were present, based on 16S rRNA gene sequences and FISH analyses, respectively. The accumulation of propionate and subsequent reactor failure after long-term exposure to tylosin are attributed to the direct inhibition of propionate-oxidizing syntrophic bacteria closely related to Syntrophobacter and the indirect inhibition of Methanosaeta by high propionate concentrations and low pH. © 2010 Wiley Periodicals, Inc.

  2. Thermal Properties of Structural Materials Found in Light Water Reactor Vessels

    SciTech Connect

    J. E. Daw; J. L. Rempe; D. L. Knudson

    2009-11-01

    High temperature material property data for structural materials used in existing Light Water Reactors (LWRs) are limited. Often, extrapolated values recommended in the literature differ significantly. To reduce such uncertainties, new data for SA533 Grade B, Class 1 (SA533B1) low alloy steel, Stainless Steel 304 (SS304), and Inconel 600, found in Light Water Reactor (LWR) vessels and penetrations, were acquired and tested using material property systems available at the High Temperature Test Laboratory (HTTL) at the Idaho National Laboratory (INL). Properties measured include thermal expansion, specific heat capacity, and thermal diffusivity for temperatures up to 1200 oC. From these results, thermal conductivity and density were calculated. Results show that, in some cases, previously recommended values for these material differ significantly from measured values at high temperatures. This is especially true for SA533B1, as previous data do not account for the phase transformation of this material between 740 oC and 840 oC.

  3. The effects of low dose rate irradiation and thermal aging on reactor structural alloys

    NASA Astrophysics Data System (ADS)

    Allen, T. R.; Trybus, C. L.; Cole, J. I.

    As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10 -8 dpa/s) irradiation at 380-410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.

  4. Microbial community structure and performance of an anaerobic reactor digesting cassava pulp and pig manure.

    PubMed

    Panichnumsin, P; Ahring, B; Nopharatana, A; Chaiprasert, P

    2012-01-01

    Microbial community dynamics in response to changes in substrate types (i.e. pig manure (PM), cassava pulp (CP) and mixtures of PM and CP) were investigated in an anaerobic continuously stirred tank reactor (CSTR). Molecular identification of bacterial and archaeal domains were performed, using a 16S rDNA clone library with polymerase chain reaction-denaturing gradient gel electrophoresis (PCR-DGGE) screening and phylogenetic analysis. Analysis of bacterial clone libraries revealed that the differences in the community structure corresponded to the substrate types. However, the Bacteroidetes were the most abundant group in all substrates, followed by the Clostridia. With pure PM, the dominant bacterial groups were Bacteroidales, Clostridia and Paludibacter. With a co-substrate, at CP to PM (CP:PM) ratio of 50:50, the sequences analysis revealed the greatest diversity of bacterial communities at class level, and the sequences affiliated with Cytophaga sp. became an exclusive predominant. With CP alone, Bacteroides sp. was the dominant species and this reactor had the lowest diversity of bacteria. Archaea observed in the CSTR fed with all substrate types were Methanosaeta sp., Methanosaeta concilii and Methanospirillum hungatei. Among the Archaea, Methanosaeta sp. was the exclusive predominant. The relative distribution of Archaea also changed regarding to the substrate types.

  5. Backwash intensity and frequency impact the microbial community structure and function in a fixed-bed biofilm reactor.

    PubMed

    Li, Xu; Yuen, Wangki; Morgenroth, Eberhard; Raskin, Lutgarde

    2012-11-01

    Linkages among bioreactor operation and performance and microbial community structure were investigated for a fixed-bed biofilm system designed to remove perchlorate from drinking water. Perchlorate removal was monitored to evaluate reactor performance during and after the frequency and intensity of the backwash procedure were changed, while the microbial community structure was studied using clone libraries and quantitative PCR targeting the 16S rRNA gene. When backwash frequency was increased from once per month to once per day, perchlorate removal initially deteriorated and then recovered, and the relative abundance of perchlorate-reducing bacteria (PRB) initially increased and then decreased. This apparent discrepancy suggested that bacterial populations other than PRB played an indirect role in perchlorate removal, likely by consuming dissolved oxygen, a competing electron acceptor. When backwash intensity was increased, the reactor gradually lost its ability to remove perchlorate, and concurrently the relative abundance of PRB decreased. The results indicated that changes in reactor operation had a profound impact on reactor performance through altering the microbial community structure. Backwashing is an important yet poorly characterized procedure when operating fixed-bed biofilm reactors. Compared to backwash intensity, changes in backwash frequency exerted less disturbance on the microbial community in the current study. If this finding can be confirmed in future work, backwash frequency may serve as the primary parameter when optimizing backwash procedures.

  6. Microbial community structural analysis of an expanded granular sludge bed (EGSB) reactor for beet sugar industrial wastewater (BSIW) treatment.

    PubMed

    Ambuchi, John Justo; Liu, Junfeng; Wang, Haiman; Shan, Lili; Zhou, Xiangtong; Mohammed, Mohammed O A; Feng, Yujie

    2016-05-01

    A looming global energy crisis has directly increased biomethanation processes using anaerobic digestion technology. However, much knowledge on the microbial community structure, their distribution within the digester and related functions remains extremely scanty and unavailable in some cases, yet very valuable in the improvement of the anaerobic bioprocesses. Using pyrosequencing technique based on Miseq PE 3000, microbial community population profiles were determined in an operated mesophilic expanded granular sludge bed (EGSB) reactor treating beet sugar industrial wastewater (BSIW) in the laboratory scale. Further, the distribution of the organisms in the lower, middle and upper sections within the reactor was examined. To our knowledge, this kind of analysis of the microbial community in a reactor treating BSIW is the first of its kind. A total of 44,204 non-chimeric reads with average length beyond 450 bp were yielded. Both bacterial and archaeal communities were identified with archaea predominance (60 %) observed in the middle section. Bayesian classifier yielded 164 families with only 0.73 % sequences which could not be classified to any taxa at family level. The overall phylum predominance in the reactor showed Firmicutes, Euryarchaeota, Chloroflexi, Proteobacteria and Bacteroidetes in the descending order. Our results clearly demonstrate a highly diverse microbial community population of an anaerobic reactor treating BSIW, with distinct distribution levels within the reactor.

  7. ITER (International Thermonuclear Experimental Reactor) reactor building design study

    SciTech Connect

    Thomson, S.L.; Blevins, J.D.; Delisle, M.W.; Canadian Fusion Fuels Technology Project, Mississauga, ON )

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs.

  8. Steam Generator Component Model in a Combined Cycle of Power Conversion Unit for Very High Temperature Gas-Cooled Reactor

    SciTech Connect

    Oh, Chang H; Han, James; Barner, Robert; Sherman, Steven R

    2007-06-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP), Very High Temperature Gas-Cooled Reactor (VHTR) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. A combined cycle is considered as one of the power conversion units to be coupled to the very high-temperature gas-cooled reactor (VHTR). The combined cycle configuration consists of a Brayton top cycle coupled to a Rankine bottoming cycle by means of a steam generator. A detailed sizing and pressure drop model of a steam generator is not available in the HYSYS processes code. Therefore a four region model was developed for implementation into HYSYS. The focus of this study was the validation of a HYSYS steam generator model of two phase flow correlations. The correlations calculated the size and heat exchange of the steam generator. To assess the model, those calculations were input into a RELAP5 model and its results were compared with HYSYS results. The comparison showed many differences in parameters such as the heat transfer coefficients and revealed the different methods used by the codes. Despite differences in approach, the overall results of heat transfer were in good agreement.

  9. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... their normal mode, and need not be vented. Systems that are normally filled with water and operating... returning the reactor to an operating mode requiring containment integrity. For primary reactor containment... surfaces of the containment structures and components shall be performed prior to any Type A test...

  10. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    NASA Technical Reports Server (NTRS)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  11. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    NASA Technical Reports Server (NTRS)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  12. Evaluation of Nb-base alloys for the divertor structure in fusion reactors

    SciTech Connect

    Purdy, I.M.

    1996-04-01

    Niobium-base alloys are candidate materials for the divertor structure in fusion reactors. For this application, an alloy should resist aqueous corrosion, hydrogen embrittlement, and radiation damage and should have high thermal conductivity and low thermal expansion. Results of corrosion and embrittlement screening tests of several binary and ternary Nb alloys in high-temperature water indicated the Mb-1Zr, Nb-5MO-1Zr, and Nb-5V-1Z4 (wt %) showed sufficient promise for further investigation. These alloys, together with pure Nb and Zircaloy-4 have been exposed to high purity water containing a low concentration of dissolved oxygen (<12 ppb) at 170, 230, and 300{degrees}C for up to {approx}3200 h. Weight-change data, microstructural observations, and qualitative mechanical-property evaluation reveal that Nb-5V-1Zr is the most promising alloy at higher temperatures. Below {approx}200{degrees}C, the alloys exhibit similiar corrosion behavior.

  13. An integrated approach to assessing the fracture safe margins of fusion reactor structures

    SciTech Connect

    Odette, G.R.

    1996-10-01

    Design and operation of fusion reactor structures will require an appropriate data base closely coupled to a reliable failure analysis method to safely manage irradiation embrittlement. However, ongoing irradiation programs will not provide the information on embrittlement necessary to accomplish these objectives. A new engineering approach is proposed based on the concept of a master toughness-temperature curve indexed on an absolute temperature scale using shifts to account for variables such as size scales, crack geometry and loading rates as well as embrittlement. While providing a simple practical engineering expedient, the proposed method can also be greatly enhanced by fundamental mechanism based models of fracture and embrittlement. Indeed, such understanding is required for the effective use of small specimen test methods, which is a integral element in developing the necessary data base.

  14. Statistically based uncertainty analysis for ranking of component importance in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor

    SciTech Connect

    Wilson, G.E.

    1992-01-01

    The Analytic Hierarchy Process (AHP) has been used to help determine the importance of components and phenomena in thermal-hydraulic safety analyses of nuclear reactors. The AHP results are based, in part on expert opinion. Therefore, it is prudent to evaluate the uncertainty of the AHP ranks of importance. Prior applications have addressed uncertainty with experimental data comparisons and bounding sensitivity calculations. These methods work well when a sufficient experimental data base exists to justify the comparisons. However, in the case of limited or no experimental data the size of the uncertainty is normally made conservatively large. Accordingly, the author has taken another approach, that of performing a statistically based uncertainty analysis. The new work is based on prior evaluations of the importance of components and phenomena in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor (ANSR), a new facility now in the design phase. The uncertainty during large break loss of coolant, and decay heat removal scenarios is estimated by assigning a probability distribution function (pdf) to the potential error in the initial expert estimates of pair-wise importance between the components. Using a Monte Carlo sampling technique, the error pdfs are propagated through the AHP software solutions to determine a pdf of uncertainty in the system wide importance of each component. To enhance the generality of the results, study of one other problem having different number of elements is reported, as are the effects of a larger assumed pdf error in the expert ranks. Validation of the Monte Carlo sample size and repeatability are also documented.

  15. Potential of direct metal deposition technology for manufacturing thick functionally graded coatings and parts for reactors components

    NASA Astrophysics Data System (ADS)

    Thivillon, L.; Bertrand, Ph.; Laget, B.; Smurov, I.

    2009-03-01

    Direct metal deposition (DMD) is an automated 3D deposition process arising from laser cladding technology with co-axial powder injection to refine or refurbish parts. Recently DMD has been extended to manufacture large-size near-net-shape components. When applied for manufacturing new parts (or their refinement), DMD can provide tailored thermal properties, high corrosion resistance, tailored tribology, multifunctional performance and cost savings due to smart material combinations. In repair (refurbishment) operations, DMD can be applied for parts with a wide variety of geometries and sizes. In contrast to the current tool repair techniques such as tungsten inert gas (TIG), metal inert gas (MIG) and plasma welding, laser cladding technology by DMD offers a well-controlled heat-treated zone due to the high energy density of the laser beam. In addition, this technology may be used for preventative maintenance and design changes/up-grading. One of the advantages of DMD is the possibility to build functionally graded coatings (from 1 mm thickness and higher) and 3D multi-material objects (for example, 100 mm-sized monolithic rectangular) in a single-step manufacturing cycle by using up to 4-channel powder feeder. Approved materials are: Fe (including stainless steel), Ni and Co alloys, (Cu,Ni 10%), WC compounds, TiC compounds. The developed coatings/parts are characterized by low porosity (<1%), fine microstructure, and their microhardness is close to the benchmark value of wrought alloys after thermal treatment (Co-based alloy Stellite, Inox 316L, stainless steel 17-4PH). The intended applications concern cooling elements with complex geometry, friction joints under high temperature and load, light-weight mechanical support structures, hermetic joints, tubes with complex geometry, and tailored inside and outside surface properties, etc.

  16. The component structure of conformal supergravity invariants in six dimensions

    NASA Astrophysics Data System (ADS)

    Butter, Daniel; Novak, Joseph; Tartaglino-Mazzucchelli, Gabriele

    2017-05-01

    In the recent paper arXiv:1606.02921, the two invariant actions for 6D N=(1,0) conformal supergravity were constructed in superspace, corresponding to the supersymmetrization of C 3 and C□ C. In this paper, we provide the translation from superspace to the component formulation of superconformal tensor calculus, and we give the full component actions of these two invariants. As a second application, we build the component form for the supersymmetric F□ F action coupled to conformal supergravity. Exploiting the fact that the N=(2,0) Weyl multiplet has a consistent truncation to N=(1,0), we then verify that there is indeed only a single N=(2,0) conformal supergravity invariant and reconstruct most of its bosonic terms by uplifting a certain linear combination of N=(1,0) invariants.

  17. Reactor safeguards against insider sabotage

    SciTech Connect

    Bennett, H.A.

    1982-03-01

    A conceptual safeguards system is structured to show how both reactor operations and physical protection resources could be integrated to prevent release of radioactive material caused by insider sabotage. Operational recovery capabilities are addressed from the viewpoint of both detection of and response to disabled components. Physical protection capabilities for preventing insider sabotage through the application of work rules are analyzed. Recommendations for further development of safeguards system structures, operational recovery, and sabotage prevention are suggested.

  18. The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

    SciTech Connect

    Pilch, M.; Tarbell, W.W.; Brockmann, J.E.

    1988-09-01

    High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that the debris does not adhere significantly to structures. The lack of retention is attributed to splashing from the surface and reentrainment in the gas flowing over the surface. These processes are shown to be applicable to reactor scale. A third experiment was done to simulate the annular gap between the reactor vessel and cavity wall. Debris collection showed that the fraction of debris exiting through the gap was greater than the gap-to-total flow area ratio. Film records indicate that dispersal was primarily by entrainment of the molten debris in the cavity. 29 refs., 36 figs., 11 tabs.

  19. Hypervelocity Wind Tunnel Components Structural Evaluation. Volume II.

    DTIC Science & Technology

    1979-05-01

    specifications for the manifolds are listed below. a Component Material u Inlet Body -- 143,000 131,000 Exit Body -- 142,000 129,500 Studs ASTM A193 , GRB-7...unlimited I’V* I’.70 STPt ~ A TEEN T Wind Tunnel Components Fatigue [ugh Pressure Crack Propagation See following page. DO 1473 E---IOD o, E DA Is 01s...Vessels in the wind tunnel facility was performed using finite element techniques coupled with fatigue and fracture mechanics analyses of the critical

  20. Reactor Materials Program - Baseline Material Property Handbook - Mechanical Properties of 1950's Vintage Stainless Steel Weldment Components, Task Number 89-23-A-1

    SciTech Connect

    Stoner, K.J.

    1999-11-05

    The Process Water System (primary coolant) piping of the nuclear production reactors constructed in the 1950''s at Savannah River Site is comprised primarily of Type 304 stainless steel with Type 308 stainless steel weld filler. A program to measure the mechanical properties of archival PWS piping and weld materials (having approximately six years of service at temperatures between 25 and 100 degrees C) has been completed. The results from the mechanical testing has been synthesized to provide a mechanical properties database for structural analyses of the SRS piping.

  1. Structure of multi-component/multi-Yukawa mixtures

    NASA Astrophysics Data System (ADS)

    Blum, L.; Arias, M.

    2006-09-01

    Recent small angle scattering experiments reveal new peaks in the structure function S(k) of colloidal systems (Liu et al 2005 J. Chem. Phys. 122 044507), in a region that was inaccessible with older instruments. It has been increasingly evident that a single (or double) Yukawa MSA-closure cannot account for these observations, and three or more terms are needed. On the other hand the MSA is not sufficiently accurate (Broccio et al 2005 Preprint); more accurate theories such as the HNC have been tried. But while the MSA is asymptotically exact at high densities (Rosenfield and Blum 1986 J. Chem. Phys. 85 1556), it does not satisfy the low density asymptotics. This has been corrected in the soft MSA (Blum et al 1972 J. Chem. Phys. 56 5197, Narten et al 1974 J. Chem. Phys. 60 3378) by adding exponential type terms. The results compared to experiment and simulation for liquid sodium by Rahman and Paskin (as shown in Blum et al 1972 J. Chem. Phys. 56 5197) are remarkably good. We use here a general closure of the Ornstein-Zernike equation, which is not necessarily the MSA closure (Blum and Hernando 2001 Condensed Matter Theories vol 16 ed Hernandez and Clark (New York: Nova) p 411). \\begin{equation} \\fl c_{ij}(r)=\\sum_{n=1}^{M}{\\cal{K}}_{ij}^{(n)}\\rme^{-z_{n}r}/r\\tqs {\\cal{K}}_{ij}^{(n)}=K^{(n)}\\delta_{i}^{(n)}\\delta_{j}^{(n)}\\tqs r\\geq \\sigma_{ij} \\label{eq1} \\end{equation} with the boundary condition for gij(r) = 0 for r<=σij. This general closure of the Ornstein-Zernike equation will go well beyond the MSA since it has been tested by Monte Carlo simulation for tetrahedral water (Blum et al 1999 Physica A 265 396), toroidal ion channels (Enriquez and Blum 2005 Mol. Phys. 103 3201) and polyelectrolytes (Blum and Bernard 2004 Proc. Int. School of Physics Enrico Fermi, Course CLV vol 155, ed Mallamace and Stanley (Amsterdam: IOS Press) p 335). For this closure we get for the Laplace transform of the pair correlation function an explicitly symmetric result

  2. Diamond machining of micro-optical components and structures

    NASA Astrophysics Data System (ADS)

    Gläbe, Ralf; Riemer, Oltmann

    2010-05-01

    Diamond machining originates from the 1950s to 1970s in the USA. This technology was originally designed for machining of metal optics at macroscopic dimensions with so far unreached tolerances. During the following decades the machine tools, the monocrystalline diamond cutting tools, the workpiece materials and the machining processes advanced to even higher precision and flexibility. For this reason also the fabrication of small functional components like micro optics at a large spectrum of geometries became technologically and economically feasible. Today, several kinds of fast tool machining and multi axis machining operations can be applied for diamond machining of micro optical components as well as diffractive optical elements. These parts can either be machined directly as single or individual component or as mold insert for mass production by plastic replication. Examples are multi lens arrays, micro mirror arrays and fiber coupling lenses. This paper will give an overview about the potentials and limits of the current diamond machining technology with respect to micro optical components.

  3. Pressure loadings of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) reactor release mitigation structures from large-break LOCAs

    SciTech Connect

    Sienicki, J.J.; Horak, W.C.; Brookhaven National Lab., Upton, NY )

    1989-01-01

    Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs.

  4. Modeling of quantitative effects of water components on the photocatalytic degradation of 17α-ethynylestradiol in a modified flat plate serpentine reactor.

    PubMed

    Wang, Dawei; Li, Yi; Li, Guoping; Wang, Chao; Zhang, Wenlong; Wang, Qing

    2013-06-15

    The effect of water components on the photocatalytic degradation of organic pollutants was incompletely understood, especially in the case of hydroxyl radical (•OH) generation and scavenging. Previous studies have used various methods to determine the rate constants for the reactions between •OH and water components, but the interactions between water components were not taken into concern. In this study, a sequential relative rate technique was used to investigate the effects of water components on the rates of •OH generation and EE2 degradation in a modified flat plate serpentine reactor, including NO₃(-), H₂PO₄(-), SO₄(2-), CO₃(2-), Cl(-), Na(+), Fe(3+), dissolved organic matter (DOM) etc. The results reflected that NO₃(-) and DOM accelerated the photodegradation of 17α-ethynylestradiol (EE2) (3.2% and 21.2%, respectively). Cl(-) and Fe(3+) inhibited that process (5.2% and 3.1%, respectively). Finally, a model for the photocatalytic degradation of EE2 was developed for the first time, taking the obtained rate constants, catalyst concentrations, flow velocities and light intensities into concern. A good agreement was observed between the model and experimental profiles.

  5. Radiological dose assessment for the dismantlement and decommissioning option for the Heavy Water Components Test Reactor facility at the Savannah River Site, Aiken, South Carolina

    SciTech Connect

    Faillace, E.R.; Kamboj, S.; Yu, C.; Chen, S.Y.

    1997-10-01

    Potential maximum radiation dose rates for a 10,000-year horizon were calculated for the dismantlement and decommissioning option for the Heavy Water Components Test Reactor facility at the Savannah River Site. The residual radioactive material guidelines (RESRAD) computer code was used. The study will help determine if it is acceptable (in terms of DOE radiation dose limits) for activated and contaminated concrete to remain in the facility, along with embedded radioactive piping and radioactive equipment. Four cases were developed to evaluate potential doses; the cases vary with regard to the definitions of the sources. Case A considers the dose from the reactor biological shield; case B considers the dose from contaminated concrete rubble; case C considers the dose from contaminated concrete rubble, the reactor biological shield, and installed equipment; and case D considers the dose from contaminated cuttings brought to the surface following the perforation of a well through the contaminated zone in case C. Site-specific parameter values were used to estimate the radiation doses. The results indicate that neither the DOE dose limit of 100 mrem/yr nor the 15-mrem/yr dose constraint would be exceeded for any of the cases. The potential maximum dose rates for cases A, B, C, and D are 0.000028, 0.015, 0.018, and 0.17 mrem/yr, respectively. The drinking water pathway is the dominant contributor to the doses in cases A through C, and the external gamma pathway is the dominant contributor in case D. Carbon-14, uranium-234, uranium-238, and americium-241 are the principal radionuclides contributing to the doses in cases A through C. Cobalt-60, europium-152, and barium-133 are the important radionuclides in case D. A sensitivity analysis was performed to determine which parameters have the greatest impact on the estimated doses. 9 refs., 11 figs., 3 tabs.

  6. 78 FR 48727 - Proposed Revisions to Design of Structures, Components, Equipment and Systems

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-09

    ... Core Support Structures.'' DATES: Comments must be filed no later than September 9, 2013. Comments... COMMISSION Proposed Revisions to Design of Structures, Components, Equipment and Systems AGENCY: Nuclear... Chapter 3, ``Design of Structures, Components, Equipment, and Systems'' and soliciting public comment...

  7. Peak earthquake response of structures under multi-component excitations

    NASA Astrophysics Data System (ADS)

    Song, Jianwei; Liang, Zach; Chu, Yi-Lun; Lee, George C.

    2007-12-01

    Accurate estimation of the peak seismic responses of structures is important in earthquake resistant design. The internal force distributions and the seismic responses of structures are quite complex, since ground motions are multi-directional. One key issue is the uncertainty of the incident angle between the directions of ground motion and the reference axes of the structure. Different assumed seismic incidences can result in different peak values within the scope of design spectrum analysis for a given structure and earthquake ground motion record combination. Using time history analysis to determine the maximum structural responses excited by a given earthquake record requires repetitive calculations to determine the critical incident angle. This paper presents a transformation approach for relatively accurate and rapid determination of the maximum peak responses of a linear structure subjected to three-dimensional excitations within all possible seismic incident angles. The responses can be deformations, internal forces, strains and so on. An irregular building structure model is established using SAP2000 program. Several typical earthquake records and an artificial white noise are applied to the structure model to illustrate the variation of the maximum structural responses for different incident angles. Numerical results show that for many structural parameters, the variation can be greater than 100%. This method can be directly applied to time history analysis of structures using existing computer software to determine the peak responses without carrying out the analyses for all possible incident angles. It can also be used to verify and/or modify aseismic designs by using response spectrum analysis.

  8. Effect of phenol on the nitrogen removal performance and microbial community structure and composition of an anammox reactor.

    PubMed

    Pereira, Alyne Duarte; Leal, Cíntia Dutra; Dias, Marcela França; Etchebehere, Claudia; Chernicharo, Carlos Augusto L; de Araújo, Juliana Calabria

    2014-08-01

    The effects of phenol on the nitrogen removal performance of a sequencing batch reactor (SBR) with anammox activity and on the microbial community within the reactor were evaluated. A phenol concentration of 300 mg L(-1) reduced the ammonium-nitrogen removal efficiency of the SBR from 96.5% to 47%. The addition of phenol changed the microbial community structure and composition considerably, as shown by denaturing gradient gel electrophoresis and 454 pyrosequencing of 16S rRNA genes. Some phyla, such as Proteobacteria, Verrucomicrobia, and Firmicutes, increased in abundance, whereas others, such as Acidobacteria, Chloroflexi, Planctomycetes, GN04, WS3, and NKB19, decreased. The diversity of the anammox bacteria was also affected by phenol: sequences related to Candidatus Brocadia fulgida were no longer detected, whereas sequences related to Ca. Brocadia sp. 40 and Ca. Jettenia asiatica persisted. These results indicate that phenol adversely affects anammox metabolism and changes the bacterial community within the anammox reactor.

  9. Joining of Components of Complex Structures for Improved Dynamic Response

    DTIC Science & Technology

    2011-10-28

    BOX STRUCTURE WITH THICK- NESS VARIATIONS To demonstrate the improved/optimal joining, a structure with a V-shaped bot - tom is considered, as shown...Rozvany, The COC algorithm, Part II: Topological, geo- metrical and generalized shape optimization, Computer Methods in Applied Mechanics and Engineering

  10. Composite load spectra for select space propulsion structural components

    NASA Technical Reports Server (NTRS)

    Newell, J. F.

    1987-01-01

    The objective of this program is to develop generic load models to simulate the composite load spectra (CLS) that are induced in space propulsion system components representative of the space shuttle main engines (SSME). These models are being developed through describing individual component loads with an appropriate mix of deterministic and state-of-the-art probabilistic models that are related to key generic variables. Combinations of the individual loads are used to synthesize the composite loads spectra. A second approach for developing the composite loads spectra load model simulation, the option portion of the contract will develop coupled models which combine the individual load models. Statistically varying coefficients of the physical models will be used to obtain the composite load spectra.

  11. STRUCTURAL DESIGN CRITERIA FOR TARGET/BLANKET SYSTEM COMPONENT MATERIALS FOR THE ACCELERATOR PRODUCTION OF TRITIUM PROJECT

    SciTech Connect

    W. JOHNSON; R. RYDER; P. RITTENHOUSE

    2001-01-01

    The design of target/blanket system components for the Accelerator Production of Tritium (APT) plant is dependent on the development of materials properties data specified by the designer. These data are needed to verify that component designs are adequate. The adequacy of the data will be related to safety, performance, and economic considerations, and to other requirements that may be deemed necessary by customers and regulatory bodies. The data required may already be in existence, as in the open technical literature, or may need to be generated, as is often the case for the design of new systems operating under relatively unique conditions. The designers' starting point for design data needs is generally some form of design criteria used in conjunction with a specified set of loading conditions and associated performance requirements. Most criteria are aimed at verifying the structural adequacy of the component, and often take the form of national or international standards such as the ASME Boiler and Pressure Vessel Code (ASME B and PV Code) or the French Nuclear Structural Requirements (RCC-MR). Whether or not there are specific design data needs associated with the use of these design criteria will largely depend on the uniqueness of the conditions of operation of the component. A component designed in accordance with the ASME B and PV Code, where no unusual environmental conditions exist, will utilize well-documented, statistically-evaluated developed in conjunction with the Code, and will not be likely to have any design data needs. On the other hand, a component to be designed to operate under unique APT conditions, is likely to have significant design data needs. Such a component is also likely to require special design criteria for verification of its structural adequacy, specifically accounting for changes in materials properties which may occur during exposure in the service environment. In such a situation it is common for the design criteria and

  12. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    DOEpatents

    Travelli, A.

    1985-10-25

    A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

  13. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    DOEpatents

    Travelli, Armando

    1988-01-01

    A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

  14. International Atomic Energy Agency (IAEA) Coordinated Research Projects on Structural Integrity of Reactor Pressure Vessels

    SciTech Connect

    Server, W. L.; Nanstad, Randy K

    2009-01-01

    The International Atomic Energy Agency (IAEA) has conducted a series of Coordinated Research Projects (CRPs) that have focused on irradiated reactor pressure vessel (RPV) steel fracture toughness properties and approaches for assuring structural integrity of RPVs throughout operating life. A series of nine CRPs have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on RPV steels. The purpose of the CRPs was to develop comparisons and correlations to test the uniformity of irradiated results through coordinated international research studies and data sharing. Consideration of dose rate effects, effects of alloying (nickel, manganese, silicon, etc.) and residual elements (eg., copper and phosphorus), and drop in upper shelf toughness are also important for assessing neutron embrittlement effects. The ultimate use of embrittlement understanding is assuring structural integrity of the RPV under current and future operation and accident conditions. Material fracture toughness is the key ingredient needed for this assessment, and many of the CRPs have focused on measurement and application of irradiated fracture toughness. This paper presents an overview of the progress made since the inception of the CRPs in the early 1970s. The chronology and importance of each CRP have been reviewed and put into context for continued and long-term safe operation of RPVs.

  15. Investigations on the fatigue behavior of high-temperature alloys for high-temperature gas-cooled reactor components

    SciTech Connect

    Meurer, H.P.; Gnirss, G.K.H.; Mergler, W.; Raule, G.; Schuster, H.; Ullrich, G.

    1984-08-01

    For the development of a high-temperature gascooled reactor that is to be operated at temperatures up to 950/sup 0/C, the low-cycle fatigue (LCF) as well as the high-cycle fatigue (HCF) behavior of several hightemperature alloys have been evaluated. The tests, performed between room temperature and 950/sup 0/C, include the influence of the environment, hold times, and strain rate in the case of LCF behavior and of mean stresses in the case of HCF behavior. At high strain ranges, alloys with a high ductility like Incoloy-800H appear to be superior, whereas at low strain ranges and under HCF conditions, high-strength alloys like Inconel-617 and Nimonic-86 show a better fatigue resistance. Hold times decrease LCF resistance, especially at low strain ranges, which can be explained by the large stress relaxation. The better LCF resistance in impure helium compared to tests in air was correlated to differences in the deformation and crack initiation mechanisms. At high temperatures, strain rate plays an important role for the stress response under LCF loading. The HCF behavior was found to be very sensitive to superimposed mean stresses because of the considerable creep strain induced.

  16. Probabilistic structural analysis of aerospace components using NESSUS

    NASA Technical Reports Server (NTRS)

    Shiao, Michael C.; Nagpal, Vinod K.; Chamis, Christos C.

    1988-01-01

    Probabilistic structural analysis of a Space Shuttle main engine turbopump blade is conducted using the computer code NESSUS (numerical evaluation of stochastic structures under stress). The goal of the analysis is to derive probabilistic characteristics of blade response given probabilistic descriptions of uncertainties in blade geometry, material properties, and temperature and pressure distributions. Probability densities are derived for critical blade responses. Risk assessment and failure life analysis is conducted assuming different failure models.

  17. The structure of galaxies : the division of stellar mass by morphological type and structural component

    NASA Astrophysics Data System (ADS)

    Kelvin, Lee Steven

    This thesis explores the relation between galaxy structure, morphology and stellar mass. In the first part I present single-Sersic two-dimensional model fits to 167,600 galaxies modelled independently in the ugrizYJHK bandpasses using reprocessed Sloan Digital Sky Survey Data Release Seven (SDSS DR7) and UKIRT Infrared Deep Sky Survey Large Area Survey (UKIDSS LAS) imaging data available via the Galaxy and Mass Assembly (GAMA) data base. In order to facilitate this study, we developed Structural Investigation of Galaxies via Model Analysis (SIGMA): an automated wrapper around several contemporary astronomy software packages. We confirm that variations in global structural measurements with wavelength arise due to the effects of dust attenuation and stellar population/metallicity gradients within galaxies. In the second part of this thesis we establish a volume-limited sample of 3,845 galaxies in the local Universe and visually classify these galaxies according to their morphological Hubble type. We find that single-Sersic photometry accurately reproduces the morphology luminosity functions predicted in the literature. We employ multi-component Sersic profiling to provide bulge-disk decompositions for this sample, allowing for the luminosity and stellar mass to be divided between the key structural components: spheroids and disks. Grouping the stellar mass in these structures by the evolutionary mechanisms that formed them, we find that hot-mode collapse, merger or otherwise turbulent mechanisms account for ~46% of the total stellar mass budget, cold-mode gas accretion and splashback mechanisms account for ~48% of the total stellar mass budget and secular evolutionary processes for ~6.5% of the total stellar mass budget in the local (z<0.06) Universe.

  18. COBRA-NC: a thermal hydraulics code for transient analysis of nuclear reactor components. Volume 4. Users' manual for containment analysis

    SciTech Connect

    Wheeler, C.L.; Thurgood, M.J.; Guidotti, T.E.; DeBellis, D.E.

    1986-08-01

    COBRA-NC is a digital computer program written in FORTRAN IV that simulates the response of nuclear reactor components and systems to thermal-hydraulic transients. The code solves the multicomponent, compressible, three-dimensional, two-fluid, three-field equations for two-phase flow. The three velocity fields are the vapor/gas field, the continuous liquid field, and the liquid drop field. This volume of the manual provides the user with an explanation of the input required to simulate the response of multicompartment nuclear containment systems to postulated loss-of-coolant accidents that result in the release of steam, water, and/or noncondensable gases into the containment.

  19. Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

    SciTech Connect

    Klueh, R.L.; Bloom, E.E.

    1981-12-01

    Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed.

  20. Structural modeling for control design (articulated multibody component representation)

    NASA Technical Reports Server (NTRS)

    Haugse, E. D.; Jones, R. E.; Salus, W. L.

    1989-01-01

    High gain, high frequency flexible responses in gimbaled multibody systems are discussed. Their origin and physical significance are described in terms of detailed mass and stiffness modeling at actuator/sensor interfaces. Guyan Reduction, Generalized Dynamic Reduction, inadequate mass modeling detail, as well as system mode truncation, are shown to suppress the high gain high frequency response and thereby lose system flexibility important for stability and performance predictions. Model validation by modal survey testing is shown to risk similar loss of accuracy. Difficulties caused by high frequency responses in component mode simulations, such as DISCOS, and also linearized system mode simulations, are described, and approaches for handling these difficulties are discussed.

  1. Detection of bondline delaminations in multilayer structures with lossy components

    NASA Technical Reports Server (NTRS)

    Madaras, Eric I.; Winfree, William P.; Smith, B. T.; Heyman, Joseph H.

    1988-01-01

    The detection of bondline delaminations in multilayer structures using ultrasonic reflection techniques is a generic problem in adhesively bonded composite structures such as the Space Shuttles's Solid Rocket Motors (SRM). Standard pulse echo ultrasonic techniques do not perform well for a composite resonator composed of a resonant layer combined with attenuating layers. Excessive ringing in the resonant layer tends to mask internal echoes emanating from the attenuating layers. The SRM is made up of a resonant steel layer backed by layers of adhesive, rubber, liner and fuel, which are ultrasonically attenuating. The structure's response is modeled as a lossy ultrasonic transmission line. The model predicts that the acoustic response of the system is sensitive to delaminations at the interior bondlines in a few narrow frequency bands. These predictions are verified by measurements on a fabricated system. Successful imaging of internal delaminations is sensitive to proper selection of the interrogating frequency. Images of fabricated bondline delaminations are presented based on these studies.

  2. Release strategies for making transferable semiconductor structures, devices and device components

    SciTech Connect

    Rogers, John A.; Nuzzo, Ralph G.; Meitl, Matthew; Ko, Heung Cho; Yoon, Jongseung; Menard, Etienne; Baca, Alfred J.

    2016-05-24

    Provided are methods for making a device or device component by providing a multi layer structure having a plurality of functional layers and a plurality of release layers and releasing the functional layers from the multilayer structure by separating one or more of the release layers to generate a plurality of transferable structures. The transferable structures are printed onto a device substrate or device component supported by a device substrate. The methods and systems provide means for making high-quality and low-cost photovoltaic devices, transferable semiconductor structures, (opto-)electronic devices and device components.

  3. Release strategies for making transferable semiconductor structures, devices and device components

    DOEpatents

    Rogers, John A; Nuzzo, Ralph G; Meitl, Matthew; Ko, Heung Cho; Yoon, Jongseung; Menard, Etienne; Baca, Alfred J

    2014-11-25

    Provided are methods for making a device or device component by providing a multilayer structure having a plurality of functional layers and a plurality of release layers and releasing the functional layers from the multilayer structure by separating one or more of the release layers to generate a plurality of transferable structures. The transferable structures are printed onto a device substrate or device component supported by a device substrate. The methods and systems provide means for making high-quality and low-cost photovoltaic devices, transferable semiconductor structures, (opto-)electronic devices and device components.

  4. Release strategies for making transferable semiconductor structures, devices and device components

    DOEpatents

    Rogers, John A.; Nuzzo, Ralph G.; Meitl, Matthew; Ko, Heung Cho; Yoon, Jongseung; Menard, Etienne; Baca, Alfred J.

    2011-04-26

    Provided are methods for making a device or device component by providing a multilayer structure having a plurality of functional layers and a plurality of release layers and releasing the functional layers from the multilayer structure by separating one or more of the release layers to generate a plurality of transferable structures. The transferable structures are printed onto a device substrate or device component supported by a device substrate. The methods and systems provide means for making high-quality and low-cost photovoltaic devices, transferable semiconductor structures, (opto-)electronic devices and device components.

  5. 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.

  6. Composite load spectra for select space propulsion structural components

    NASA Technical Reports Server (NTRS)

    Newell, J. F.; Kurth, R. E.; Ho, H.

    1991-01-01

    The objective of this program is to develop generic load models with multiple levels of progressive sophistication to simulate the composite (combined) load spectra that are induced in space propulsion system components, representative of Space Shuttle Main Engines (SSME), such as transfer ducts, turbine blades, and liquid oxygen posts and system ducting. The first approach will consist of using state of the art probabilistic methods to describe the individual loading conditions and combinations of these loading conditions to synthesize the composite load spectra simulation. The second approach will consist of developing coupled models for composite load spectra simulation which combine the deterministic models for composite load dynamic, acoustic, high pressure, and high rotational speed, etc., load simulation using statistically varying coefficients. These coefficients will then be determined using advanced probabilistic simulation methods with and without strategically selected experimental data.

  7. Composite load spectra for select space propulsion structural components

    NASA Technical Reports Server (NTRS)

    Newell, J. F.; Kurth, R. E.; Ho, H.

    1991-01-01

    The objective of this program is to develop generic load models with multiple levels of progressive sophistication to simulate the composite load spectra that are induced in space propulsion system components, representative of Space Shuttle Main Engines (SSME), such as transfer ducts, turbine blades, and liquid oxygen (LOX) posts and system ducting. These models will be developed using two independent approaches. The first approach consists of using state-of-the-art probabilistic methods to describe the individual loading conditions and combinations of these loading conditions to synthesize the composite load spectra simulation. The methodology required to combine the various individual load simulation models (hot-gas dynamic, vibrations, instantaneous position, centrifugal field, etc.) into composite load spectra simulation models will be developed under this program. A computer code incorporating the various individual and composite load spectra models will be developed to construct the specific load model desired. The second approach, which is covered under the options portion of the contract, will consist of developing coupled models for composite load spectra simulation which combine the (deterministic) models for composite load dynamic, acoustic, high-pressure and high rotational speed, etc., load simulation using statistically varying coefficients. These coefficients will then be determined using advanced probabilistic simulation methods with and without strategically selected experimental data. This report covers the efforts of the third year of the contract. The overall program status is that the turbine blade loads have been completed and implemented. The transfer duct loads are defined and are being implemented. The thermal loads for all components are defined and coding is being developed. A dynamic pressure load model is under development. The parallel work on the probabilistic methodology is essentially completed. The overall effort is being

  8. An Ongoing Role for Structural Sarcomeric Components in Maintaining Drosophila melanogaster Muscle Function and Structure

    PubMed Central

    Perkins, Alexander D.; Tanentzapf, Guy

    2014-01-01

    Animal muscles must maintain their function while bearing substantial mechanical loads. How muscles withstand persistent mechanical strain is presently not well understood. The basic unit of muscle is the sarcomere, which is primarily composed of cytoskeletal proteins. We hypothesized that cytoskeletal protein turnover is required to maintain muscle function. Using the flight muscles of Drosophila melanogaster, we confirmed that the sarcomeric cytoskeleton undergoes turnover throughout adult life. To uncover which cytoskeletal components are required to maintain adult muscle function, we performed an RNAi-mediated knockdown screen targeting the entire fly cytoskeleton and associated proteins. Gene knockdown was restricted to adult flies and muscle function was analyzed with behavioural assays. Here we analyze the results of that screen and characterize the specific muscle maintenance role for several hits. The screen identified 46 genes required for muscle maintenance: 40 of which had no previously known role in this process. Bioinformatic analysis highlighted the structural sarcomeric proteins as a candidate group for further analysis. Detailed confocal and electron microscopic analysis showed that while muscle architecture was maintained after candidate gene knockdown, sarcomere length was disrupted. Specifically, we found that ongoing synthesis and turnover of the key sarcomere structural components Projectin, Myosin and Actin are required to maintain correct sarcomere length and thin filament length. Our results provide in vivo evidence of adult muscle protein turnover and uncover specific functional defects associated with reduced expression of a subset of cytoskeletal proteins in the adult animal. PMID:24915196

  9. Design-Load Basis for LANL Structures, Systems, and Components

    SciTech Connect

    I. Cuesta

    2004-09-01

    This document supports the recommendations in the Los Alamos National Laboratory (LANL) Engineering Standard Manual (ESM), Chapter 5--Structural providing the basis for the loads, analysis procedures, and codes to be used in the ESM. It also provides the justification for eliminating the loads to be considered in design, and evidence that the design basis loads are appropriate and consistent with the graded approach required by the Department of Energy (DOE) Code of Federal Regulation Nuclear Safety Management, 10, Part 830. This document focuses on (1) the primary and secondary natural phenomena hazards listed in DOE-G-420.1-2, Appendix C, (2) additional loads not related to natural phenomena hazards, and (3) the design loads on structures during construction.

  10. Optimal glass-ceramic structures: Components of giant mirror telescopes

    NASA Technical Reports Server (NTRS)

    Eschenauer, Hans A.

    1990-01-01

    Detailed investigations are carried out on optimal glass-ceramic mirror structures of terrestrial space technology (optical telescopes). In order to find an optimum design, a nonlinear multi-criteria optimization problem is formulated. 'Minimum deformation' at 'minimum weight' are selected as contradictory objectives, and a set of further constraints (quilting effect, optical faults etc.) is defined and included. A special result of the investigations is described.

  11. The Effects of Polyunsaturated Lipid Components on bilayer Structure

    NASA Astrophysics Data System (ADS)

    Pramudya, Y.; Kiss, A.; Nguyen, Lam T.; Yuan, J.; Hirst, Linda S.

    2007-03-01

    Polyunsaturated fatty acids (PUFAs), such as DHA (Docosahexanoic Acid) and AA (Alphalinoleic Acid) have been the focus of much research attention in recent years, due to their apparent health benefits and effects on cell physiology. They are found in a variety of biological membranes and have been implicated with lipid raft formation and possible function, particularly in the retinal rod cells and the central nervous system. In this work lipid bilayer structure has been investigated in lipid mixtures, incorporating polyunsaturated fatty acid moieties. The structural effects of increasing concentrations of both symmetric and asymmetric PUFA materials on the bilayer structure are investigated via synchrotron x-ray diffraction on solution samples. We observe bilayer spacings to increase with the percentage of unsaturated fatty acid lipid in the membrane, whilst the degree of ordering significantly decreases. In fact above 20% of fatty acid, well defined bilayers are no longer observed to form. Evidence of phase separation can be clearly seen from these x-ray results and in combination with AFM measurements.

  12. Design component method for sensitivity analysis of built-up structures

    NASA Technical Reports Server (NTRS)

    Choi, Kyung K.; Seong, Hwai G.

    1986-01-01

    A 'design component method' that provides a unified and systematic organization of design sensitivity analysis for built-up structures is developed and implemented. Both conventional design variables, such as thickness and cross-sectional area, and shape design variables of components of built-up structures are considered. It is shown that design of components of built-up structures can be characterized and system design sensitivity expressions obtained by simply adding contributions from each component. The method leads to a systematic organization of computations for design sensitivity analysis that is similar to the way in which computations are organized within a finite element code.

  13. Seismic Soil-Structure Interaction Analyses of a Deeply Embedded Model Reactor – SASSI Analyses

    SciTech Connect

    Nie J.; Braverman J.; Costantino, M.

    2013-10-31

    This report summarizes the SASSI analyses of a deeply embedded reactor model performed by BNL and CJC and Associates, as part of the seismic soil-structure interaction (SSI) simulation capability project for the NEAMS (Nuclear Energy Advanced Modeling and Simulation) Program of the Department of Energy. The SASSI analyses included three cases: 0.2 g, 0.5 g, and 0.9g, all of which refer to nominal peak accelerations at the top of the bedrock. The analyses utilized the modified subtraction method (MSM) for performing the seismic SSI evaluations. Each case consisted of two analyses: input motion in one horizontal direction (X) and input motion in the vertical direction (Z), both of which utilized the same in-column input motion. Besides providing SASSI results for use in comparison with the time domain SSI results obtained using the DIABLO computer code, this study also leads to the recognition that the frequency-domain method should be modernized so that it can better serve its mission-critical role for analysis and design of nuclear power plants.

  14. Studies of low temperature, low flux radiation embrittlement of nuclear reactor structural materials. Final report

    SciTech Connect

    Odette, G.R.; Lucas, G.E.

    1998-09-02

    A large matrix of simple alloys and complex commercial type steels was irradiated over a range of fluxes at 60 C up to a fast fluence of about 3 {times} 10{sup 22} n/m{sup 2}. Combined with data in the literature, these results show a negligible effect of flux on irradiation hardening in the range of 2 {times} 10{sup 13} to 5 {times} 10{sup 18} n/m{sup 2}-s. This observation lends indirect support to the proposal that the accelerated embrittlement in the High Flux Isotope Reactor surveillance steels was due to an anomalously high level of damage from gamma rays. A weak dependence of hardening on a number of elements, including copper, nickel, phosphorus, molybdenum and manganese, can be described by a simple empirical chemistry factor. Particular combinations of elements resulted in hardening differences of up to about 60% in the complex commercial type steels and up to about 100% in simple model alloys. Direct effects of microstructure appear to be minimal. Hardening varies with the square root of fluence above a threshold around 4 {times} 10{sup 20} n/m{sup 2}. The results suggest that low temperature hardening is dominated by local intracascade processes leading to the formation of small defect-solute clusters/complexes. The observed hardening corresponds to nominal maximum end-of-life transition temperature shifts in support structure steels of about 120 C.

  15. Different substrates and starter inocula govern microbial community structures in biogas reactors.

    PubMed

    Satpathy, Preseela; Steinigeweg, Sven; Cypionka, Heribert; Engelen, Bert

    2016-01-01

    The influence of different starter inocula on the microbial communities in biogas batch reactors fed with fresh maize and maize silage as substrates was investigated. Molecular biological analysis by Denaturing Gradient Gel Electrophoresis (DGGE) of 16S rRNA gene fragments showed that each inoculum bore specific microbial communities with varying predominant phylotypes. Both, bacterial and archaeal DGGE profiles displayed three distinct communities that developed depending on the type of inoculum. Although maize and silage are similar substrates, different communities dominated the lactate-rich silage compared to lactate-free fresh maize. Cluster analysis of DGGE gels showed the communities of the same substrates to be stable with their respective inoculum. Bacteria-specific DGGE analysis revealed a rich diversity with Firmicutes being predominant. The other abundant phylotypes were Bacteroidetes and Synergistetes. Archaea-specific DGGE analysis displayed less diverse community structures, identifying members of the Methanosarcinales as the dominant methanogens present in all the three biogas digesters. In general, the source of inoculum played a significant role in shaping microbial communities. Adaptability of the inoculum to the substrates fed also influenced community compositions which further impacted the rates of biogas production.

  16. A miniaturized test method for the mechanical characterization of structural materials for fusion reactors

    NASA Astrophysics Data System (ADS)

    Gondi, P.; Donato, A.; Montanari, R.; Sili, A.

    1996-10-01

    This work deals with a non-destructive method for mechanical tests which is based on the indentation of materials at a constant rate by means of a cylinder with a small radius and penetrating flat surface. The load versus penetration depth curves obtained using this method have shown correspondences with those of tensile tests and have given indications about the mechanical properties on a reduced scale. In this work penetration tests have been carried out on various kinds of Cr martensitic steels (MANET-2, BATMAN and modified F82H) which are of interest for first wall and structural applications in future fusion reactors. The load versus penetration depth curves have been examined with reference to data obtained in tensile tests and to microhardness measurements. Penetration tests have been performed at various temperature (from -180 to 100°C). Conclusions, which can be drawn for the ductile to brittle transition, are discussed for MANET-2 steel. Preliminary results obtained on BATMAN and modified F82H steels are reported. The characteristics of the indenter imprints have been studied by scanning electron microscopy.

  17. Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

    SciTech Connect

    Koshelev, A. S. Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

    2016-12-15

    The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

  18. Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

    NASA Astrophysics Data System (ADS)

    Koshelev, A. S.; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

    2016-12-01

    The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

  19. 78 FR 15755 - Proposed Revision to Design of Structures, Components, Equipment and Systems; Correction

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-03-12

    ... COMMISSION Proposed Revision to Design of Structures, Components, Equipment and Systems; Correction AGENCY... Chapter 3, ``Design of Structures, Components, Equipment, and Systems'' and is soliciting public comment on NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power...

  20. Application of Seismic Design Requirements to Cold Vacuum Drying (CVD) Facility Structures and Systems and Components

    SciTech Connect

    CREA, B.A.

    1999-11-15

    The methodology followed in assignment of Performance Class (PC) for Natural Phenomena Hazards (NPH) seismic loads for Cold Vacuum Drying Facility (CVDF) Structures, Systems and Components is defined. The loading definition associated with each PC and structure, system and component is then defined.

  1. Nuclear reactor

    DOEpatents

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  2. Design procedures for fiber composite structural components: Rods, columns and beam columns

    NASA Technical Reports Server (NTRS)

    Chamis, C. C.

    1983-01-01

    Step by step procedures are described which are used to design structural components (rods, columns, and beam columns) subjected to steady state mechanical loads and hydrothermal environments. Illustrative examples are presented for structural components designed for static tensile and compressive loads, and fatigue as well as for moisture and temperature effects. Each example is set up as a sample design illustrating the detailed steps that are used to design similar components.

  3. Composite load spectra for select space propulsion structural components

    NASA Technical Reports Server (NTRS)

    Newell, J. F.; Kurth, R. E.; Ho, H.

    1986-01-01

    A multiyear program is performed with the objective to develop generic load models with multiple levels of progressive sophistication to simulate the composite (combined) load spectra that are induced in space propulsion system components, representative of Space Shuttle Main Engines (SSME), such as transfer ducts, turbine blades, and liquid oxygen (LOX) posts. Progress of the first year's effort includes completion of a sufficient portion of each task -- probabilistic models, code development, validation, and an initial operational code. This code has from its inception an expert system philosophy that could be added to throughout the program and in the future. The initial operational code is only applicable to turbine blade type loadings. The probabilistic model included in the operational code has fitting routines for loads that utilize a modified Discrete Probabilistic Distribution termed RASCAL, a barrier crossing method and a Monte Carlo method. An initial load model was developed by Battelle that is currently used for the slowly varying duty cycle type loading. The intent is to use the model and related codes essentially in the current form for all loads that are based on measured or calculated data that have followed a slowly varying profile.

  4. Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

    SciTech Connect

    Wang, C.Y.

    1993-06-01

    This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts` ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

  5. Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

    SciTech Connect

    Wang, C.Y.

    1993-01-01

    This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts' ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

  6. Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

    NASA Astrophysics Data System (ADS)

    Wang, C. Y.

    This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts' ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

  7. Flight-service evaluation of composite structural components

    NASA Technical Reports Server (NTRS)

    Dexter, H. B.

    1973-01-01

    A review of programs aimed at flight-service evaluation of composite materials in various applications is presented. These flight-service programs are expected to continue for up to 5 years and include selective reinforcement of an airplane center wing box a helicopter tail cone, and composite replacements for commercial aircraft spoilers and fairings. These longtime flight-service programs will help provide the necessary information required by commercial airlines to commit advanced composites to aircraft structures with confidence. Results of these programs will provide information concerning the stability of composite materials when subjected to various flight environments.

  8. Probabilistic Structural Analysis Methods (PSAM) for select space propulsion system components, part 2

    NASA Technical Reports Server (NTRS)

    1991-01-01

    The technical effort and computer code enhancements performed during the sixth year of the Probabilistic Structural Analysis Methods program are summarized. Various capabilities are described to probabilistically combine structural response and structural resistance to compute component reliability. A library of structural resistance models is implemented in the Numerical Evaluations of Stochastic Structures Under Stress (NESSUS) code that included fatigue, fracture, creep, multi-factor interaction, and other important effects. In addition, a user interface was developed for user-defined resistance models. An accurate and efficient reliability method was developed and was successfully implemented in the NESSUS code to compute component reliability based on user-selected response and resistance models. A risk module was developed to compute component risk with respect to cost, performance, or user-defined criteria. The new component risk assessment capabilities were validated and demonstrated using several examples. Various supporting methodologies were also developed in support of component risk assessment.

  9. [Innovation and practice of component structure theory on material basis of traditional Chinese medicine prescriptions].

    PubMed

    Feng, Liang; Zhang, Ming-Hu; Gu, Jun-Feil; Wu, Chan; Jia, Xiao-Bin

    2013-11-01

    The component structure theory on material basis of traditional Chinese medicine prescriptions provides a new research thought and method for studies on traditional Chinese medicine prescriptions in line with integrated and systemic characteristics of traditional Chinese medicine. Through years of exploration and accumulation, studies on component structures have made achievements. On the basis of summarizing the component structure development of material basis of traditional Chinese medicine prescriptions, we systematically explained the background of component structures and their roles and progress in quality control of traditional Chinese medicine prescriptions and modern innovative traditional Chinese medicine preparations. Studies on component structures promote the changes in material basis of traditional Chinese medicine prescriptions, and point out the direction for the modernization development of traditional Chinese medicine prescriptions.

  10. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1961-01-24

    A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

  11. Electron microscopic examination of wastewater biofilm formation and structural components.

    PubMed Central

    Eighmy, T T; Maratea, D; Bishop, P L

    1983-01-01

    This research documents in situ wastewater biofilm formation, structure, and physiochemical properties as revealed by scanning and transmission electron microscopy. Cationized ferritin was used to label anionic sites of the biofilm glycocalyx for viewing in thin section. Wastewater biofilm formation paralleled the processes involved in marine biofilm formation. Scanning electron microscopy revealed a dramatic increase in cell colonization and growth over a 144-h period. Constituents included a variety of actively dividing morphological types. Many of the colonizing bacteria were flagellated. Filaments were seen after primary colonization of the surface. Transmission electron microscopy revealed a dominant gram-negative cell wall structure in the biofilm constituents. At least three types of glycocalyces were observed. The predominant glycocalyx possessed interstices and was densely labeled with cationized ferritin. Two of the glycocalyces appeared to mediate biofilm adhesion to the substratum. The results suggest that the predominant glycocalyx of this thin wastewater biofilm serves, in part, to: (i) enclose the bacteria in a matrix and anchor the biofilm to the substratum and (ii) provide an extensive surface area with polyanionic properties. Images PMID:6881965

  12. Effects of non-structural components and soil-structure interaction on the seismic response of framed structures

    NASA Astrophysics Data System (ADS)

    Ditommaso, Rocco; Auletta, Gianluca; Iacovino, Chiara; Nigro, Antonella; Carlo Ponzo, Felice

    2017-04-01

    In this paper, several nonlinear numerical models of reinforced concrete framed structures have been defined in order to evaluate the effects of non-structural elements and soil-structure interaction on the elastic dynamic behaviour of buildings. In the last few years, many and various studies have highlighted the significant effects derived from the interaction between structural and non-structural components on the main dynamic characteristics of a building. Usually, structural and non-structural elements act together, adding both masses and stiffness. The presence of infill panels is generally neglected in the design process of structural elements, although these elements can significantly increase the lateral stiffness of a structure leading to a modification in the dynamic properties. Particularly, at the Damage Limit State (where an elastic behaviour is expected), soil-structure interaction effects and non-structural elements may further affect the elastic natural period of buildings, changing the spectral accelerations compared with those provided by seismic codes in case of static analyses. In this work, a parametric study has been performed in order to evaluate the elastic fundamental period of vibration of buildings as a function of structural morphology (height, plan area, ratio between plan dimensions), infills presence and distribution and soil characteristics. Acknowledgements This study was partially funded by the Italian Department of Civil Protection within the project DPC-RELUIS 2016 - RS4 ''Seismic observatory of structures and health monitoring'' and by the "Centre of Integrated Geomorphology for the Mediterranean Area - CGIAM" within the Framework Agreement with the University of Basilicata "Study, Research and Experimentation in the Field of Analysis and Monitoring of Seismic Vulnerability of Strategic and Relevant Buildings for the purposes of Civil Protection and Development of Innovative Strategies of Seismic Reinforcement".

  13. Media arrangement impacts cell growth in anaerobic fixed-bed reactors treating sugarcane vinasse: Structured vs. randomic biomass immobilization.

    PubMed

    de Aquino, Samuel; Fuess, Lucas Tadeu; Pires, Eduardo Cleto

    2017-03-23

    This study reports on the application of an innovative structured-bed reactor (FVR) as an alternative to conventional packed-bed reactors (PBRs) to treat high-strength solid-rich wastewaters. Using the FVR prevents solids from accumulating within the fixed-bed, while maintaining the advantages of the biomass immobilization. The long-term operation (330days) of a FVR and a PBR applied to sugarcane vinasse under increasing organic loads (2.4-18.0kgCODm(-3)day(-1)) was assessed, focusing on the impacts of the different media arrangements over the production and retention of biomass. Much higher organic matter degradation rates, as well as long-term operational stability and high conversion efficiencies (>80%) confirmed that the FVR performed better than the PBR. Despite the equivalent operating conditions, the biomass growth yield was different in both reactors, i.e., 0.095gVSSg(-1)COD (FVR) and 0.066gVSSg(-1)COD (PBR), indicating a clear control of the media arrangement over the biomass production in fixed-bed reactors.

  14. Structural evolution of a two-component organogel.

    PubMed

    Singh, Mohit; Tan, Grace; Agarwal, Vivek; Fritz, Gerhard; Maskos, Karol; Bose, Arijit; John, Vijay; McPherson, Gary

    2004-08-31

    Dry reverse micelles of AOT in isooctane spontaneously undergo a microstructural transition to an organogel upon the addition of a phenolic dopant, p-chlorophenol. This microstructural evolution has been studied through a combination of light scattering, small-angle neutron scattering (SANS), NMR, and rheology. Several equilibrium stages between the system of dry reverse micelles of AOT and a 1:1 AOT/p-chlorophenol (molar ratio) gel in isooctane have been examined. To achieve this, p-chlorophenol is added progressively to the dilute solutions of AOT in isooctane, and this concentration series is then analyzed. The dry micelles of AOT in isooctane do not undergo any detectable structural change up to a certain p-chlorophenol concentration. Upon a very small increment in the concentration of p-chlorophenol beyond this "threshold" concentration, large strandlike aggregates are observed which then evolve to the three-dimensional gel network.

  15. Silicon carbide tritium permeation barrier for steel structural components.

    SciTech Connect

    Causey, Rion A.; Garde, Joseph Maurico; Buchenauer, Dean A.; Calderoni, Pattrick; Holschuh, Thomas, Jr.; Youchison, Dennis Lee; Wright, Matt; Kolasinski, Robert D.

    2010-09-01

    Chemical vapor deposited (CVD) silicon carbide (SiC) has superior resistance to tritium permeation even after irradiation. Prior work has shown Ultrametfoam to be forgiving when bonded to substrates with large CTE differences. The technical objectives are: (1) Evaluate foams of vanadium, niobium and molybdenum metals and SiC for CTE mitigation between a dense SiC barrier and steel structure; (2) Thermostructural modeling of SiC TPB/Ultramet foam/ferritic steel architecture; (3) Evaluate deuterium permeation of chemical vapor deposited (CVD) SiC; (4) D testing involved construction of a new higher temperature (> 1000 C) permeation testing system and development of improved sealing techniques; (5) Fabricate prototype tube similar to that shown with dimensions of 7cm {theta} and 35cm long; and (6) Tritium and hermeticity testing of prototype tube.

  16. Fabrication and evaluation of SiC/Cu functionally graded material used for plasma facing components in a fusion reactor

    NASA Astrophysics Data System (ADS)

    Ling, Yun-Han; Li, Jiang-Tao; Ge, Chang-Chun; Bai, Xin-De

    2002-06-01

    A new SiC/Cu functionally graded material that contains a spectrum of 0-100% compositional distributions of SiC used for plasma facing component was proposed and fabricated by a novel process termed graded sintering under ultra-high pressure, by which a near dense graded composite has been successfully obtained. Tests on plasma relevant performances showed that in SiC/Cu graded composite the CD 4 production due to chemical sputtering is 85% lower than that of SMF800 nuclear graphite, while its thermal desorption is about 10% of that graphite; fatigue cracks and chemical decomposition were found on the surface of SiC/Cu FGM after 300 cyclic impacts of laser pulse with power density of 398 MW/m 2; slight damage was also observed on the material surface after in situ plasma irradiation in a Tokamak facility.

  17. Vertical distribution of structural components in corn stover

    SciTech Connect

    Johnson, Jane M. F.; Karlen, Douglas L.; Gresham, Garold L.; Cantrell, Keri B.; Archer, David W.; Wienhold, Brian J.; Varvel, Gary E.; Laird, David A.; Baker, John; Ochsner, Tyson E.; Novak, Jeff M.; Halvorson, Ardell D.; Arriaga, Francisco; Lightle, David T.; Hoover, Amber; Emerson, Rachel; Barbour, Nancy W.

    2014-11-17

    In the United States, corn (Zea mays L.) stover has been targeted for second generation fuel production and other bio-products. Our objective was to characterize sugar and structural composition as a function of vertical distribution of corn stover (leaves and stalk) that was sampled at physiological maturity and about three weeks later from multiple USA locations. A small subset of samples was assessed for thermochemical composition. Concentrations of lignin, glucan, and xylan were about 10% greater at grain harvest than at physiological maturity, but harvestable biomass was about 25% less due to stalk breakage. Gross heating density above the ear averaged 16.3 ± 0.40 MJ kg⁻¹, but with an alkalinity measure of 0.83 g MJ⁻¹, slagging is likely to occur during gasification. Assuming a stover harvest height of 10 cm, the estimated ethanol yield would be >2500 L ha⁻¹, but it would be only 1000 L ha⁻¹ if stover harvest was restricted to the material from above the primary ear. Vertical composition of corn stover is relatively uniform; thus, decision on cutting height may be driven by agronomic, economic and environmental considerations.

  18. Vertical distribution of structural components in corn stover

    DOE PAGES

    Johnson, Jane M. F.; Karlen, Douglas L.; Gresham, Garold L.; ...

    2014-11-17

    In the United States, corn (Zea mays L.) stover has been targeted for second generation fuel production and other bio-products. Our objective was to characterize sugar and structural composition as a function of vertical distribution of corn stover (leaves and stalk) that was sampled at physiological maturity and about three weeks later from multiple USA locations. A small subset of samples was assessed for thermochemical composition. Concentrations of lignin, glucan, and xylan were about 10% greater at grain harvest than at physiological maturity, but harvestable biomass was about 25% less due to stalk breakage. Gross heating density above the earmore » averaged 16.3 ± 0.40 MJ kg⁻¹, but with an alkalinity measure of 0.83 g MJ⁻¹, slagging is likely to occur during gasification. Assuming a stover harvest height of 10 cm, the estimated ethanol yield would be >2500 L ha⁻¹, but it would be only 1000 L ha⁻¹ if stover harvest was restricted to the material from above the primary ear. Vertical composition of corn stover is relatively uniform; thus, decision on cutting height may be driven by agronomic, economic and environmental considerations.« less

  19. Coherent states, vacuum structure and infinite component relativistic wave equations

    NASA Astrophysics Data System (ADS)

    Cirilo-Lombardo, Diego Julio

    2016-11-01

    It is commonly claimed in the recent literature that certain solutions to wave equations of positive energy of Dirac-type with internal variables are characterized by a non-thermal spectrum. As part of that statement, it was said that the transformations and symmetries involved in equations of such type corresponded to a particular representation of the Lorentz group. In this paper, we give the general solution to this problem emphasizing the interplay between the group structure, the corresponding algebra and the physical spectrum. This analysis is completed with a strong discussion and proving that: (i) the physical states are represented by coherent states; (ii) the solutions in [Yu. P. Stepanovsky, Nucl. Phys. B (Proc. Suppl.) 102 (2001) 407-411; 103 (2001) 407-411] are not general, (iii) the symmetries of the considered physical system in [Yu. P. Stepanovsky, Nucl. Phys. B (Proc. Suppl.) 102 (2001) 407-411; 103 (2001) 407-411] (equations and geometry) do not correspond to the Lorentz group but to the fourth covering: the Metaplectic group Mp(n).

  20. Structural design criteria for high heat flux components.

    SciTech Connect

    Majumdar, S.

    1999-07-14

    The high temperature design rules of the ITER Structural Design Criteria (ISDC), are applied to first wall designs with high heat flux. The maximum coolant pressure and surface heat flux capabilities are shown to be determined not only by the mechanical properties of the first wall material but also by the details of the blanket design. In a high power density self-cooled lithium blanket, the maximum primary stress in the first wall is controlled by many of the geometrical parameters of the blanket, such as, first wall span, first wall curvature, first wall thickness, side wall thickness, and second wall thickness. The creep ratcheting lifetime of the first wall is also shown to be controlled by many of the same geometrical parameters as well as the coolant temperature. According to most high temperature design codes, the time-dependent primary membrane stress allowable are based on the average temperature (ignoring thermal stress). Such a procedure may sometimes be unconservative, particularly for embrittled first walls with large temperature gradients. The effect of secondary (thermal) stresses on the accumulation of creep deformation is illustrated with a vanadium alloy flat plate first wall design.

  1. Vertical distribution of structural components in corn stover

    SciTech Connect

    Jane M. F. Johnson; Douglas L. Karlen; Garold L. Gresham; Keri B. Cantrell; David W. Archer; Brian J. Wienhold; Gary E. Varvel; David A. Laird; John Baker; Tyson E. Ochsner; Jeff M. Novak; Ardell D. Halvorson; Francisco Arriaga; David T. Lightle; Amber Hoover; Rachel Emerson; Nancy W. Barbour

    2014-11-01

    In the United States, corn (Zea mays L.) stover has been targeted for second generation fuel production and other bio-products. Our objective was to characterize sugar and structural composition as a function of vertical distribution of corn stover (leaves and stalk) that was sampled at physiological maturity and about three weeks later from multiple USA locations. A small subset of samples was assessed for thermochemical composition. Concentrations of lignin, glucan, and xylan were about 10% greater at grain harvest than at physiological maturity, but harvestable biomass was about 25% less due to stalk breakage. Gross heating density above the ear averaged 16.3 ± 0.40 MJ kg?¹, but with an alkalinity measure of 0.83 g MJ?¹, slagging is likely to occur during gasification. Assuming a stover harvest height of 10 cm, the estimated ethanol yield would be >2500 L ha?¹, but it would be only 1000 L ha?¹ if stover harvest was restricted to the material from above the primary ear. Vertical composition of corn stover is relatively uniform; thus, decision on cutting height may be driven by agronomic, economic and environmental considerations.

  2. A study on nuclear properties of Zr, Nb, and Ta nuclei used as structural material in fusion reactor

    NASA Astrophysics Data System (ADS)

    Sahan, Halide; Tel, Eyyup; Sahan, Muhittin; Aydin, Abdullah; Hakki Sarpun, Ismail; Kara, Ayhan; Doner, Mesut

    2015-07-01

    Fusion has a practically limitless fuel supply and is attractive as an energy source. The main goal of fusion research is to construct and operate an energy generating system. Fusion researches also contains fusion structural materials used fusion reactors. Material issues are very important for development of fusion reactors. Therefore, a wide range of fusion structural materials have been considered for fusion energy applications. Zirconium (Zr), Niobium (Nb) and Tantalum (Ta) containing alloys are important structural materials for fusion reactors and many other fields. Naturally Zr includes the 90Zr (%51.5), 91Zr (%11.2), 92Zr (%17.1), 94Zr (%17.4), 96Zr (%2.80) isotopes and 93Nb and 181Ta include the 93Nb (%100) and 181Ta (%99.98), respectively. In this study, the charge, mass, proton and neutron densities and the root-mean-square (rms) charge radii, rms nuclear mass radii, rms nuclear proton, and neutron radii have been calculated for 87-102Zr, 93Nb, 181Ta target nuclei isotopes by using the Hartree-Fock method with an effective Skyrme force with SKM*. The calculated results have been compared with those of the compiled experimental taken from Atomic Data and Nuclear Data Tables and theoretical values of other studies.

  3. Emergency heat removal system for a nuclear reactor

    DOEpatents

    Dunckel, Thomas L.

    1976-01-01

    A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

  4. 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.

  5. RSMASS: A preliminary reactor/shield mass model for SDI applications

    SciTech Connect

    Marshall, A.C.

    1986-08-01

    A simple mathematical model (RSMASS) has been developed to provide rapid estimates of reactor and shield masses for space-based reactor power systems. Approximations are used rather than correlations or detailed calculations to estimate the reactor fuel mass and the masses of the moderator, structure, reflector, pressure vessel, miscellaneous components, and the reactor shield. The fuel mass is determined either by neutronics limits, specific power limits, or fuel burnup limits - whichever yields the largest mass. RSMASS requires the reactor power and energy, 24 reactor parameters, and 20 shield parameters to be specified. This parametric approach should provide good mass estimates for a very broad range of reactor types. Reactor and shield masses calculated by RSMASS were found to be in good agreement with the masses obtained from detailed calculations.

  6. Savannah River reactor process water heat exchanger tube structural integrity margin Task Number 92-005-1

    SciTech Connect

    Mertz, G.E.; Barnes, D.M.; Sindelar, R.L.

    1992-02-01

    Twelve process water heat exchangers are designed to remove heat generated in the reactor tank. Each heat exchanger has approximately 9000, 1/2 inch diameter {times} 0.049 inches thick tubes. Minimum structural tubing requirements and the leak rate through postulated tubing defects are developed in this report A comparison of the structural requirements and the defect size calculated to produce leak rates of 0.5 lbs./day demonstrate adequate structural margins against gross tube rupture. Commercial nuclear experience with pressurized water reactor (PWR) steam generator plugging criteria are used for guidance in performing this analysis. It is important to note that the SRS reactors are low energy systems with normal operating pressures of 203 psig at 130{degree}F while the PWR is a high energy system with operating pressures near 2200 psig at 600{degree}F. Clearly the PVM steam generator has loadings which are more severe than the SRS heat exchangers. Consistent with the Regulatory Guide 1.121 criteria both wastage (wall thinning) and cracking are addressed. Structural limits on wall thinning and crack size are developed to preclude gross rupture. ASME Section XI criteria, with the factors of safety recommended by Regulatory Guide 1.121 are used to develop the allowable crack size criteria. Normal operating conditions (pressure, dead weight, and hydraulic drag) are considered with seismic and water hammer accident conditions. Both the wall thinning and crack size criteria are developed for the end-of-evaluation period. Allowances for corrosion, wear, or crack growth have not been included in this analysis Structurally, the tubing is over designed and can tolerate large defects with adequate margins against gross rupture. The structural margins of heat exchanger tubing are evident by contrasting the tubing's structural capacity, per the ASME Code, with its operating conditions/configuration.

  7. Savannah River reactor process water heat exchanger tube structural integrity margin Task Number 92-005-1

    SciTech Connect

    Mertz, G.E.; Barnes, D.M.; Sindelar, R.L.

    1992-02-01

    Twelve process water heat exchangers are designed to remove heat generated in the reactor tank. Each heat exchanger has approximately 9000, 1/2 inch diameter {times} 0.049 inches thick tubes. Minimum structural tubing requirements and the leak rate through postulated tubing defects are developed in this report A comparison of the structural requirements and the defect size calculated to produce leak rates of 0.5 lbs./day demonstrate adequate structural margins against gross tube rupture. Commercial nuclear experience with pressurized water reactor (PWR) steam generator plugging criteria are used for guidance in performing this analysis. It is important to note that the SRS reactors are low energy systems with normal operating pressures of 203 psig at 130{degree}F while the PWR is a high energy system with operating pressures near 2200 psig at 600{degree}F. Clearly the PVM steam generator has loadings which are more severe than the SRS heat exchangers. Consistent with the Regulatory Guide 1.121 criteria both wastage (wall thinning) and cracking are addressed. Structural limits on wall thinning and crack size are developed to preclude gross rupture. ASME Section XI criteria, with the factors of safety recommended by Regulatory Guide 1.121 are used to develop the allowable crack size criteria. Normal operating conditions (pressure, dead weight, and hydraulic drag) are considered with seismic and water hammer accident conditions. Both the wall thinning and crack size criteria are developed for the end-of-evaluation period. Allowances for corrosion, wear, or crack growth have not been included in this analysis Structurally, the tubing is over designed and can tolerate large defects with adequate margins against gross rupture. The structural margins of heat exchanger tubing are evident by contrasting the tubing`s structural capacity, per the ASME Code, with its operating conditions/configuration.

  8. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

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

  9. The structure of decommissioning plan for VVR-S research reactor

    SciTech Connect

    Popescu, C.; Paunescu, A.; Garlea, I.; Garlea, C.

    1996-12-31

    This paper presents the activity in preparing the decommissioning plan for the VVR-S research reactor taking into account that there is no experience in this field in Hungary. VVR-S IPNE reactor, situated in Magurele Village, near Bucharest, was put into operation on the 27th of July 1957 and it has continuously been operating without major events, up to now. During this period, no modifications concerning the core and main circuits were made. The reactor is still operating with the original equipment and control instrumentation supplied by Ex-Soviets Union. The reactor was designed and built at the scientific and technological level of the 50`s. VVR-S IPNE reactor is a thermal one, of 2 MW nominal power and tank-type. It is light water cooled and moderated. The core operates with EK-10 type fuel assemblies (made of UO{sub 2} - MgO, with 10% U{sup 235} enrichment); after 1984 a mixed core with EK-10 and S-36 type assemblies was employed. The present fuel stock is S-36 type (36.6% U{sup 235} enrichment).

  10. Principal component analysis for surface reflection components and structure in facial images and synthesis of facial images for various ages

    NASA Astrophysics Data System (ADS)

    Hirose, Misa; Toyota, Saori; Ojima, Nobutoshi; Ogawa-Ochiai, Keiko; Tsumura, Norimichi

    2017-08-01

    In this paper, principal component analysis is applied to the distribution of pigmentation, surface reflectance, and landmarks in whole facial images to obtain feature values. The relationship between the obtained feature vectors and the age of the face is then estimated by multiple regression analysis so that facial images can be modulated for woman aged 10-70. In a previous study, we analyzed only the distribution of pigmentation, and the reproduced images appeared to be younger than the apparent age of the initial images. We believe that this happened because we did not modulate the facial structures and detailed surfaces, such as wrinkles. By considering landmarks and surface reflectance over the entire face, we were able to analyze the variation in the distributions of facial structures and fine asperity, and pigmentation. As a result, our method is able to appropriately modulate the appearance of a face so that it appears to be the correct age.

  11. Community structure and function in a H(2)-based membrane biofilm reactor capable of bioreduction of selenate and chromate.

    PubMed

    Chung, Jinwook; Ryu, Hodon; Abbaszadegan, Morteza; Rittmann, Bruce E

    2006-10-01

    Two different H(2)-based, denitrifying membrane-biofilm reactors (MBfRs) initially reduced Se(VI) or Cr(VI) stably to Se(0) or Cr(III). When the oxidized contaminants in the influent were switched, each new oxidized contaminant was reduced immediately, and its reduction soon was approximately the same or greater than it had been in its original MBfR. The precipitation of reduced selenium and chromium in the biofilm was verified by scanning electron microscopy and energy dispersive X-ray analysis. These results on selenate and chromate reduction are consistent with the interpretation that the H(2)-based biofilm community had a high level of functional diversity. The communities' structures were assessed by cloning analysis. Dechloromonas spp., a known perchlorate-reducing bacteria, dominated the clones from both reactors during selenate and chromate reductions, which suggests that it may have functional diversity capable of reducing selenate and chromate as secondary and dissimilatory acceptors.

  12. Measurement of wavefront structure from large aperture optical components by phase shifting interferometry

    SciTech Connect

    Wolfe, C.R.; Lawson, J.K.; Kellam, M.; Maney, R.T.; Demiris, A.

    1995-05-12

    This paper discusses the results of high spatial resolution measurement of the transmitted or reflected wavefront of optical components using phase shifting interferometry with a wavelength of 6328 {angstrom}. The optical components studied range in size from approximately 50 mm {times} 100 mm to 400 mm {times} 750 mm. Wavefront data, in the form of 3-D phase maps, have been obtained for three regimes of scale length: ``micro roughness``, ``mid-spatial scale``, and ``optical figure/curvature.`` Repetitive wavefront structure has been observed with scale lengths from 10 mm to 100 mm. The amplitude of this structure is typically {lambda}/100 to {lambda}/20. Previously unobserved structure has been detected in optical materials and on the surfaces of components. We are using this data to assist in optimizing laser system design, to qualify optical components and fabrication processes under study in our component development program.

  13. REACTOR UNLOADING

    DOEpatents

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  14. Analysis of the low temperature ceramics structure with consideration for polydispersity of initial refractory components

    NASA Astrophysics Data System (ADS)

    Leytsin, Vladimir N.; Dmitrieva, Mariya A.; Tovpinets, Alexandr O.; Ivonin, Ivan V.; Ponomarev, Sergey V.

    2016-11-01

    The results of computer simulation of the structure and physical properties of sintered low-temperature ceramics specimens with different volume fractions of different components of refractory components are presented. Properties of sintered ceramics, residual porosity, and shrinkage anisotropy are determined by features of packing of various fractions of refractory particles. The results indicate the determining factor of the presence of particles of the coarse fraction of refractory components capable of forming a internal skeleton of interacting particles.

  15. Proceedings of the Office of Fusion Energy/DOE workshop on ceramic matrix composites for structural applications in fusion reactors

    SciTech Connect

    Jones, R.H. ); Lucas, G.E. )

    1990-11-01

    A workshop to assess the potential application of ceramic matrix composites (CMCs) for structural applications in fusion reactors was held on May 21--22, 1990, at University of California, Santa Barbara. Participants included individuals familiar with materials and design requirements in fusion reactors, ceramic composite processing and properties and radiation effects. The primary focus was to list the feasibility issues that might limit the application of these materials in fusion reactors. Clear advantages for the use of CMCs are high-temperature operation, which would allow a high-efficiency Rankine cycle, and low activation. Limitations to their use are material costs, fabrication complexity and costs, lack of familiarity with these materials in design, and the lack of data on radiation stability at relevant temperatures and fluences. Fusion-relevant feasibility issues identified at this workshop include: hermetic and vacuum properties related to effects of matrix porosity and matrix microcracking; chemical compatibility with coolant, tritium, and breeder and multiplier materials, radiation effects on compatibility; radiation stability and integrity; and ability to join CMCs in the shop and at the reactor site, radiation stability and integrity of joints. A summary of ongoing CMC radiation programs is also given. It was suggested that a true feasibility assessment of CMCs for fusion structural applications could not be completed without evaluation of a material tailored'' to fusion conditions or at least to radiation stability. It was suggested that a follow-up workshop be held to design a tailored composite after the results of CMC radiation studies are available and the critical feasibility issues are addressed.

  16. Visible Light Driven Photocatalytic Reactor Based on Micro-structured Polymer Optical Fiber Preform

    NASA Astrophysics Data System (ADS)

    Li, Dong-Dong; She, Jiang-Bo; Wang, Chang-Shun; Peng, Bo

    2014-05-01

    A novel visible light driven photocatalytic reactor with 547 pieces of Ag/AgBr-film-modified capillaries is reported and it is derived from a microstructured polymer optical fiber (MPOF) preform. The MPOF preform not only plays the role of a light-transmitting media, but it is also a Ag/AgBr supporting and waste-water pipe to supply the photocatalytic degradation of dyes solute. The photocatalytic reactor has such a large surface area for Ag/AgBr loading, which is a visible light driven photocatalyst that photodegradation efficiency is enhanced.

  17. Probabilistic Structural Analysis Methods (PSAM) for Select Space Propulsion System Components

    NASA Technical Reports Server (NTRS)

    1999-01-01

    Probabilistic Structural Analysis Methods (PSAM) are described for the probabilistic structural analysis of engine components for current and future space propulsion systems. Components for these systems are subjected to stochastic thermomechanical launch loads. Uncertainties or randomness also occurs in material properties, structural geometry, and boundary conditions. Material property stochasticity, such as in modulus of elasticity or yield strength, exists in every structure and is a consequence of variations in material composition and manufacturing processes. Procedures are outlined for computing the probabilistic structural response or reliability of the structural components. The response variables include static or dynamic deflections, strains, and stresses at one or several locations, natural frequencies, fatigue or creep life, etc. Sample cases illustrates how the PSAM methods and codes simulate input uncertainties and compute probabilistic response or reliability using a finite element model with probabilistic methods.

  18. 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.

  19. Extended Aging Theories for Predictions of Safe Operational Life of Critical Airborne Structural Components

    NASA Technical Reports Server (NTRS)

    Ko, William L.; Chen, Tony

    2006-01-01

    The previously developed Ko closed-form aging theory has been reformulated into a more compact mathematical form for easier application. A new equivalent loading theory and empirical loading theories have also been developed and incorporated into the revised Ko aging theory for the prediction of a safe operational life of airborne failure-critical structural components. The new set of aging and loading theories were applied to predict the safe number of flights for the B-52B aircraft to carry a launch vehicle, the structural life of critical components consumed by load excursion to proof load value, and the ground-sitting life of B-52B pylon failure-critical structural components. A special life prediction method was developed for the preflight predictions of operational life of failure-critical structural components of the B-52H pylon system, for which no flight data are available.

  20. BDDR, a new CEA technological and operating reactor database

    SciTech Connect

    Soldevilla, M.; Salmons, S.; Espinosa, B.

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  1. Measurement of neutron spectra in the experimental reactor LR-0

    SciTech Connect

    Prenosil, Vaclav; Mravec, Filip; Veskrna, Martin; Kostal, Michal; Matej, Zdenek; Cvachovec, Frantisek

    2015-07-01

    The measurement of fast neutron fluxes is important in many areas of nuclear technology. It affects the stability of the reactor structural components, performance of fuel, and also the fuel manner. The experiments performed at the LR-0 reactor were in the past focused on the measurement of neutron field far from the core, in reactor pressure vessel simulator or in biological shielding simulator. In the present the measurement in closer regions to core became more important, especially measurements in structural components like reactor baffle. This importance increases with both reactor power increase and also long term operation. Other important task is an increasing need for the measurement close to the fuel. The spectra near the fuel are aimed due to the planned measurements with the FLIBE salt, in FHR / MSR research, where one of the task is the measurement of the neutron spectra in it. In both types of experiments there is strong demand for high working count rate. The high count rate is caused mainly by high gamma background and by high fluxes. The fluxes in core or in its vicinity are relatively high to ensure safe reactor operation. This request is met in the digital spectroscopic apparatus. All experiments were realized in the LR-0 reactor. It is an extremely flexible light water zero-power research reactor, operated by the Research Center Rez (Czech Republic). (authors)

  2. Experimental Study on Flow Optimization in Upper Plenum of Reactor Vessel for a Compact Sodium-Cooled Fast Reactor

    SciTech Connect

    Kimura, Nobuyuki; Hayashi, Kenji; Kamide, Hideki; Itoh, Masami; Sekine, Tadashi

    2005-11-15

    An innovative sodium-cooled fast reactor has been investigated in a feasibility study of fast breeder reactor cycle systems in Japan. A compact reactor vessel and a column-type upper inner structure with a radial slit for an arm of a fuel-handling machine (FHM) are adopted. Dipped plates are set in the reactor vessel below the free surface to prevent gas entrainment. We performed a one-tenth-scaled model water experiment for the upper plenum of the reactor vessel. Gas entrainment was not observed in the experiment under the same velocity condition as the reactor. Three vortex cavitations were observed near the hot-leg inlet. A vertical rib on the reactor vessel wall was set to restrict the rotating flow near the hot leg. The vortex cavitation between the reactor vessel wall and the hot leg was suppressed by the rib under the same cavitation factor condition as in the reactor. The cylindrical plug was installed through the hole in the dipped plates for the FHM to reduce the flow toward the free surface. It was effective when the plug was submerged into the middle height in the upper plenum. This combination of two components had a possibility to optimize the flow in the compact reactor vessel.

  3. Transient heat and mass transfer analysis in a porous ceria structure of a novel solar redox reactor

    SciTech Connect

    Chandran, RB; Bader, R; Lipinski, W

    2015-06-01

    Thermal transport processes are numerically analyzed for a porous ceria structure undergoing reduction in a novel redox reactor for solar thermochemical fuel production. The cylindrical reactor cavity is formed by an array of annular reactive elements comprising the porous ceria monolith integrated with gas inlet and outlet channels. Two configurations are considered, with the reactor cavity consisting of 10 and 20 reactive elements, respectively. Temperature dependent boundary heat fluxes are obtained on the irradiated cavity wall by solving for the surface radiative exchange using the net radiation method coupled to the heat and mass transfer model of the reactive element. Predicted oxygen production rates are in the range 40-60 mu mol s(-1) for the geometries considered. After an initial rise, the average temperature of the reactive element levels off at 1660 and 1680 K for the two geometries, respectively. For the chosen reduction reaction rate model, oxygen release continues after the temperature has leveled off which indicates that the oxygen release reaction is limited by chemical kinetics and/or mass transfer rather than by the heating rate. For a fixed total mass of ceria, the peak oxygen release rate is doubled for the cavity with 20 reactive elements due to lower local oxygen partial pressure. (C) 2015 Elsevier Masson SAS. All rights reserved.

  4. Use of molybdenum as a structural material of fuel elements for improving the safety of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Shmelev, A. N.; Kozhahmet, B. K.

    2017-01-01

    Main purpose of the study is justifying the use of molybdenum as a structural material of fuel elements for improving the safety of nuclear reactors. Particularity of used molybdenum is that its isotopic composition corresponds to molybdenum, which is obtained as the tailing during operation of the separation cascade for producing a material for medical diagnostics of cancer. When performing the study the neutron-physical properties of isotopes of natural molybdenum (nuclear data library JENDL-4.0) and thermal properties of metallic molybdenum were used. The following results were obtained: 1. A method for reducing the thermal constant of fuel elements for light water and fast reactors by using dispersion fuel in cylindrical fuel rods containing, for example, granules of metallic U-Mo-alloy into Mo-matrix was proposed. 2. The necessity of molybdenum enrichment by weakly absorbing isotopes was shown. 3. Total use of isotopic molybdenum will be more than 50%. A method for reducing the thermal constant of the fuel elements, allowing us to increase the safety of light water and fast nuclear reactors by using dispersion fuel in cylindrical fuel rods containing, for example, granules of metallic U-Mo-alloy into Mo-matrix with enrichment by weakly absorbing isotopes of molybdenum is proposed.

  5. Prediction of stainless steel activation in experimental breeder reactor 2 (EBR-II) reflector and blanket subassemblies

    SciTech Connect

    Bunde, K.A.

    1996-12-31

    Stainless steel structural components in nuclear reactors become radioactive wastes when no longer useful. Prior to disposal, certain physical attributes must be analyzed. These attributes include structural integrity, chemical stability, and the radioactive material content among others. The focus of this work is the estimation of the radioactive material content of stainless steel wastes from a research reactor operated by Argonne National Laboratory.

  6. [Research thoughts on structural components of Chinese medicine combined with bioinformatics].

    PubMed

    Wang, Cheng-cheng; Feng, Liang; Liu, Dan; Cui, Li; Tan, Xiao-bin; Jia, Xiao-bin

    2015-11-01

    Traditional Chinese medicine(TCM) is a complex system, featured with integrity and characteristics. Structural component TCM is a well-organized integrity of traditional Chinese medicine, reflecting multi-component integration effect of TCM. It gives us a new view on the material basis of TCM. Currently, conventional researching strategies are not enough to deal with the relationship between material basis and efficacy, multi-composition, multi-targets, and multi-section mechanism. Post-genome area gives a birth to bioinformatics, which involves systematic biology, different levels of omics, corresponding mathematics and computer techniques. It increasingly becomes a powerful tool to understand complicated system and life essential laws. Research ideas, methods. and knowledge of data mining technology of bioinformatics combined with the theory of structural components of Chinese medicine bring a new opportunity for developing structural components of Chinese medicine, systematically exploring the essence of TCM and promoting the modernization of TCM.

  7. 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.

  8. Virtual ultrasound sources for inspecting nuclear components of coarse-grained structure

    SciTech Connect

    Brizuela, J.; Katchadjian, P.; Desimone, C.; Garcia, A.

    2014-02-18

    This work describes an ultrasonic inspection procedure designed for verifying coarse-grained structure materials, which are commonly used on nuclear reactors. In this case, conventional phased array techniques cannot be used due to attenuating characteristics and backscattered noise from microstructures inside the material. Thus, synthetic aperture ultrasonic imaging (SAFT) is used for this approach in contact conditions. In order to increase energy transferred to the medium, synthetic transmit aperture is formed by several elements which generate a diverging wavefront equivalent to a virtual ultrasound source behind the transducer. On the other hand, the phase coherence technique has been applied to reduce more structural noise and improve the image quality. The beamforming process has been implemented over a GPU platform to reduce computing time.

  9. Measuring School Improvement Implementation: Validation of the School Effectiveness Structural Components Inventory.

    ERIC Educational Resources Information Center

    Koppel, Sheree P.; And Others

    The School Effectiveness Structural Components Inventory (SESCI) was designed to assess structural elements in a school environment that contribute to effectiveness. The preliminary validation took place in 12 elementary schools in or near a large metropolitan school district. The schools were classified by overall school socioeconomic status, as…

  10. On bi-Hamiltonian structure of two-component Novikov equation

    NASA Astrophysics Data System (ADS)

    Li, Nianhua; Liu, Q. P.

    2013-01-01

    In this Letter, we present a bi-Hamiltonian structure for the two-component Novikov equation. We also show that proper reduction of this bi-Hamiltonian structure leads to the Hamiltonian operators found by Hone and Wang for the Novikov equation.

  11. 78 FR 41434 - Proposed Revisions to Design of Structures, Components, Equipment and Systems

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-10

    ...) reflect the current staff's review methods and practices based on lessons learned from NRC reviews of... Effects of the Light-water Reactor Environment for New Reactors,'' (ADAMS Accession No. ML070380586). The...

  12. Vlasov Simulation of Electrostatic Solitary Structures in Multi-Component Plasmas

    NASA Technical Reports Server (NTRS)

    Umeda, Takayuki; Ashour-Abdalla, Maha; Pickett, Jolene S.; Goldstein, Melvyn L.

    2012-01-01

    Electrostatic solitary structures have been observed in the Earth's magnetosheath by the Cluster spacecraft. Recent theoretical work has suggested that these solitary structures are modeled by electron acoustic solitary waves existing in a four-component plasma system consisting of core electrons, two counter-streaming electron beams, and one species of background ions. In this paper, the excitation of electron acoustic waves and the formation of solitary structures are studied by means of a one-dimensional electrostatic Vlasov simulation. The present result first shows that either electron acoustic solitary waves with negative potential or electron phase-space holes with positive potential are excited in four-component plasma systems. However, these electrostatic solitary structures have longer duration times and higher wave amplitudes than the solitary structures observed in the magnetosheath. The result indicates that a high-speed and small free energy source may be needed as a fifth component. An additional simulation of a five-component plasma consisting of a stable four-component plasma and a weak electron beam shows the generation of small and fast electron phase-space holes by the bump-on-tail instability. The physical properties of the small and fast electron phase-space holes are very similar to those obtained by the previous theoretical analysis. The amplitude and duration time of solitary structures in the simulation are also in agreement with the Cluster observation.

  13. A robust nonlinear attitude control law for space stations with flexible structural components

    NASA Technical Reports Server (NTRS)

    Wang, P. K. C.

    1985-01-01

    In this paper, a nonlinear attitude control law for space stations with flexible structural components is derived using a rigid-body model. This control law, depending on the Cayley-Rodriguez parameters, globally stabilizes the equilibrium of the rigid-body model. The effect of elastic deformations of the flexible structural components on the resulting feedback system dynamics is analyzed. It is found that the system's stability property is highly robust with respect to structural vibrations and inertial variations. The time-domain behavior of the feedback system is studied numerically using a model of a typical space station with flexible solar panels.

  14. Boron carbide: Consistency of components, lattice parameters, fine structure and chemical composition makes the complex structure reasonable

    NASA Astrophysics Data System (ADS)

    Werheit, Helmut

    2016-10-01

    The complex, highly distorted structure of boron carbide is composed of B12 and B11C icosahedra and CBC, CBB and B□B linear elements, whose concentration depends on the chemical composition each. These concentrations are shown to be consistent with lattice parameters, fine structure data and chemical composition. The respective impacts on lattice parameters are estimated and discussed. Considering the contributions of the different structural components to the energy of the overall structure makes the structure and its variation within the homogeneity range reasonable; in particular that of B4.3C representing the carbon-rich limit of the homogeneity range. Replacing in B4.3C virtually the B□B components by CBC yields the hypothetical moderately distorted B4.0C (structure formula (B11C)CBC). The reduction of lattice parameters related is compatible with recently reported uncommonly prepared single crystals, whose compositions deviate from B4.3C.

  15. Fast Reactors

    NASA Astrophysics Data System (ADS)

    Esposito, S.; Pisanti, O.

    The following sections are included: * Elementary Considerations * The Integral Equation to the Neutron Distribution * The Critical Size for a Fast Reactor * Supercritical Reactors * Problems and Exercises

  16. Structures for attaching or sealing a space between components having different coefficients or rates of thermal expansion

    DOEpatents

    Corman, Gregory Scot; Dean, Anthony John; Tognarelli, Leonardo; Pecchioli, Mario

    2005-06-28

    A structure for attaching together or sealing a space between a first component and a second component that have different rates or amounts of dimensional change upon being exposed to temperatures other than ambient temperature. The structure comprises a first attachment structure associated with the first component that slidably engages a second attachment structure associated with the second component, thereby allowing for an independent floating movement of the second component relative to the first component. The structure can comprise split rings, laminar rings, or multiple split rings.

  17. Preliminary evaluation of beta-spodumene as a fusion reactor structural material

    SciTech Connect

    Kelsey, P.V. Jr.; Schmunk, R.E.; Henslee, S.P.

    1981-01-01

    Beta-spodumene was investigated as a candidate material for use in fusion reactor environments. Properties which support the use of beta-spodumene include good thermal shock resistance, a very low coefficient of thermal expansion, a low-Z composition which would result in minimum impact on the plasma, and flexibility in fabrication processes. Specimens were irradiated in the Advanced Test Reactor (ATR) to a fluence of 5.3 x 10/sup 22/ n/m/sup 2/, E > 0.1 MeV, and 4.9 x 10/sup 23/ n/m/sup 2/ thermal fluence in order to obtain a preliminary evaluation of the impact of irradiation on the material. Preliminary data indicate that the mechanical properties of beta-spodumene are little affected by irradiation. Gas production and release have also been investigated.

  18. Reactor pressure vessel structural integrity research in the US Nuclear Regulatory Commission HSST and HSSI Programs

    SciTech Connect

    Pennell, W.E.; Corwin, W.R.

    1994-02-01

    This report discusses development on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels containing flaws. Fracture mechanics tests on reactor pressure vessel steel have shown that local brittle zones do not significantly degrade the material fracture toughness, constraint relaxation at the crack tip of shallow surface flaws results in increased fracture toughness, and biaxial loading reduces but does not eliminate the shallow-flaw fracture toughness elevation. Experimental irradiation investigations have shown that the irradiation-induced shift in Charpy V-notch versus temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement and the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.

  19. A Project Management and Systems Engineering Structure for a Generation IV Very High Temperature Reactor

    SciTech Connect

    Ed Gorski; Dennis Harrell; Finis Southworth

    2004-09-01

    The Very High Temperature Reactor (VHTR) will be an advanced, very high temperature (approximately 1000o C. coolant outlet temperature), gas cooled nuclear reactor and is the nearest term of six Generation IV reactor technologies for nuclear assisted hydrogen production. In 2001, the Generation IV International Forum (GIF), a ten nation international forum working together with the Department of Energy’s (DOE) Nuclear Energy Research Advisory Committee (NERAC), agreed to proceed with the development of a technology roadmap and identified the next generation of nuclear reactor systems for producing new sources of power. Since a new reactor has not been licensed in the United States since the 1970s, the risks are too large for a single utility to assume in the development of an unprecedented Generation IV reactor. The government must sponsor and invest in the research to resolve major first of a kind (FOAK) issues through a full-scale demonstration prior to industry implementation. DOE’s primary mission for the VHTR is to demonstrate nuclear reactor assisted cogeneration of electricity and hydrogen while meeting the Generation IV goals for safety, sustainability, proliferation resistance and physical security and economics. The successful deployment of the VHTR as a demonstration project will aid in restarting the now atrophied U.S. nuclear power industry infrastructure. It is envisioned that VHTR project participants will include DOE Laboratories, industry partners such as designers, constructors, manufacturers, utilities, and Generation IV international countries. To effectively mange R&D, engineering, procurement, construction, and operation for this multi-organizational and technologically complex project, systems engineering will be used extensively to ensure delivery of the final product. Although the VHTR is an unprecedented FOAK system, the R&D, when assessed using the Office of Science and Technology Gate Model, falls primarily in the 3rd - Exploratory

  20. Enabling Propulsion Materials (EPM) Structural Component Successfully Tested Under Pseudo-Operating Conditions

    NASA Technical Reports Server (NTRS)

    Bartolotta, Paul A.

    1997-01-01

    A fabrication feasibility demonstration component for the Enabling Propulsion Materials (EPM) program was evaluated under prototypical engine loading conditions at the Structural Benchmark Test Facility at the NASA Lewis Research Center. The purpose for this test was to verify EPM casting, joining, coating, and life-prediction methods. Electron beam welding techniques developed in the EPM program were used to join two large superalloy cast sections of an exhaust nozzle flap to fabricate the demonstration component. After the joints were inspected, the component was coated with an oxidation-resistant barrier coating and was sent to Lewis for testing. The special test fixture shown in the photo (the Structural Benchmark Test Facility) was designed and built at Lewis to produce a biaxial bending condition similar to the loading condition this part would encounter during engine operation. Several finite element analyses were conducted to validate the mechanical test method. A floating furnace was then designed to provide prototypical thermal profiles in the component. An isothermal low-cycle fatigue test was used to evaluate the component at a cyclic load of 13 kN (maximum) to 1 kN (minimum) at a frequency of 1 Hz. Component failure was defined as a 30-percent increase in the component's compliance. On the basis of this definition, the low-cycle fatigue life of this component would be 35,000 cycles.

  1. Flow visualization of bubble structure in bubble column reactor for fluid mixing

    NASA Astrophysics Data System (ADS)

    Ibrahim, Nur Afizah; Khalid, Amir; Zaman, Izzuddin; Sapit, Azwan; Manshoor, Bukhari

    2017-04-01

    Bubble columns reactor is widely used as gas-liquid mixing and as reactors in many industries especially in chemical, petrochemical and biochemical processing. High interfacial area between the gas and liquid phase will enhanced an effective mixing, leading to improved heat and mass transfer characteristics under bubble columns become an attractive choice as reactors for the described processes. In this research, experimental work by using cylindrical acrylic bubble column with internal diameter of 0.15 m and height of 1 m was done. The bubble column is equipped by four nozzles with orifice diameter of 5mm function as gas distributor attach at the bottom of the column. For this study, gas phase and liquid phase used are air and water respectively. The investigated parameter was mechanism of bubble formation, regime analysis and the relationship between superficial gas holdup and gas holdup. The techniques used in collecting data were visual observation, measurement technique and photographic method. The result showed that there were five stage of bubble formation based on experiment conducted. For gas holdup and superficial gas velocity relationship, it was discovered that the gas holdup increased with the increasing of superficial gas velocity.

  2. 105-H Reactor Interim Safe Storage Project Final Report

    SciTech Connect

    E.G. Ison

    2008-11-08

    The following information documents the decontamination and decommissioning of the 105-H Reactor facility, and placement of the reactor core into interim safe storage. The D&D of the facility included characterization, engineering, removal of hazardous and radiologically contaminated materials, equipment removal, decontamination, demolition of the structure, and restoration of the site. The ISS work also included construction of the safe storage enclosure, which required the installation of a new roofing system, power and lighting, a remote monitoring system, and ventilation components.

  3. Liquid Metal Cooled Reactor for Space Power

    NASA Astrophysics Data System (ADS)

    Weitzberg, Abraham

    2003-01-01

    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  4. Radiation Resistance of Structural Materials of Nuclear Reactors on Irradiation with High-Energy Hydrogen and Helium Ions

    NASA Astrophysics Data System (ADS)

    Komarov, F. F.; Komarov, A. F.; Pil‧ko, Vl. V.; Pil‧ko, V. V.

    2013-11-01

    Basic principles of determination of the radiation resistance of structural materials of nuclear reactors with implantation of high-energy hydrogen and helium atoms have been presented. The parameters of the process of implantation of light irons have been calculated. By scanning-electron-microscopy, optical-microscopy, and interference methods, the authors have studied the surface structure of samples of steel-3, stainless steel, and D16 alloy immediately after irradiating them with hydrogen and helium atoms with an energy of 200 to 400 keV in the range of doses from 1016 to 3 · 1017 ions/cm2 and after annealing these samples thermally at temperatures from 300 to 550°C. Threshold blistering doses for all the studied materials and annealing temperatures for visualizing structural defects have been determined.

  5. Lessons Learned about Liquid Metal Reactors from FFTF Experience

    SciTech Connect

    Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.; Burke, Thomas M.; Grandy, Christopher

    2016-09-20

    The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens. In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports

  6. Nuclear reactor insulation and preheat system

    DOEpatents

    Wampole, Nevin C.

    1978-01-01

    An insulation and preheat system for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the compartment. An external surface of the compartment or enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair.

  7. An expert system for probabilistic description of loads on space propulsion system structural components

    NASA Technical Reports Server (NTRS)

    Spencer, B. F., Jr.; Hopkins, D. A.

    1988-01-01

    LDEXPT, an expert system that generates probabilistic characterizations of the loads spectra borne by spacecraft propulsion systems' structural components, is found by recent experience at NASA-Lewis to be useful in the cases of components representative of the Space Shuttle Main Engine's turbopumps and fluid transfer ducting. LDEXPT is composed of a knowledge base management system and a rule base management system. The ANLOAD load-modeling module of LDEXPT encompasses three independent probabilistic analysis techniques.

  8. Fabrication of structural components from commercial aluminum alloys using superplastic forming

    NASA Technical Reports Server (NTRS)

    Hales, S. J.; Bales, T. T.; Shinn, J. M.; James, W. F.

    1990-01-01

    SPF technology was used to fabricate structural components from the 7475 Al and 8090 Al-Li commercial alloys. Gas-pressurization cycles were established for SPF three-hat stiffener configurations on the basis of uniaxial data and component-geometry considerations. It is established that higher forming rates than the optimum strain rates selected from the uniaxial data for each alloy could be used in the later stages of forming without reducing SPF components' dimensional conformity. Cavitation was precluded through the use of back pressure during forming.

  9. Fabrication of structural components from commercial aluminum alloys using superplastic forming

    NASA Technical Reports Server (NTRS)

    Hales, S. J.; Bales, T. T.; Shinn, J. M.; James, W. F.

    1990-01-01

    SPF technology was used to fabricate structural components from the 7475 Al and 8090 Al-Li commercial alloys. Gas-pressurization cycles were established for SPF three-hat stiffener configurations on the basis of uniaxial data and component-geometry considerations. It is established that higher forming rates than the optimum strain rates selected from the uniaxial data for each alloy could be used in the later stages of forming without reducing SPF components' dimensional conformity. Cavitation was precluded through the use of back pressure during forming.

  10. Component-based syntheses of trioxacarcin A, DC-45-A1 and structural analogues

    NASA Astrophysics Data System (ADS)

    Magauer, Thomas; Smaltz, Daniel J.; Myers, Andrew G.

    2013-10-01

    The trioxacarcins are polyoxygenated, structurally complex natural products that potently inhibit the growth of cultured human cancer cells. Here we describe syntheses of trioxacarcin A, DC-45-A1 and structural analogues by late-stage stereoselective glycosylation reactions of fully functionalized, differentially protected aglycon substrates. Key issues addressed in this work include the identification of an appropriate means to activate and protect each of the two 2-deoxysugar components, trioxacarcinose A and trioxacarcinose B, as well as a viable sequencing of the glycosidic couplings. The convergent, component-based sequence we present allows for rapid construction of structurally diverse, synthetic analogues that would be inaccessible by any other means, in amounts required to support biological evaluation. Analogues that arise from the modification of four of five modular components are assembled in 11 steps or fewer. The majority of these are found to be active in antiproliferative assays using cultured human cancer cells.

  11. Nonlinear low frequency electrostatic structures in a magnetized two-component auroral plasma

    SciTech Connect

    Rufai, O. R.; Bharuthram, R.; Singh, S. V. Lakhina, G. S.

    2016-03-15

    Finite amplitude nonlinear ion-acoustic solitons, double layers, and supersolitons in a magnetized two-component plasma composed of adiabatic warm ions fluid and energetic nonthermal electrons are studied by employing the Sagdeev pseudopotential technique and assuming the charge neutrality condition at equilibrium. The model generates supersoliton structures at supersonic Mach numbers regime in addition to solitons and double layers, whereas in the unmagnetized two-component plasma case only, soliton and double layer solutions can be obtained. Further investigation revealed that wave obliqueness plays a critical role for the evolution of supersoliton structures in magnetized two-component plasmas. In addition, the effect of ion temperature and nonthermal energetic electron tends to decrease the speed of oscillation of the nonlinear electrostatic structures. The present theoretical results are compared with Viking satellite observations.

  12. Nonlinear low frequency electrostatic structures in a magnetized two-component auroral plasma

    NASA Astrophysics Data System (ADS)

    Rufai, O. R.; Bharuthram, R.; Singh, S. V.; Lakhina, G. S.

    2016-03-01

    Finite amplitude nonlinear ion-acoustic solitons, double layers, and supersolitons in a magnetized two-component plasma composed of adiabatic warm ions fluid and energetic nonthermal electrons are studied by employing the Sagdeev pseudopotential technique and assuming the charge neutrality condition at equilibrium. The model generates supersoliton structures at supersonic Mach numbers regime in addition to solitons and double layers, whereas in the unmagnetized two-component plasma case only, soliton and double layer solutions can be obtained. Further investigation revealed that wave obliqueness plays a critical role for the evolution of supersoliton structures in magnetized two-component plasmas. In addition, the effect of ion temperature and nonthermal energetic electron tends to decrease the speed of oscillation of the nonlinear electrostatic structures. The present theoretical results are compared with Viking satellite observations.

  13. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  14. An Integrated Theory for Predicting the Hydrothermomechanical Response of Advanced Composite Structural Components

    NASA Technical Reports Server (NTRS)

    Chamis, C. C.; Lark, R. F.; Sinclair, J. H.

    1977-01-01

    An integrated theory is developed for predicting the hydrothermomechanical (HDTM) response of fiber composite components. The integrated theory is based on a combined theoretical and experimental investigation. In addition to predicting the HDTM response of components, the theory is structured to assess the combined hydrothermal effects on the mechanical properties of unidirectional composites loaded along the material axis and off-axis, and those of angleplied laminates. The theory developed predicts values which are in good agreement with measured data at the micromechanics, macromechanics, laminate analysis and structural analysis levels.

  15. Size and shape of grain boundary network components and their atomic structures in polycrystalline nanoscale materials

    SciTech Connect

    Xu, Tao; Li, Mo

    2015-10-28

    Microstructure in polycrystalline materials is composed of grain boundary plane, triple junction line, and vertex point. They are the integral parts of the grain boundary network structure and the foundation for the structure-property relations. In polycrystalline, especially nanocrystalline, materials, it becomes increasingly difficult to probe the atomistic structure of the microstructure components directly in experiment due to the size limitation. Here, we present a numerical approach using pair correlation function from atomistic simulation to obtain the detailed information for atomic order and disorder in the grain boundary network in nanocrystalline materials. We show that the atomic structures in the different microstructural components are related closely to their geometric size and shape, leading to unique signatures for atomic structure in microstructural characterization at nanoscales. The dependence varies systematically with the characteristic dimension of the microstructural component: liquid-like disorder is found in vertex points, but a certain order persists in triple junctions and grain boundaries along the extended dimensions of these microstructure components.

  16. Prescribed burning consumes key forest structural components: implications for landscape heterogeneity.

    PubMed

    Holland, Greg J; Clarke, Michael F; Bennett, Andrew F

    2017-04-01

    Prescribed burning to achieve management objectives is a common practice in fire-prone regions worldwide. Structural components of habitat that are combustible and slow to develop are particularly susceptible to change associated with prescribed burning. We used an experimental, "whole-landscape" approach to investigate the effect of differing patterns of prescribed burning on key habitat components (logs, stumps, dead trees, litter cover, litter depth, and understorey vegetation). Twenty-two landscapes (each ~100 ha) were selected in a dry forest ecosystem in southeast Australia. Experimental burns were conducted in 16 landscapes (stratified by burn extent) while six served as untreated controls. We measured habitat components prior to and after burning. Landscape burn extent ranged from 22% to 89% across the 16 burn treatments. With the exception of dead standing trees (no change), all measures of habitat components declined as a consequence of burning. The degree of loss increased as the extent to which a landscape was burned also increased. Prescribed burning had complex effects on the spatial heterogeneity (beta diversity) of structural components within landscapes. Landscapes that were more heterogeneous pre-fire were homogenized by burning, while those that were more homogenous pre-fire tended to display greater differentiation post-burning. Thus, the notion that patch mosaic burning enhances heterogeneity at the landscape-scale depends on prior conditions. These findings have important management implications. Where prescribed burns must be undertaken, effects on important resources can be moderated via control of burn characteristics (e.g., burn extent). Longer-term impacts of prescribed burning will be strongly influenced by the return interval, given the slow rate at which some structural components accumulate (decades to centuries). Management of habitat structural components is important given the critical role they play in (1) provision of habitat

  17. Isolation and characterization of structural components of Aloe vera L. leaf pulp.

    PubMed

    Ni, Y; Turner, D; Yates, K M; Tizard, I

    2004-12-20

    The clear pulp, also known as inner gel, of Aloe vera L. leaf is widely used in various medical, cosmetic and nutraceutical applications. Many beneficial effects of this plant have been attributed to the polysaccharides present in the pulp. However, discrepancies exist regarding the composition of pulp polysaccharide species and an understanding of pulp structure in relation to its chemical composition has been lacking. Thus, we examined pulp structure, isolated structural components and determined their carbohydrate compositions along with analyzing a partially purified pulp-based product (Acemannan hydrogel) used to make Carrisyn hydrogel wound dressing. Light and electron microscopy showed that the pulp consisted of large clear mesophyll cells with a diameter as large as 1000 microm. These cells were composed of cell walls and cell membranes along with a very limited number of degenerated cellular organelles. No intact cellular organelles were found in mesophyll cells. Following disruption of pulp by homogenization, three components were isolated by sequential centrifugation. They were thin clear sheets, microparticles and a viscous liquid gel, which corresponded to cell wall, degenerated cellular organelles and liquid content of mesophyll cells based on morphological and chemical analysis. These three components accounted for 16.2% (+/-3.8), 0.70% (+/-0) and 83.1% of the pulp on a dry weight basis. The carbohydrate composition of each component was distinct; liquid gel contained mannan, microparticles contained galactose-rich polysaccharide(s) and cell walls contained an unusually high level of galacturonic acid (34%, w/w; Gal A). The same three components were also found in Acemannan Hydrogel with mannan as the predominant component. Thus, different pulp structural components are associated with different polysaccharides and thus may potentially be different functionally. These findings may help lay a basis for further studies and development of better

  18. ELECTRONUCLEAR REACTOR

    DOEpatents

    Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

    1960-04-19

    An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

  19. Kinetics of deactivation of catalysts for vinyl acetate synthesis in the fluidized-bed reactor: The optimal loading and distribution of zinc acetate in the porous structure of a support

    SciTech Connect

    Romanchuk, S.V.; Makhlin, V.A.

    1995-03-01

    The deactivation of a catalyst (zinc acetate on activated carbon) including a change of the phase state of the active component is considered. The mechanism and relevant kinetic model of the deactivation are presented. A degree of thermal decomposition of zinc acetate controls the deactivation rate, which depends on the loading and distribution of zinc acetate in the porous structure of a support. A modeling of the process in an industrial reactor is performed with regard to the deactivation, attrition, and loss of a catalyst. Each carbon support has an optimal loading of zinc acetate (equal to the critical value), which provides both a high activity and stability of catalyst operation. The reasons behind the fast deactivation of the commercial catalyst are revealed. The possibility is demonstrated of extending the life time of a catalyst on available carbon supports by a factor of {approximately}2.5, due to the optimal loading and distribution of the active component in the porous support structure.

  20. Seismic performance of non-structural components and contents in buildings: an overview of NZ research

    NASA Astrophysics Data System (ADS)

    Dhakal, Rajesh P.; Pourali, Atefeh; Tasligedik, Ali Sahin; Yeow, Trevor; Baird, Andrew; MacRae, Gregory; Pampanin, Stefano; Palermo, Alessandro

    2016-03-01

    This paper summarizes the research on non-structural elements and building contents being conducted at University of Canterbury in New Zealand. Since the 2010-2011 series of Canterbury earthquakes, in which damage to non-structural components and contents contributed heavily to downtime and overall financial loss, attention to seismic performance and design of non-structural components and contents in buildings has increased exponentially in NZ. This has resulted in an increased allocation of resources to research leading to development of more resilient non-structural systems in buildings that would incur substantially less damage and cause little downtime during earthquakes. In the last few years, NZ researchers have made important developments in understanding and improving the seismic performance of secondary building elements such as partitions, facades, ceilings and contents.

  1. Probabilistic Structural Analysis Methods for select space propulsion system components (PSAM). Volume 2: Literature surveys of critical Space Shuttle main engine components

    NASA Technical Reports Server (NTRS)

    Rajagopal, K. R.

    1992-01-01

    The technical effort and computer code development is summarized. Several formulations for Probabilistic Finite Element Analysis (PFEA) are described with emphasis on the selected formulation. The strategies being implemented in the first-version computer code to perform linear, elastic PFEA is described. The results of a series of select Space Shuttle Main Engine (SSME) component surveys are presented. These results identify the critical components and provide the information necessary for probabilistic structural analysis. Volume 2 is a summary of critical SSME components.

  2. Optical formation of stable waveguiding structures from a photopolymerisable composition with a nonpolymerisable component

    SciTech Connect

    Mensov, Sergei N; Polushtaitsev, Yu V

    2012-06-30

    We report formation of stable dielectric waveguiding structures from a photopolymerisable composition containing a nonpolymerisable component by optical radiation. A computer simulation has shown that the use of nonpolymerisable additives not only retains the self-trapping modes of incident radiation but also provides matching conditions for the synthesised waveguiding structure with standard optical fibres at telecommunication wavelengths. The efficiency of these nonlinear wave processes for connecting single-mode fibres SMF-28 is experimentally confirmed.

  3. Variance component model to account for sample structure in genome-wide association studies.

    PubMed

    Kang, Hyun Min; Sul, Jae Hoon; Service, Susan K; Zaitlen, Noah A; Kong, Sit-Yee; Freimer, Nelson B; Sabatti, Chiara; Eskin, Eleazar

    2010-04-01

    Although genome-wide association studies (GWASs) have identified numerous loci associated with complex traits, imprecise modeling of the genetic relatedness within study samples may cause substantial inflation of test statistics and possibly spurious associations. Variance component approaches, such as efficient mixed-model association (EMMA), can correct for a wide range of sample structures by explicitly accounting for pairwise relatedness between individuals, using high-density markers to model the phenotype distribution; but such approaches are computationally impractical. We report here a variance component approach implemented in publicly available software, EMMA eXpedited (EMMAX), that reduces the computational time for analyzing large GWAS data sets from years to hours. We apply this method to two human GWAS data sets, performing association analysis for ten quantitative traits from the Northern Finland Birth Cohort and seven common diseases from the Wellcome Trust Case Control Consortium. We find that EMMAX outperforms both principal component analysis and genomic control in correcting for sample structure.

  4. Data structure characterization of miltispectral data using principal component and principal factor analysis

    NASA Technical Reports Server (NTRS)

    Lee, Jae K.; Mausel, Paul W.; Lulla, Kamlesh P.

    1989-01-01

    Both principal component analysis (PCA) and principal factor analysis (PFA) were used to analyze an experimental multispectral data structure in terms of common and unique variance. Only the common variance of the multispectral data was associated with the principal factor, while higher-order principal components were associated with both common and unique variance. The unique variance was found to represent small spectral variations within each cover type as well as noise vectors, and was most abundant in the lower-order principal components. The lower-order principal components can be useful in research designed to discriminate minor physical variations within features, and to highlight localized change when using multitemporal-multispectral data. Conversely, PFA of the multispectral data provided an insight into a great potential for discriminating basic land-cover types by excluding the unique variance which was related to the noise and minor spectral variations.

  5. [Scalability and extensionality of innovative Chinese medicine preparation under guidance of component structure theory].

    PubMed

    Yang, Nan; Feng, Liang; Jia, Xiao-Bin

    2016-01-01

    Chinese medicine preparation is a science to study how to make raw material into suitable dosage forms to be used in clinical operations. Its study scope is significantly different from traditional chemical drugs. As is known, the ingredients of Chinese medicine are complex and various, as a result, the composition of the ingredients is not clear and the property is not unified. The pre-treatment process is the key factor to affect the druggability, safety and efficacy of the Chinese medicine. The connotation of Chinese medicine is a huge systematic project, not only including the traditional dosage form design process but also including the components structure optimization process on material basis, components extraction, separation and purification process, components characterization process as well as the pharmacological and toxicological study processes which aim to select suitable dosage form to exert the maximum drug efficacy and evaluate the properties of preparations. Therefore, according to the requirement of modern innovative preparations and based on the integrity and systematicness of Chinese medicine, in this paper we will explore the scalability and extensionality of modern Chinese medicine, including the material basis of Chinese medicine based on component structure theory; separation, refining and purification of components; component structure optimization and network pharmacology and regulation, as well as biopharmaceutics properties of representative components, and we will construct the multi-unit drug delivery system and Chinese medicine multi-dimensional dynamic quality control system. This paper elaborates the scientific connotation and systematicness of modern Chinese preparation, and provides ideas and methods for the development of modern innovative Chinese preparations. Copyright© by the Chinese Pharmaceutical Association.

  6. Structure and Mechanism of the S Component of a Bacterial ECF Transporter

    SciTech Connect

    P Zhang; J Wang; Y Shi

    2011-12-31

    The energy-coupling factor (ECF) transporters, responsible for vitamin uptake in prokaryotes, are a unique family of membrane transporters. Each ECF transporter contains a membrane-embedded, substrate-binding protein (known as the S component), an energy-coupling module that comprises two ATP-binding proteins (known as the A and A' components) and a transmembrane protein (known as the T component). The structure and transport mechanism of the ECF family remain unknown. Here we report the crystal structure of RibU, the S component of the ECF-type riboflavin transporter from Staphylococcus aureus at 3.6-{angstrom} resolution. RibU contains six transmembrane segments, adopts a previously unreported transporter fold and contains a riboflavin molecule bound to the L1 loop and the periplasmic portion of transmembrane segments 4-6. Structural analysis reveals the essential ligand-binding residues, identifies the putative transport path and, with sequence alignment, uncovers conserved structural features and suggests potential mechanisms of action among the ECF transporters.

  7. Silicon Nitride and Silicon Carbide Ceramics Structural Components in Avionics and Space

    NASA Astrophysics Data System (ADS)

    Berroth, Karl

    2014-06-01

    In the paper, Silicon Nitride and silicon carbide components for avionics and space are described. These lightweight stiff and strong materials with low and very low CTE and high thermal conductivity provide means for new designs and higher resolution in passive structures for optical instruments. Material properties and application examples are discussed.

  8. Next Generation Nuclear Plant Structures, Systems, and Components Safety Classification White Paper

    SciTech Connect

    Pete Jordan

    2010-09-01

    This white paper outlines the relevant regulatory policy and guidance for a risk-informed approach for establishing the safety classification of Structures, Systems, and Components (SSCs) for the Next Generation Nuclear Plant and sets forth certain facts for review and discussion in order facilitate an effective submittal leading to an NGNP Combined Operating License application under 10 CFR 52.

  9. [Isolation and partial structural characteristics of major toxic components of Latrodectus pallidus venom].

    PubMed

    Charakha, A R; Shevchenko, L V; Molodkin, A K; Pluzhnikov, K A; Volkova, T M; Grishin, E V

    1997-03-01

    Toxic components of the Latrodectus pallidus spider venom were isolated and characterized. The venom was shown to contain a toxin specific for mammals and at least one insectospecific toxin. Partial amino acid sequences of both toxins were determined, and their high structural homology with previously studied alpha-latrotoxin and alpha-latroinsectotoxin from L. mactans tredecimguttatus was found.

  10. X-ray Crystal Structure of the B Component of Hemolysin BL from Bacillus cereus

    SciTech Connect

    Madegowda,M.; Eswaramoorthy, S.; Burley, S.; Swaminathan, S.

    2008-01-01

    Bacillus cereus Hemolysin BL enterotoxin, a ternary complex of three proteins, is the causative agent of food poisoning and requires all three components for virulence. The X-ray structure of the binding domain of HBL suggests that it may form a pore similar to other soluble channel forming proteins. A putative pathway of pore formation is discussed.

  11. An engineering approach for the application of textile composites to a structural component

    NASA Technical Reports Server (NTRS)

    Baldwin, Jack W.; Gracias, Brian K.; Clark, Steven R.

    1993-01-01

    An engineering approach for the application of textile composites to a structural component is addressed. The main objective is to improve impact resistance of composite blades by using some form of 3-D reinforcement. Project goals, results, and conclusions are discussed.

  12. 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.

  13. Preliminary assessment of the effects of biaxial loading on reactor pressure vessel structural-integrity-assessment technology

    SciTech Connect

    Pennell, W.E.; Bass, B.R.; Bryson, J.W.; Dickson, T.L.; McAfee, W.J.; Merkle, J.G.

    1996-04-01

    Effects of biaxial loading on shallow-flaw fracture toughness were studied to determine potential impact on structural integrity assessment of a reactor pressure vessel (RPV) under pressurized thermal shock (PTS) transient loading and pressure-temperature (PT) loading produced by reactor heatup and cooldown transients. Biaxial shallow-flaw fracture-toughness tests results were also used to determine the parameter controlling fracture in the transition temperature range, and to develop a related dual-parameter fracture-toughness correlation. Shallow-flaw and biaxial loading effects were found to reduce the conditional probability of crack initiation by a factor of nine when the shallow-flaw fracture-toughness K{sub Jc} data set, with biaxial-loading effects adjustments, was substituted in place of ASME Code K{sub Ic} data set in PTS analyses. Biaxial loading was found to reduce the shallow-flaw fracture toughness of RPV steel such that the lower-bound curve was located between ASME K{sub Ic} and K{sub IR} curves. This is relevant to future development of P-T curve analysis procedures. Fracture in shallow-flaw biaxial samples tested in the lower transition temperature range was shown to be strain controlled. A strain-based dual-parameter fracture-toughness correlation was developed and shown to be capable of predicting the effect of crack-tip constraint on fracture toughness for strain-controlled fracture.

  14. Removal of COD and nitrogen from animal food plant wastewater in an intermittently-aerated structured-bed reactor.

    PubMed

    Wosiack, Priscila Arcoverde; Lopes, Deize Dias; Rissato Zamariolli Damianovic, Márcia Helena; Foresti, Eugenio; Granato, Daniel; Barana, Ana Cláudia

    2015-05-01

    This study evaluated the performance of a continuous flow structured-bed reactor in the simultaneous removal of total nitrogen (TN) and chemical oxygen demand (COD) in the effluent from an animal food plant. The reactor had an intermittent aeration system; hydraulic retention time (HRT) of one day; temperature of 30 °C; and recirculation ratio of five times the flow. An experimental central composite rotational delineation (CCRD) type design was used to define the aeration conditions and nitrogen load (factors) to be studied. Response surface methodology was used to analyse the influence of the factors above the results, the removal of TN and COD. It was observed that the aeration factor showed the greatest significance for the results and that the affluent TKN concentration did not have a significant effect, at a 95% level of confidence, on COD removal. Throughout the experiment, the COD/N ratio remained between 3.2 and 3.8. The best results for COD and TN removal, 80% and 88%, respectively, were obtained with 158 min of aeration on a cycle of 180 min and 255 mg L(-1) of Total Kjeldahl Nitrogen (TKN) in the substrate. Copyright © 2015 Elsevier Ltd. All rights reserved.

  15. Reactive sputter magnetron reactor for preparation of thin films and simultaneous in situ structural study by X-ray diffraction.

    PubMed

    Bürgi, J; Neuenschwander, R; Kellermann, G; García Molleja, J; Craievich, A F; Feugeas, J

    2013-01-01

    The purpose of the designed reactor is (i) to obtain polycrystalline and∕or amorphous thin films by controlled deposition induced by a reactive sputtering magnetron and (ii) to perform a parallel in situ structural study of the deposited thin films by X-ray diffraction, in real time, during the whole growth process. The designed reactor allows for the control and precise variation of the relevant processing parameters, namely, magnetron target-to-sample distance, dc magnetron voltage, and nature of the gas mixture, gas pressure and temperature of the substrate. On the other hand, the chamber can be used in different X-ray diffraction scanning modes, namely, θ-2θ scanning, fixed α-2θ scanning, and also low angle techniques such as grazing incidence small angle X-ray scattering and X-ray reflectivity. The chamber was mounted on a standard four-circle diffractometer located in a synchrotron beam line and first used for a preliminary X-ray diffraction analysis of AlN thin films during their growth on the surface of a (100) silicon wafer.

  16. Reactor operation safety information document

    SciTech Connect

    Not Available

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  17. Effect of mechanical disruption on the effectiveness of three reactors used for dilute acid pretreatment of corn stover Part 2: morphological and structural substrate analysis

    PubMed Central

    2014-01-01

    Background Lignocellulosic biomass is a renewable, naturally mass-produced form of stored solar energy. Thermochemical pretreatment processes have been developed to address the challenge of biomass recalcitrance, however the optimization, cost reduction, and scalability of these processes remain as obstacles to the adoption of biofuel production processes at the industrial scale. In this study, we demonstrate that the type of reactor in which pretreatment is carried out can profoundly alter the micro- and nanostructure of the pretreated materials and dramatically affect the subsequent efficiency, and thus cost, of enzymatic conversion of cellulose. Results Multi-scale microscopy and quantitative image analysis was used to investigate the impact of different biomass pretreatment reactor configurations on plant cell wall structure. We identify correlations between enzymatic digestibility and geometric descriptors derived from the image data. Corn stover feedstock was pretreated under the same nominal conditions for dilute acid pretreatment (2.0 wt% H2SO4, 160°C, 5 min) using three representative types of reactors: ZipperClave® (ZC), steam gun (SG), and horizontal screw (HS) reactors. After 96 h of enzymatic digestion, biomass treated in the SG and HS reactors achieved much higher cellulose conversions, 88% and 95%, respectively, compared to the conversion obtained using the ZC reactor (68%). Imaging at the micro- and nanoscales revealed that the superior performance of the SG and HS reactors could be explained by reduced particle size, cellular dislocation, increased surface roughness, delamination, and nanofibrillation generated within the biomass particles during pretreatment. Conclusions Increased cellular dislocation, surface roughness, delamination, and nanofibrillation revealed by direct observation of the micro- and nanoscale change in accessibility explains the superior performance of reactors that augment pretreatment with physical energy. PMID:24690534

  18. Spherical torus fusion reactor

    DOEpatents

    Martin Peng, Y.K.M.

    1985-10-03

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  19. Structural Analysis of the Flagellar Component Proteins in Solution by Small Angle X-Ray Scattering.

    PubMed

    Lee, Lawrence K

    2017-01-01

    Small angle X-ray scattering is an increasingly utilized method for characterizing the shape and structural properties of proteins in solution. The technique is amenable to very large protein complexes and to dynamic particles with different conformational states. It is therefore ideally suited to the analysis of some flagellar motor components. Indeed, we recently used the method to analyze the solution structure of the flagellar motor protein FliG, which when combined with high-resolution snapshots of conformational states from crystal structures, led to insights into conformational transitions that are important in mediating the self-assembly of the bacterial flagellar motor. Here, we describe procedures for X-ray scattering data collection of flagellar motor components, data analysis, and interpretation.

  20. ROCOPT: A user friendly interactive code to optimize rocket structural components

    NASA Technical Reports Server (NTRS)

    Rule, William K.

    1989-01-01

    ROCOPT is a user-friendly, graphically-interfaced, microcomputer-based computer program (IBM compatible) that optimizes rocket components by minimizing the structural weight. The rocket components considered are ring stiffened truncated cones and cylinders. The applied loading is static, and can consist of any combination of internal or external pressure, axial force, bending moment, and torque. Stress margins are calculated by means of simple closed form strength of material type equations. Stability margins are determined by approximate, orthotropic-shell, closed-form equations. A modified form of Powell's method, in conjunction with a modified form of the external penalty method, is used to determine the minimum weight of the structure subject to stress and stability margin constraints, as well as user input constraints on the structural dimensions. The graphical interface guides the user through the required data prompts, explains program options and graphically displays results for easy interpretation.