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Sample records for structural reactor components

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Properties of V-4Cr-4Ti for application as fusion reactor structural components

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1994-08-01

    Vanadium-base alloys are promising candidate materials for application in fusion reactor first-wall and blanket structures because they offer several important advantages, i.e., inherently low irradiation-induced activity, good mechanical properties, good compatibility with lithium, high thermal conductivity, and good resistance to irradiation-induced swelling and damage. As part of a program to screen candidate alloys and develop an optimized vanadium-base alloy, extensive investigations of various V-Ti, V-Cr-Ti, and V-Ti-Si alloys have been conducted after irradiation in lithium in fission reactors. From these investigations, V-4 wt.% Cr-4 wt.% Ti was identified as the most promising alloy. The alloy exhibited attractive mechanical and physical properties that are prerequisites for first-wall and blanket structures, i.e., high tensile strength, high ductility, good creep properties, high impact energy, low ductile-brittle transition temperature before and after irradiation, excellent resistance to irradiation-induced swelling and microstructural instability, and good resistance to corrosion in lithium. In particular, the alloy is virtually immune to irradiation-induced embrittlement, a remarkable property compared to other candidate materials being investigated in the fusion-reactor-materials community. Effects of helium, charged dynamically in simulation of realistic fusion reactor conditions, on tensile, ductile-brittle transition, and swelling properties were insignificant.

  6. Structural mechanics in reactor technology

    SciTech Connect

    Wittmann, F.H.

    1987-01-01

    This series consists of 14 volumes. Each contains several papers. The volume subtitles are: Indexes, Abbreviations, Supplement; Computational Mechanics and Computer-Aided Engineering; Fuel Elements and Assemblies; Experience with Structures and Components in Operating Reactors; Fast Reactor Core and Coolant Circuit Structures; LWR Pressure Components; Fracture Mechanics and NDE; Concrete and Concrete Structures; Extreme Loading and Response of Reactor Containments; Seismic Response Analysis of Nuclear Power Plant Systems; Mechanical and Thermal Problems of Fusion Reactors; Structural Reliability Probabilistic Safety Assessment; and Inelastic Behaviour of Metals and Constitutive Equations.

  7. 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.

  8. Residual life assessment of major light water reactor components: Overview

    SciTech Connect

    Shah, V.N.; MacDonald, P.E.; Amar, A.S.; Bakr, M.H.; Beaudoin, B.F.; Buescher, B.J.; Conley, D.A.; Drahos, F.R.; Gardner, J.B.; Garner, R.W.; Kirkwood, B.J.; Meyer, L.C.; Server, W.L.; Shah, V.N.; Siegel, E.A.; Sinha, U.P.; Ware, A.G. )

    1989-11-01

    This report presents an assessment of the aging (time-dependent degradation) of selected major light water reactor components and structures. The stressors, possible degradation sites and mechanisms, potential failure modes, and current inservice inspection requirements are discussed for eleven major light water reactor components: reactor coolant pumps, pressurized water reactor (PWR) pressurizers, PWR pressurizer surge and spray lines, PWR reactor coolant system charging and safety injection nozzles, PWR feedwater lines, PWR control rod drive mechanisms and reactor internals, boiling water reactor (BWR) containments, BWR feedwater and main steam lines, BWR control rod drive mechanisms and reactor internals, electrical cables and connections, and emergency diesel generators. Unresolved technical issues related to understanding and managing the aging of these major components are identified. 575 refs., 148 figs., 96 tabs.

  9. Solid tags for identifying failed reactor components

    DOEpatents

    Bunch, Wilbur L.; Schenter, Robert E.

    1987-01-01

    A solid tag material which generates stable detectable, identifiable, and measurable isotopic gases on exposure to a neutron flux to be placed in a nuclear reactor component, particularly a fuel element, in order to identify the reactor component in event of its failure. Several tag materials consisting of salts which generate a multiplicity of gaseous isotopes in predetermined ratios are used to identify different reactor components.

  10. 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.

  11. 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.

  12. 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.

  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. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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…

  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

    Greenstreet, B.L.

    1963-12-31

    A system for maintaining the alignment of moderator block structures in reactors is presented. Integral restraining grids are placed between each layer of blocks in the moderator structure, at the top of the uppermost layer, and at the bottom of the lowermost layer. Slots are provided in the top and bottom surfaces of the moderator blocks so as to provide a keying action with the grids. The grids are maintained in alignment by vertical guiding members disposed about their peripheries. (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. 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.

  3. 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)

  4. 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.

  5. 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.

  6. 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)

  7. 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…

  8. 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)

  9. 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.

  10. Reactor component inventory system at FFTF

    SciTech Connect

    Ordonez, C.R.; Redekopp, R.D.; Reed, E.A.

    1985-02-01

    A reliable inventory control system was developed at the Fast Flux Test Facility (FFTF) to keep track of the occupancy of 900 refueling facility locations, to compile historical data on the movement of each reactor assembly, and to simulate assembly moves. The simulate capability is valuable because it allows verification of documents before they are issued for use in the plant, and eliminates the possibility of planning illegal or impossible moves. The system is installed on a UNIVAC 1100 computer and is maintained using a data base management system by Sperry Univac called MAPPER.

  11. 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...

  12. 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).

  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. 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.

  15. 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.

  16. 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.

  17. 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

  18. 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)

  19. 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.

  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. 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.

  2. 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.

  3. Efficient testing of ITER materials and components at the Research Institute of Atomic Reactors` experimental facilities

    SciTech Connect

    Ivanov, V.; Kazakov, V.; Pokrovsky, A.; Shamardin, V.; Melder, R.; Revyakin, Yu.; Sandakov, V.

    1995-12-31

    The Research Institute of Atomic Reactors (RIAR) of the State Scientific Centre of the Russian Federation has carried out reactor tests of fusion reactor materials and components. RIAR contains an ideal complex of installations, experimental setups, and diagnostics for such investigations. It includes several different types of reactors, including a fast neutron reactor, a high-flux intermediate-neutron SM-3 reactor, a intermediate-neutron loop reactor, and two RBT-type reactors, and a hot cells complex with remote handling facilities to allow study of the physical-mechanical properties, structure, and elemental composition of irradiated materials. RIAR has carried out a number of initial experiments, including testing of copper and vanadium alloys, electro-insulative coatings, steels, ceramics, diagnostic systems materials, and in-core and hot cell set-ups for divertor mock-up testing, and has collaborative efforts underway with the Scientific Research Institute Electrophysical Apparatus-St. Petersburg (SRIEA), Oak Ridge National Laboratory (ORNL), Argonne National Laboratory (ANL), Red Star, the Institute of Physics and Power engineering (IPPE), the Scientific Research Institute of Inorganic Materials (SRIIM), and Pacific Northwest Laboratory (PNL).

  4. Reactor materials program process water component failure probability

    SciTech Connect

    Daugherty, W. L.

    1988-04-12

    The maximum rate loss of coolant accident for the Savannah River Production Reactors is presently specified as the abrupt double-ended guillotine break (DEGB) of a large process water pipe. This accident is not considered credible in light of the low applied stresses and the inherent ductility of the piping materials. The Reactor Materials Program was initiated to provide the technical basis for an alternate, credible maximum rate LOCA. The major thrust of this program is to develop an alternate worst case accident scenario by deterministic means. In addition, the probability of a DEGB is also being determined; to show that in addition to being mechanistically incredible, it is also highly improbable. The probability of a DEGB of the process water piping is evaluated in two parts: failure by direct means, and indirectly-induced failure. These two areas have been discussed in other reports. In addition, the frequency of a large bread (equivalent to a DEGB) in other process water system components is assessed. This report reviews the large break frequency for each component as well as the overall large break frequency for the reactor system.

  5. 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.

  6. 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.

  7. Component failures at pressurized water reactors. Final report

    SciTech Connect

    Reisinger, M.F.

    1980-10-01

    Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis.

  8. Tritium retention in fusion reactor plasma facing components

    SciTech Connect

    Langley, R.A.

    1995-03-01

    The IAEA has proposed a coordinated research program to address tritium retention and release in fusion reactor plasma facing components. This program will address materials which are mainly of interest to the design and construction of ITER, namely beryllium, carbon based materials and medium and high-Z metals, e.g. tungsten, vanadium and molybdenum, but will not be limited to these materials. Experimental data are needed for: recycling models, tritium inventory estimates, tritium permeation calculations and hydrogen embrittlement characterization. The ultimate use of the data would be to influence the formation of models for use by fusion reactor designers. Judicious material choices must be made by the designers and accurate predictive codes are required in order to make these choices. The proposed coordinated research program will provide a forum for discussions between experimentalists, theoreticians, modelers and reactor designers, provide financial support for relevant research projects and collect and evaluate experimental and theoretical data. This paper briefly reviews existing data, addresses the data gaps and points out experiments designed to obtain the needed data. 18 refs., 3 figs., 1 tab.

  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. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Materials for the plasma-facing components of fusion reactors

    NASA Astrophysics Data System (ADS)

    Bolt, H.; Barabash, V.; Krauss, W.; Linke, J.; Neu, R.; Suzuki, S.; Yoshida, N.; ASD. E. X. Upgrade Team

    2004-08-01

    During reactor operation the plasma-facing materials have to fulfil very complex and sometimes contradicting requirements. At present, tungsten shows the highest promise as plasma-facing material. Experiments in the ASDEX Upgrade tokamak indicate that plasma operation is feasible with walls and divertor surfaces mostly covered with tungsten. Thick tungsten coatings have been deposited by plasma spraying on EUROFER first wall mock-ups and show good adhesion and stability. The performance of tungsten surfaces under intense transient thermal loads is another critical issue, since the formation of a melt layer may favour the generation of highly activated dust particles. Work on `nanocrystalline' tungsten shall improve the mechanical properties under neutron irradiation which is especially important for designs, where tungsten has also to fulfil structural functions. Alternative divertor heat sink materials with very high thermal conductivity like SiC-fibre reinforced copper composites are presently being developed and should allow operation at reactor relevant coolant temperatures.

  15. Evaluation of aging degradation of structural components

    SciTech Connect

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

    1992-03-01

    Irradiation embrittlement of the neutron shield tank (NST) A212 Grade B steel from the Shippingport reactor, as well as thermal embrittlement of CF-8 cast stainless steel components from the Shippingport and KRB reactors, has been characterized. Increases in Charpy transition temperature (CTT), yield stress, and hardness of the NST material in the low-temperature low-flux environment are consistent with the test reactor data for irradiations at < 232{degrees}C. The shift in CTT is not as severe as that observed in surveillance samples from the High Flux Isotope Reactor (HFIR): however, it shows very good agreement with the results for HFIR A212-B steel irradiated in the Oak Ridge Research Reactor. The results indicate that fluence rate has not effect on radiation embrittlement at rates as low as 2 {times} 10{sup 8} n/cm{sup 2}{center dot}s at the low operating temperature of the Shippingport NST, i.e., 55{degrees}C. This suggest that radiation damage in Shippingport NST and HFIR surveillance samples may be different because of the neutron spectra and/or Cu and Ni content of the two materials. Cast stainless steel components show relatively modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength. Correlations for estimating mechanical properties of cast stainless steels predict accurate or slightly conservative 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 predict the mechanical properties of the Ringhals 2 reactor hot- and crossover-leg elbows (CF-8M steel) after service of {approx}15 y.

  16. Evaluation of aging degradation of structural components

    SciTech Connect

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

    1992-03-01

    Irradiation embrittlement of the neutron shield tank (NST) A212 Grade B steel from the Shippingport reactor, as well as thermal embrittlement of CF-8 cast stainless steel components from the Shippingport and KRB reactors, has been characterized. Increases in Charpy transition temperature (CTT), yield stress, and hardness of the NST material in the low-temperature low-flux environment are consistent with the test reactor data for irradiations at < 232{degrees}C. The shift in CTT is not as severe as that observed in surveillance samples from the High Flux Isotope Reactor (HFIR): however, it shows very good agreement with the results for HFIR A212-B steel irradiated in the Oak Ridge Research Reactor. The results indicate that fluence rate has not effect on radiation embrittlement at rates as low as 2 {times} 10{sup 8} n/cm{sup 2}{center_dot}s at the low operating temperature of the Shippingport NST, i.e., 55{degrees}C. This suggest that radiation damage in Shippingport NST and HFIR surveillance samples may be different because of the neutron spectra and/or Cu and Ni content of the two materials. Cast stainless steel components show relatively modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength. Correlations for estimating mechanical properties of cast stainless steels predict accurate or slightly conservative 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 predict the mechanical properties of the Ringhals 2 reactor hot- and crossover-leg elbows (CF-8M steel) after service of {approx}15 y.

  17. 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

  18. Structured functional principal component analysis.

    PubMed

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

    2015-03-01

    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

  19. 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....

  20. 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.

  1. 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

  2. Structural interaction of cytoskeletal components.

    PubMed

    Schliwa, M; van Blerkom, J

    1981-07-01

    Three-dimensional cytoskeletal organization of detergent-treated epithelial African green monkey kidney cells (BSC-1) and chick embryo fibroblasts was studied in whole-mount preparations visualized in a high voltage electron microscope. Stereo images are generated at both low and high magnification to reveal both overall cytoskeletal morphology and details of the structural continuity of different filament types. By the use of an improved extraction procedure in combination with heavy meromyosin subfragment 1 decoration of actin filaments, several new features of filament organization are revealed that suggest that the cytoskeleton is a highly interconnected structural unit. In addition to actin filaments, intermediate filaments, and microtubules, a new class of filaments of 2- to 3-nm diameter and 30- to 300-nm length that do not bind heavy merymyosin is demonstrated. They form end-to-side contacts with other cytoskeletal filaments, thereby acting as linkers between various fibers, both like (e.g., actin- actin) and unlike (e.g., actin-intermediate filament, intermediate filament-microtubule). Their nature is unknown. In addition to 2- to 3-nm filaments, actin filaments are demonstrated to form end-to-side contacts with other filaments. Y-shaped actin filament "branches" are observed both in the cell periphery close to ruffles and in more central cell areas also populated by abundant intermediate filaments and microtubules. Arrowhead complexes formed by subfragment 1 decoration of actin filaments point towards the contact site. Actin filaments also form end-to-side contacts with microtubules and intermediate filaments. Careful inspection of numerous actin-microtubule contacts shows that microtubules frequently change their course at sites of contact. A variety of experimentally induced modifications of the frequency of actin-microtubule contacts can be shown to influence the course of microtubules. We conclude that bends in microtubules are imposed by structural

  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. 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.

  5. Cryogenic system component development for fusion experimental reactor at JAERI

    SciTech Connect

    Kato, T.; Kamiya, S.; Tada, E.; Hiyama, T.; Kawano, K.; Shimamoto, S.

    1986-11-01

    A supercritical helium (SHE) circulation pump, a jet pump, and a cold compressor were designed and manufactured as the first step of cryogenic component development for a large-scale cryogenic system which is required for the Fusion Experimental Reactor (FER). The SHE circulation pump achieved 320-g/s flow rate with an 0.88-MPa pressure head at 4.6 K, making it the biggest cold pump in the world. The jet pump's mass flow ratio was about 1.0 with an 0.07-MPa pressure head at about 10 K. The cold compressor was successfully operated with an inlet vapor pressure of 0.053 MPa (3.7 K), and outlet pressure of 0.12 MPa, and a mass flow rate of 60 g/s. The designs and test results are described in this paper.

  6. 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

  7. Probabilistic evaluation of SSME structural components

    NASA Astrophysics Data System (ADS)

    Rajagopal, K. R.; Newell, J. F.; Ho, H.

    1991-05-01

    The application is described of Composite Load Spectra (CLS) and Numerical Evaluation of Stochastic Structures Under Stress (NESSUS) family of computer codes to the probabilistic structural analysis of four Space Shuttle Main Engine (SSME) space propulsion system components. These components are subjected to environments that are influenced by many random variables. The applications consider a wide breadth of uncertainties encountered in practice, while simultaneously covering a wide area of structural mechanics. This has been done consistent with the primary design requirement for each component. The probabilistic application studies are discussed using finite element models that have been typically used in the past in deterministic analysis studies.

  8. 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

  9. 76 FR 16842 - Request for a License To Export Reactor Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-25

    ... systems, related reactors. operation of AP- equipment, and 1000 (design) spare parts. nuclear reactors... 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...

  10. 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 reactors... COMMISSION Request for a License To Export Reactor Components Pursuant to 10 CFR 110.70 (b) ``Public...

  11. 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 .

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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.

  17. 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

  18. Experience with non-fuel-bearing components in LWR (light-water reactor) fuel systems

    SciTech Connect

    Bailey, W.J.; Berting, F.M.

    1990-12-01

    Many non-fuel-bearing components are so closely associated with the spent fuel assemblies that their integrity and behavior must be taken into consideration with the fuel assemblies, when handling spent fuel of planning waste management activities. Presented herein is some of the experience that has been gained over the past two decades from non-fuel-bearing components in light-water reactors (LWRs), both pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). Among the most important of these components are the control rod systems, the absorber and burnable poison rods, and the fuel assembly channels. 15 refs., 5 figs., 2 tabs.

  19. 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.

  20. 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.

  1. Associated neural network independent component analysis structure

    NASA Astrophysics Data System (ADS)

    Kim, Keehoon; Kostrzweski, Andrew

    2006-05-01

    Detection, classification, and localization of potential security breaches in extremely high-noise environments are important for perimeter protection and threat detection both for homeland security and for military force protection. Physical Optics Corporation has developed a threat detection system to separate acoustic signatures from unknown, mixed sources embedded in extremely high-noise environments where signal-to-noise ratios (SNRs) are very low. Associated neural network structures based on independent component analysis are designed to detect/separate new acoustic sources and to provide reliability information. The structures are tested through computer simulations for each critical component, including a spontaneous detection algorithm for potential threat detection without a predefined knowledge base, a fast target separation algorithm, and nonparametric methodology for quantified confidence measure. The results show that the method discussed can separate hidden acoustic sources of SNR in 5 dB noisy environments with an accuracy of 80%.

  2. 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.

  3. Service evaluation of aircraft composite structural components

    NASA Technical Reports Server (NTRS)

    Brooks, W. A., Jr.; Dow, M. B.

    1973-01-01

    The advantages of the use of composite materials in structural applications have been identified in numerous engineering studies. Technology development programs are underway to correct known deficiencies and to provide needed improvements. However, in the final analysis, flight service programs are necessary to develop broader acceptance of, and confidence in, any new class of materials such as composites. Such flight programs, initiated by NASA Langley Research Center, are reviewed. These programs which include the selectively reinforced metal and the all-composite concepts applied to both secondary and primary aircraft structural components, are described and current status is indicated.

  4. 78 FR 57904 - Request for a License To Export; Reactor Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-20

    ..., and 1000 (design) spare parts. nuclear reactors. Dated this 16th day of September 2013 in Rockville... 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...

  5. Generic component failure data base for light water and liquid sodium reactor PRAs (probabilistic risk assessments)

    SciTech Connect

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

    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. 21 refs., 9 tabs.

  6. Packaging of structural health monitoring components

    NASA Astrophysics Data System (ADS)

    Kessler, Seth S.; Spearing, S. Mark; Shi, Yong; Dunn, Christopher T.

    2004-07-01

    Structural Health Monitoring (SHM) technologies have the potential to realize economic benefits in a broad range of commercial and defense markets. Previous research conducted by Metis Design and MIT has demonstrated the ability of Lamb waves methods to provide reliable information regarding the presence, location and type of damage in composite specimens. The present NSF funded program was aimed to study manufacturing, packaging and interface concepts for critical SHM components. The intention is to be able to cheaply manufacture robust actuating/sensing devices, and isolate them from harsh operating environments including natural, mechanical, or electrical extremes. Currently the issues related to SHM system durability have remained undressed. During the course of this research several sets of test devices were fabricated and packaged to protect the piezoelectric component assemblies for robust operation. These assemblies were then tested in hot and wet conditions, as well as in electrically noisy environments. Future work will aim to package the other supporting components such as the battery and wireless chip, as well as integrating all of these components together for operation. SHM technology will enable the reduction or complete elimination of scheduled inspections, and will allow condition-based maintenance for increased reliability and reduced overall life-cycle costs.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Flaw assessment procedure for high temperature reactor components

    SciTech Connect

    Ainsworth, R.A. . Berkeley Nuclear Labs.); Ruggles, M.B. ); Takahashi, Y. . Komae Research Lab.)

    1990-01-01

    An interim high-temperature flaw assessment procedure is described. This is a result of a collaborative effort between Electric Power Research Institute in the USA, Central Research Institute of Electric Power Industry in Japan, and Nuclear Electric plc in the UK. The procedure addresses preexisting defects subject to creep-fatigue loading conditions. Laws employed to calculate the crack growth per cycle are defined in terms of fracture mechanics parameters and constants related to the component material. The crack growth laws may be integrated to calculate the remaining life of a component or to predict the amount of crack extension in a given period. Fatigue and creep crack growth per cycle are calculated separately, and the total crack extension is taken as the simple sum of the two contributions. An interaction between the two propagation modes is accounted for in the material properties in the separate calculations. In producing the procedure, limitations of the approach have been identified. Some of these limitations are to be addressed in an extension of the current collaborative program. 20 refs.

  14. 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.

  15. 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

  16. Power Systems Development Facility: Performance and development of components in the transport reactor train

    SciTech Connect

    Powell, C.A.; Vimalchand, P.; Leonard, R.F.

    1998-12-31

    The Power Systems Development Facility (PSDF) will develop and demonstrate advanced power generation technologies and system components needed to improve process reliability. This paper will provide an introduction to the PSDF and discuss in detail the operation and performance of the M.W. Kellogg Company`s (MWK) Transport reactor train system components. There will also be brief discussions on the operation and performance of the Transport reactor and the Particulate Collection Device (PCD). Discussions will focus on the major operational challenges faced during the commissioning and operation of various components and the significant equipment modifications that were made to improve the reliability and performance. These include: modifications to the pulverizers, corrective actions taken to the transport air and recycle gas systems, improvements to the process gas analysis system, and changes to the steam generation package. Also included are operational findings of the particle disengagement and collection system, experiences with solids handling systems, and continued development of the reactor`s startup burner, pressure letdown valve, process air systems and impacts of corrosion downstream of the PCD. Much can be inferred from the experiences gained at the PSDF as to the impact each component or system had on the successful operation of the MWK Transport reactor train and similar technologies in the future.

  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. 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.

  20. Progress Towards Prognostic Health Management of Passive Components in Advanced Small Modular Reactors

    SciTech Connect

    Meyer, Ryan M.; Ramuhalli, Pradeep; Hirt, Evelyn H.; Pardini, Allan F.; Suter, Jonathan D.; Prowant, Matthew S.

    2014-08-01

    Sustainable nuclear power to promote energy security and to reduce greenhouse gas emissions are two key national energy priorities. The development of deployable small modular reactors (SMRs) is expected to support these objectives by developing technologies that improve the reliability, sustain safety, and improve affordability of new reactors. Advanced SMRs (AdvSMRs) refer to a specific class of SMRs and are based on modularization of advanced reactor concepts. Prognostic health management (PHM) systems can benefit both the safety and economics of deploying AdvSMRs and can play an essential role in managing the inspection and maintenance of passive components in AdvSMR systems. This paper describes progress on development of a prototypic PHM system for AdvSMR passive components, with thermal creep chosen as the target degradation mechanism.

  1. 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 components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear...

  2. 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 components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear...

  3. 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

    ... supplementing a notice published in the Federal Register on March 20, 2012 (77 FR 16270), that requested public...; email: Evelyn.Gettys@nrc.gov . SUPPLEMENTARY INFORMATION: On March 20, 2012 (77 FR 16270), the NRC... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized...

  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 components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear...

  5. 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 components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear...

  6. 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 components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear...

  7. 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

  8. 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.

  9. 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.

  10. 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.

  11. Estimators for variance components in structured stair nesting models

    NASA Astrophysics Data System (ADS)

    Monteiro, Sandra; Fonseca, Miguel; Carvalho, Francisco

    2016-06-01

    The purpose of this paper is to present the estimation of the components of variance in structured stair nesting models. The relationship between the canonical variance components and the original ones, will be very important in obtaining that estimators.

  12. 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).

  13. Application of component mode synthesis in structural dynamics

    NASA Technical Reports Server (NTRS)

    Craig, R. R.

    1986-01-01

    The principal analytical techniques used for component mode synthesis (CMS) of undamped systems and their application to structural dynamics are discussed. In the CMS, a system is divided into components or substructures, and for each of these components, the number of degrees of freedom is reduced by expressing the physical coordinates in terms of a reduced set of component modal coordinates. Among a number of component modes, a new form of component mode, called an applied force attachment mode, is described. Consideration is given to literature studies of damped structures and recent combined analytical/experimental studies.

  14. Structural components and architectures of RNA exosomes.

    PubMed

    Januszyk, Kurt; Lima, Christopher D

    2010-01-01

    A large body of structural work conducted over the past ten years has elucidated mechanistic details related to 3' to 5' processing and decay of RNA substrates by the RNA exosome. This chapter will focus on the structural organization of eukaryotic exosomes and their evolutionary cousins in bacteria and archaea with an emphasis on mechanistic details related to substrate recognition and to 3' to 5' phosphorolytic exoribonucleolytic activities of bacterial and archaeal exosomes as well as the hydrolytic exoribonucleolytic and endoribonucleolytic activities of eukaryotic exosomes. These points will be addressed in large part through presentation of crystal structures ofphosphorolytic enzymes such as bacterial RNase PH, PNPase and archaeal exosomes and crystal structures ofthe eukaryotic exosome and exosome sub-complexes in addition to standalone structures of proteins that catalyze activities associated with the eukaryotic RNA exosome, namely Rrp44, Rrp6 and their bacterial counterparts. PMID:21618871

  15. 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.

  16. High-Resolution Crack Imaging Reveals Degradation Processes in Nuclear Reactor Structural Materials

    SciTech Connect

    Olszta, Matthew J.; Schreiber, Daniel K.; Thomas, Larry E.; Bruemmer, Stephen M.

    2012-04-01

    Corrosion and cracking represent critical failure mechanisms for structural materials in many applications. Although a crack can often be seen with the unaided eye, higher resolution imaging techniques are required to understand the nature of the crack tips and underlying degradation processes. Researchers at Pacific Northwest National Laboratory (PNNL) employ a suite of microscopy techniques and site-specific material sampling to analyze corrosion and crack structures, producing images and compositional analyses with near-atomic spatial resolution. The samples are cracked components removed from commercial light-water reactor service or laboratory samples tested in simulated reactor environments.

  17. 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.

  18. 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.

  19. A kinetic model for impact/sliding wear of pressurized water reactor internal components: Application to rod cluster control assemblies

    SciTech Connect

    Zbinden, M.; Durbec, V.

    1996-12-01

    Certain internal components of Pressurized Water Reactors are damaged by wear when subjected to vibration induced by flow. In order to enable predictive calculation of such wear, one must have a model which takes account reliably of real damages. The modelling of wear represents a final link in a succession of numerical calculations which begins by the determination of hydraulic excitations induced by the flow. One proceeds, then, in the dynamic response calculation of the structure to finish up with an estimation of volumetric wear and of the depth of wear scars. A new concept of industrial wear model adapted to components of nuclear plants is proposed. Its originality is to be supported, on one hand, by experimental results obtained via wear machines of relatively short operational times, and, on the other hand, by the information obtained from the operating feedback over real wear kinetics of the reactors components. The proposed model is illustrated by an example which correspond to a specific real situation. The determination of the coefficients permitting to cover all assembly of configurations and the validation of the model in these configurations have been the object of the most recent work.

  20. 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.

  1. 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.

  2. Electronics speckle interferometry applications for NDE of spacecraft structural components

    NASA Astrophysics Data System (ADS)

    Rao, M. V.; Samuel, R.; Ananthan, A.; Dasgupta, S.; Nair, P. S.

    2008-09-01

    The spacecraft components viz., central cylinder, deck plates, solar panel substrates, antenna reflectors are made of aluminium/composite honeycomb sandwich construction. Detection of these defects spacecraft structural components is important to assess the integrity of the spacecraft structure. Electronic Speckle Interferometry (ESI) techniques identify the defects as anomalous regions in the interferometric fringe patterns of the specklegram while the component is suitably stressed to give rise to differential displacement/strain around the defective region. Calibration studies, different phase shifting methods associated with ESI and the development of a prototype Twin Head ESSI System (THESSIS) and its use for the NDE of a typical satellite structural component are presented.

  3. 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.

  4. The universal scissor component: Optimization of a reconfigurable component for deployable scissor structures

    NASA Astrophysics Data System (ADS)

    Alegria Mira, Lara; Thrall, Ashley P.; De Temmerman, Niels

    2016-02-01

    Deployable scissor structures are well equipped for temporary and mobile applications since they are able to change their form and functionality. They are structural mechanisms that transform from a compact state to an expanded, fully deployed configuration. A barrier to the current design and reuse of scissor structures, however, is that they are traditionally designed for a single purpose. Alternatively, a universal scissor component (USC)-a generalized element which can achieve all traditional scissor types-introduces an opportunity for reuse in which the same component can be utilized for different configurations and spans. In this article, the USC is optimized for structural performance. First, an optimized length for the USC is determined based on a trade-off between component weight and structural performance (measured by deflections). Then, topology optimization, using the simulated annealing algorithm, is implemented to determine a minimum weight layout of beams within a single USC component.

  5. Nonthermal Components in the Large Scale Structure

    NASA Astrophysics Data System (ADS)

    Miniati, Francesco

    2004-12-01

    I address the issue of nonthermal processes in the large scale structure of the universe. After reviewing the properties of cosmic shocks and their role as particle accelerators, I discuss the main observational results, from radio to γ-ray and describe the processes that are thought be responsible for the observed nonthermal emissions. Finally, I emphasize the important role of γ-ray astronomy for the progress in the field. Non detections at these photon energies have already allowed us important conclusions. Future observations will tell us more about the physics of the intracluster medium, shocks dissipation and CR acceleration.

  6. 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.

  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. 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

  9. 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.

  10. 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.

  11. 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

  12. Block-Krylov component synthesis method for structural model reduction

    NASA Technical Reports Server (NTRS)

    Craig, Roy R., Jr.; Hale, Arthur L.

    1988-01-01

    A new analytical method is presented for generating component shape vectors, or Ritz vectors, for use in component synthesis. Based on the concept of a block-Krylov subspace, easily derived recurrence relations generate blocks of Ritz vectors for each component. The subspace spanned by the Ritz vectors is called a block-Krylov subspace. The synthesis uses the new Ritz vectors rather than component normal modes to reduce the order of large, finite-element component models. An advantage of the Ritz vectors is that they involve significantly less computation than component normal modes. Both 'free-interface' and 'fixed-interface' component models are derived. They yield block-Krylov formulations paralleling the concepts of free-interface and fixed-interface component modal synthesis. Additionally, block-Krylov reduced-order component models are shown to have special disturbability/observability properties. Consequently, the method is attractive in active structural control applications, such as large space structures. The new fixed-interface methodology is demonstrated by a numerical example. The accuracy is found to be comparable to that of fixed-interface component modal synthesis.

  13. Probabilistic structural analysis methods for space propulsion system components

    NASA Technical Reports Server (NTRS)

    Chamis, C. C.

    1986-01-01

    The development of a three-dimensional inelastic analysis methodology for the Space Shuttle main engine (SSME) structural components is described. The methodology is composed of: (1) composite load spectra, (2) probabilistic structural analysis methods, (3) the probabilistic finite element theory, and (4) probabilistic structural analysis. The methodology has led to significant technical progress in several important aspects of probabilistic structural analysis. The program and accomplishments to date are summarized.

  14. Probabilistic structural analysis methods for space propulsion system components

    NASA Technical Reports Server (NTRS)

    Chamis, Christos C.

    1987-01-01

    The development of a three-dimensional inelastic analysis methodology for the Space Shuttle main engine (SSME) structural components is described. The methodology is composed of: (1) composite load spectra, (2) probabilistic structural analysis methods, (3) the probabilistic finite element theory, and (4) probabilistic structural analysis. The methodology has led to significant technical progress in several important aspects of probabilistic structural analysis. The program and accomplishments to date are summarized.

  15. 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.

  16. Decontamination of liquid-metal fast breeder reactor components for reuse; The French experience

    SciTech Connect

    Michaille, P. ); Moroni, J.C. ); Lambert, I. )

    1991-02-01

    Decontamination of stainless steel liquid-metal fast breeder reactor components for reuse in France began with the decontamination of Rapsodie components. At that time, dilute phosphoric acid was used. To cope with additional irradiated components after Phenix came into operation, an extensive study was performed, which led to the selection of a procedure involving two baths. The first bath, alkaline permanganate (AP), is applied for 3 h; the second bath, sulfo-phosphoric acid (SP), is applied for 6 h, both at 60{degrees}C. Up to three cycles are repeated until the residual dose rate is sufficiently low. Eight intermediate heat exchangers (IHXs) and two primary pumps from Phenix were decontaminated using this method. This paper reports that because SP can pickle only a limited depth ({approximately} 3{mu}m), due to the passivation effect of phosphoric acid, and because of the waste treatment problems associated with phosphates, new solutions were explored. One possibility involves improvement of the AP-SP procedure: In the SPm procedure, the AP bath is omitted and the phosphoric concentration is reduced by a factor of 4. A second approach is the use of a new formula, called SECA, a mixture of maleic and citric acid used in reducing conditions (imposed by hydrazine). Since the Phenix and Superphenix waste treatment facilities are not designed to reprocess maleic-citric acid, only the SPm procedure has been used on reactor components. A low-contaminated IHX from Rapsodie served as a test benchmark, not only for the decontamination procedure, but also for the requalification criteria, before the SPm procedure was applied to a highly contaminated IHX from Phenix. Recent results are presented.

  17. Sensor modules for structural health monitoring and reliability of components

    NASA Astrophysics Data System (ADS)

    Kroening, Michael; Berthold, Axel; Meyendorf, Norbert

    2005-05-01

    Safety and availability of ageing infrastructures require periodic or continuous monitoring of the structure"s integrity. Innovative design criteria for new infrastructure components may allow material and energy conservation if components are continuously monitored by using advanced sensor systems. This concept for recurring Structural Health Monitoring will replace a significant part of conventional NDE by new maintenance concepts. The goal consists in sensor networks based on advanced principles of testing technology with integrated signal/data processing and data communication. NDE modeling is required for the quantification of measurement results. Finally, a decision on the integrity of the structure based on sensor results requires detailed knowledge about material behavior and modeling capacity for materials and components. IZFP has developed sensor concepts for complex solutions applicable to Structural Health Monitoring for different applications. These applications include railroad inspection, aircraft inspection, inspection of wind energy systems, power electric switches and micro gas valves. Basic concepts and applications of sensor networks will be presented.

  18. Characterization of damped structural connections for multi-component systems

    NASA Technical Reports Server (NTRS)

    Lawrence, Charles; Huckelbridge, Arthur A.

    1989-01-01

    The inability to model connections adequately has historically limited the ability to predict overall system dynamic response. Connections between structural components are often mechanically complex and difficult to accurataely model analytically. Improved analytical models for connections are needed to improve system dynamic predictions. This study explores combining Component Mode Synthesis methods for coupling structural components with Parameter Identification procedures for improving the analytical modeling of the connections. Improvements in the connection stiffness and damping properties are computed in terms of physical parameters so the physical characteristics of the connections can be better understood, in addition to providing improved input for the system model.

  19. Characterization of damped structural connections for multi-component systems

    NASA Technical Reports Server (NTRS)

    Lawrence, Charles; Huckelbridge, Arthur A.

    1988-01-01

    The inability to model connections adequately has historically limited the ability to predict overall system dynamic response. Connections between structural components are often mechanically complex and difficult to accurately model analytically. Improved analytical models for connections are needed to improve system dynamic predictions. This study explores combining Component Mode Synthesis methods for coupling structural components with Parameter Identification procedures for improving the analytical modeling of the connections. Improvements in the connection stiffness and damping properties are computed in terms of physical parameters so the physical characteristics of the connections can be better understood, in addition to providing improved input for the system model.

  20. 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.

  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. 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

  3. NASA service experience with composite components. [for aircraft structures

    NASA Technical Reports Server (NTRS)

    Dexter, H. B.; Chapman, A. J.

    1980-01-01

    NASA Langley has been active in sponsoring flight service programs with advanced composites during the past decade. A broad data base and confidence in the durability of composite structures are being developed. Flight service experience is reported for more than 140 composite aircraft components with up to 8 years service and almost two million successful component flight hours. Composite components are being evaluated on Boeing, Douglas, and Lockheed transport aircraft. Components are currently under development for service evaluation on Bell and Sikorsky helicopters. Design concepts and inspection and maintenance results are reported for components currently in service. Components under development in the NASA Aircraft Energy Efficiency (ACEE) program are discussed. Results of flight, outdoor ground, and controlled laboratory environmental tests on composite materials used in the flight service programs are also presented.

  4. 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.

  5. 10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Issuance, Limitations, and Conditions of Licenses and Construction Permits §...

  6. 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

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Use of laser flow visualization techniques in reactor component thermal-hydraulic studies

    SciTech Connect

    Oras, J.J.; Kasza, K.E.

    1984-01-01

    To properly design reactor components, an understanding of the various thermal hydraulic phenomena, i.e., thermal stratification flow channeling, recirculation regions, shear layers, etc., is necessary. In the liquid metal breeder reactor program, water is commonly used to replace sodium in experimental testing to facilitate the investigations, (i.e., reduce cost and allow fluid velocity measurement or flow pattern study). After water testing, limited sodium tests can be conducted to validate the extrapolation of the water results to sodium. This paper describes a novel laser flow visualization technique being utilized at ANL together with various examples of its use and plans for further development. A 3-watt argon-ion laser, in conjunction with a cylindrical opticallens, has been used to create a thin (approx. 1-mm) intense plane of laser light for the illuminiation of various flow tracers in precisely defined regions of interest within a test article having windows. Both fluorescing dyes tuned to the wavelength of the laser light (to maximize brightness and sharpness of flow image) and small (< 0.038-mm, 0.0015-in. dia.) opaque, nearly neutrally buoyant polystyrene spheres (to ensure that the particles trace out the fluid motion) have been used as flow tracers.

  12. 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.

  13. 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.

  14. Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

    NASA Astrophysics Data System (ADS)

    Murty, K. L.; Charit, I.

    2008-12-01

    Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

  15. 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.

  16. 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

  17. 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

  18. Probabilistic structural analysis methods for critical SSME propulsion components

    NASA Technical Reports Server (NTRS)

    Chamis, C. C.

    1986-01-01

    The development of a three-dimensional inelastic analysis methodology for the Space Shuttle main engine (SSME) structural components is described. The methodology is composed of: (1) composite load spectra, (2) probabilistic structural analysis methods, (3) the probabilistic finite element theory, and (4) probabilistic structural analysis. The progress in the development of generic probabilistic models for various individual loads which consist of a steady state load, a periodic load, a random load, and a spike, is discussed. The capabilities of the Numerical Evaluation of Stochastic Structures Under Stress finite element code designed for probabilistic structural analysis of the SSME are examined. Variation principles for formulation probabilistic finite elements and a structural analysis for evaluating the geometric and material properties tolerances on the structural response of turbopump blades are being designed.

  19. Biological Insights from Structures of Two-Component Proteins

    PubMed Central

    Gao, Rong; Stock, Ann M.

    2013-01-01

    Two-component signal transduction based on phosphotransfer from a histidine protein kinase to a response regulator protein is a prevalent strategy for coupling environmental stimuli to adaptive responses in bacteria. In both histidine kinases and response regulators, modular domains with conserved structures and biochemical activities adopt different conformational states in the presence of stimuli or upon phosphorylation, enabling a diverse array of regulatory mechanisms based on inhibitory and/or activating protein-protein interactions imparted by different domain arrangements. This review summarizes some of the recent structural work that has provided insight to the functioning of bacterial histidine kinases and response regulators. Particular emphasis is placed on identifying features that are expected to be conserved among different two-component proteins from those that are expected to differ, with the goal of defining the extent to which knowledge of previously characterized two-component proteins can be applied to newly discovered systems. PMID:19575571

  20. Identification of structural interface characteristics using component mode synthesis

    NASA Technical Reports Server (NTRS)

    Huckelbridge, A. A.; Lawrence, C.

    1987-01-01

    The inability to adequately model connections has limited the ability to predict overall system dynamic response. Connections between structural components are often mechanically complex and difficult to accurately model analytically. Improved analytical models for connections are needed to improve system dynamic predictions. This study explores combining Component Mode synthesis methods for coupling structural components with Parameter Identification procedures for improving the analytical modeling of the connections. Improvements in the connection properties are computed in terms of physical parameters so the physical characteristics of the connections can be better understood, in addition to providing improved input for the system model. Two sample problems, one utilizing simulated data, the other using experimental data from a rotor dynamic test rig are presented.

  1. Identification of structural interface characteristics using component mode synthesis

    NASA Technical Reports Server (NTRS)

    Huckelbridge, A. A.; Lawrence, C.

    1987-01-01

    The inability to adequately model connections has limited the ability to predict overall system dynamic response. Connections between structural components are often mechanically complex and difficult to accurately model analytically. Improved analytical models for connections are needed to improve system dynamic predictions. This study explores combining Component Mode synthesis methods for coupling structural components with Parameter Identification procedures for improving the analytical modeling of the connections. Improvements in the connection properties are computed in terms of physical parameters so the physical characteristics of the connections can be better understood, in addition to providing improved input for the system model. Two sample problems, one utilizing simulated data, the other using experimental data from a rotor dynamic test rig, are presented.

  2. Identification of structural interface characteristics using component mode synthesis

    NASA Technical Reports Server (NTRS)

    Huckelbridge, A. A.; Lawrence, C.

    1989-01-01

    The inability to adequately model connections has limited the ability to predict overall system dynamic response. Connections between structural components are often mechanically complex and difficult to accurately model analytically. Improved analytical models for connections are needed to improve system dynamic predictions. This study explores combining Component Mode synthesis methods for coupling structural components with Parameter Identification procedures for improving the analytical modeling of the connections. Improvements in the connection properties are computed in terms of physical parameters so the physical characteristics of the connections can be better understood, in addition to providing improved input for the system model. Two sample problems, one utilizing simulated data, the other using experimental data from a rotor dynamic test rig, are presented.

  3. Experimental component mode synthesis of structures with sloppy joints

    NASA Technical Reports Server (NTRS)

    Blackwood, Gary H.; Von Flotow, A. H.

    1988-01-01

    The accuracy of component mode synthesis is investigated experimentally for substructures coupled by nonideal joints. The work is based upon a segmented experimental beam for which free-interface frequency response matrices are measured for each segment. These measurements are used directly in component mode synthesis to predict the behavior of the assembled structure; the segments are then physically joined, and the resulting frequency response of the superstructure is compared to the prediction. Rotational freeplay is then introduced into the connecting joint, and the new superstructure frequency response is compared to the original linear component mode synthesis prediction. The level of accuracy to be expected in component mode synthesis is discussed in terms of the degree of nonlinearity in the joints, mode number, and mode shapes.

  4. Development of optical components for in-vessel viewing systems used for fusion experimental reactor

    NASA Astrophysics Data System (ADS)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi; Tada, Eisuke; Morita, Yosuke; Seki, Masahiro

    1994-12-01

    Optical components including imagefiber, periscope, glass, reflecting mirror and adhesive for lens are essential elements of in-vessel viewing system use for fusion experimental reactor and extensive of gamma irradiation tests have been conducted. These components were irradiated in the range of 1 MGy - 100 MGy under the average exposure dose rate of 1 X 106 R/h. As a result, the observation limit of the imagefiber specially fabricated for radiation hard is obtained to be 12 MGy at a illuminance of 8500 lx. Deterioration of transmissivity of three kinds of glass (alkaline barium glass, lead glass and synthetic quartz glass) is small compared with standard glass for commercial periscope. A periscope which was made of these glasses is visible even after 20 MGy at 8500 lx and in case of the standard periscope, the observation limit is 1 kGy at 8500 lx. Decrease in the reflectance on chromium nitride coated reflecting mirror is extremely small than aluminum coated and platinum coated mirrors at accumulated dose of 100 MGy. Two types of adhesive made of polyester resin and epoxy resin became discolored and exfoliated after 50 MGy.

  5. Cryogenic system component development for the fusion experimental reactor at JAERI

    SciTech Connect

    Kato, T.; Kamiya, S.; Tada, E.; Hiyama, T.; Kawano, K.; Shimamoto, S.

    1986-01-01

    The major objective of fusion R and D at the Japan Atomic Energy Research Institute (JAERI) is to construct the Fusion Experimental Reactor (FER) to follow JT-60. The construction of FER inevitably requires development of a large, reliable, and efficient helium liquefier/refrigerator and the more advanced cryogenic technology for cooling superconducting toroidal and poloidal coils. Typical characteristics required for the cryogenic system of FER are 10 to 20 kW at 4 K as one unit, reliability for > 8000 h, a stable pulsed heat load, and high-energy efficiency of > 1/500. In this cryogenic system, the major components such as the helium compressor, turbo-expander, cold circulation pump for supercritical helium, and cold compressor to reduce operating temperature below 4 K should be scaled up to a mass flow rate of > 1000 g/s. For this purpose, JAERI has developed cryogenics since 1980 in accordance with the development program in which the scaling up of the major components mentioned above are involved as well as cooling technology development.

  6. Beyond lamins: other structural components of the nucleoskeleton

    PubMed Central

    Zhong, Zhixia; Wilson, Katherine L.; Dahl, Kris Noel

    2010-01-01

    The nucleus is bordered by a double bilayer nuclear envelope, communicates with the cytoplasm via embedded nuclear pore complexes, and is structurally supported by an underlying nucleoskeleton. The nucleoskeleton includes nuclear intermediate filaments formed by lamin proteins, which provide major structural and mechanical support to the nucleus. However other structural proteins also contribute to the function of nucleoskeleton and help connect it to the cytoskeleton. This chapter reviews nucleoskeletal components beyond lamins, and summarizes specific methods and strategies useful for analyzing nuclear structural proteins including actin, spectrin, titin, LINC complex proteins and nuclear spindle matrix proteins. These components can localize to highly specific functional subdomains at the nuclear envelope or nuclear interior, and can interact either stably or dynamically with a variety of partners. These components confer upon the nucleoskeleton a functional diversity and mechanical resilience that appears to rival the cytoskeleton. To facilitate the exploration of this understudied area of biology, we summarize methods useful for localizing, solubilizing and immunoprecipitating nuclear structural proteins, and a state-of-the-art method to measure a newly-recognized mechanical property of nucleus. PMID:20816232

  7. Beyond lamins other structural components of the nucleoskeleton.

    PubMed

    Zhong, Zhixia; Wilson, Katherine L; Dahl, Kris Noel

    2010-01-01

    The nucleus is bordered by a double bilayer nuclear envelope, communicates with the cytoplasm via embedded nuclear pore complexes, and is structurally supported by an underlying nucleoskeleton. The nucleoskeleton includes nuclear intermediate filaments formed by lamin proteins, which provide major structural and mechanical support to the nucleus. However, other structural proteins also contribute to the function of the nucleoskeleton and help connect it to the cytoskeleton. This chapter reviews nucleoskeletal components beyond lamins and summarizes specific methods and strategies useful for analyzing nuclear structural proteins including actin, spectrin, titin, linker of nucleoskeleton and cytoskeleton (LINC) complex proteins, and nuclear spindle matrix proteins. These components can localize to highly specific functional subdomains at the nuclear envelope or nuclear interior and can interact either stably or dynamically with a variety of partners. These components confer upon the nucleoskeleton a functional diversity and mechanical resilience that appears to rival the cytoskeleton. To facilitate the exploration of this understudied area of biology, we summarize methods useful for localizing, solubilizing, and immunoprecipitating nuclear structural proteins, and a state-of-the-art method to measure a newly-recognized mechanical property of nucleus. PMID:20816232

  8. Equilibrium Structures of Differentially Rotating Primary Components of Binary Stars

    NASA Astrophysics Data System (ADS)

    Mohan, C.; Lal, A. K.; Singh, V. P.

    1997-11-01

    In this paper a method is proposed for computing the equilibrium structures and various other observable physical parameters of the primary components of stars in binary systems assuming that the primary is more massive than the secondary and is rotating differentially about its axis. Kippenhahn and Thomas averaging approach (1970) is used in a manner earlier used by Mohan, Saxena and Agarwal (1990) to incorporate the rotational and tidal effects in the equations of stellar structure. Explicit expressions for the distortional terms appearing in the stellar structure equations have been obtained by assuming a general law of differential rotation of the typeω2 = b 0+b 1 s 2+b 2 s 4, where ω is the angular velocity of rotation of a fluid element in the star at a distance s from the axis of rotation, and b 0, b 1, b 2 are suitably chosen numerical constants. The expressions incorporate the effects of differential rotation and tidal distortions up to second order terms. The use of the proposed method has been illustrated by applying it to obtain the structures and observable parameters of certain differentially rotating primary components of the binary stars assuming the primary components to have polytropic structures.

  9. Neutronic analysis of alternative structural materials for fusion reactor blankets

    NASA Astrophysics Data System (ADS)

    Santos, Raul dos

    1988-07-01

    The neutronic performance of the International Tokamak Reactor (INTOR) blanket was studied when several alternative structural materials were used instead of the INTOR reference structural material, type 316 stainless steel. The alternative structural materials included: ferritic-, vanadium-, titanium-, long range ordered-, manganese austenitic-, and nimonic-alloys. All were treated both with and without a first-wall coating of beryllium or graphite. The tritium breeding ratio, the nuclear heating, and the gas (hydrogen and helium) production rates in the structural materials were calculated for the possible combinations of structural material and first-wall coating. These parameters were compared with those obtained by using SS-316. The nimonic alloy was the only one with worse neutronic performance than the SS-316.

  10. 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

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

    SciTech Connect

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

    2010-03-08

    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 {angstrom} 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.

  12. 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.

  13. 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

  14. 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.

  15. Experimental component mode systhesis of structures with nonlinear joints

    NASA Technical Reports Server (NTRS)

    Blackwood, Gary H.; Vonflotow, A. H.

    1988-01-01

    The accuracy of component mode synthesis is investigated experimentally for substructures coupled by non-ideal joints. The work is based upon a segmented experimental beam for which the free-interface frequency response matrices are measured for each segment. These measurements are used directly in component mode synthesis to predict the behavior of the assembled structure; the segments are then physically joined and the resulting frequency response of the superstructure is compared to the prediction. Rotational freeplay is then introduced into the connecting joint and the new superstructure frequency response is compared to the original linear CMS prediction. The level of accuracy to be expected in component mode synthesis is discussed in terms of the degree of nonlinearity in the joints, mode number and mode shapes.

  16. Damage rates for FFTF structural components and surveillance assemblies

    SciTech Connect

    Simons, R.L.

    1993-08-01

    The Fast Flux Test Facility (FFTF) surveillance program provides coupon surveillance materials that are irradiated to the expected lifetime damage dose that the represented component will experience. This methodology requires a knowledge of the damage dose rates to the surveillance assemblies and to the critical locations of the structural components. This analysis updates the predicted exposures from a total fluence to a displacement per atom (dpa) basis using Monte Carlo (computer code for) neutron photon (transport) code (MCNP). The MCNP calculation improves the relative consistency and lowers the predicted damage rates uncertainty in a number of out-of-core locations. The results were used an part of the evaluation to extend the lifetime of the invessel components to 30 years in support of multiple missions for FFTF.

  17. Calvert Cliffs Nuclear Power Plant Life Cycle Management/License Renewal Program: System, structure, and component screening. Final report

    SciTech Connect

    Doroshuk, B.W.; Tilden, B.M.; Hostetler, D.R.; Klein, D.J.; Negin, C.A.

    1994-09-01

    Central to the Life Cycle Management (LCM) Program for the Calvert Cliffs Nuclear Plant is its Integrated Plant Assessment (IPA) process; a comprehensive, systematic evaluation of the effectiveness of age-related degradation management for the plant`s important systems, structures, and components. The first step in this process is the screening of functionally important systems, structures that warrant further evaluation of aging issues. A detailed method and procedures for conducting this screening have been developed and thoroughly tested. The development and application of these procedures at Calvert Cliffs should permit other utilities to avoid implementation problems and avoid substantial front-end development costs. The IPA process is initiated by a screening step that identifies important systems, structures, and components for further evaluation. This report contains the screening methodology, provides procedures for System Level Screening and Component Level Screening, and summarizes results for five systems that represent a wide range of use. These are the Reactor Coolant System, Compressed Air System, Saltwater Cooling System, Electrical 4 Kv Transformers and Buses, and the Reactor Protective System. Examples of component screening are included for the Reactor Coolant System. These screening results show how to determine which equipment`s maintenance programs should be checked for degradation management effectiveness.

  18. 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.

  19. A structured overview of simultaneous component based data integration

    PubMed Central

    Van Deun, Katrijn; Smilde, Age K; van der Werf, Mariët J; Kiers, Henk AL; Van Mechelen, Iven

    2009-01-01

    Background Data integration is currently one of the main challenges in the biomedical sciences. Often different pieces of information are gathered on the same set of entities (e.g., tissues, culture samples, biomolecules) with the different pieces stemming, for example, from different measurement techniques. This implies that more and more data appear that consist of two or more data arrays that have a shared mode. An integrative analysis of such coupled data should be based on a simultaneous analysis of all data arrays. In this respect, the family of simultaneous component methods (e.g., SUM-PCA, unrestricted PCovR, MFA, STATIS, and SCA-P) is a natural choice. Yet, different simultaneous component methods may lead to quite different results. Results We offer a structured overview of simultaneous component methods that frames them in a principal components setting such that both the common core of the methods and the specific elements with regard to which they differ are highlighted. An overview of principles is given that may guide the data analyst in choosing an appropriate simultaneous component method. Several theoretical and practical issues are illustrated with an empirical example on metabolomics data for Escherichia coli as obtained with different analytical chemical measurement methods. Conclusion Of the aspects in which the simultaneous component methods differ, pre-processing and weighting are consequential. Especially, the type of weighting of the different matrices is essential for simultaneous component analysis. These types are shown to be linked to different specifications of the idea of a fair integration of the different coupled arrays. PMID:19671149

  20. [Structural components and peculiarities of Pseudomonas aeruginosa biofilm organization].

    PubMed

    Balko, O B; Avdieieva, L V

    2010-01-01

    Peculiarities of the structural organization of bacterial biofilm during its formation and disintegration have been investigated on the model of Pseudomonas aeruginosa UCM B-900 (ATCC 9027). It was shown, that development of the biofilm in a stationary system on glass was a two-vector process with changes in time and space. P. aeruginosa UCM B-900 biofilm is formed from single cells, passes through the stages of base components, net structure, islands and comes to the end with integration into a complete monolayer. The biofilm degradation repeats the stages of its formation in the reverse sequence. PMID:20812507

  1. 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.

  2. Aircraft fatigue and crack growth considering loads by structural component

    NASA Technical Reports Server (NTRS)

    Yost, J. D.

    1994-01-01

    The indisputable 1968 C-130 fatigue/crack growth data is reviewed to obtain additional useful information on fatigue and crack growth. The proven Load Environment Model concept derived empirically from F-105D multichannel recorder data is refined to a simpler method by going from 8 to 5 variables in the spectra without a decrease in accuracy. This approach provides the true fatigue/crack growth and load environment by structural component for both fatigue and strength design. Methods are presented for defining fatigue scatter and damage at crack initiation. These design tools and criteria may be used for both metal and composite aircraft structure.

  3. 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.

  4. Structural thermal tests on Advanced Neutron Source reactor fuel plates

    SciTech Connect

    Swinson, W.F.; Battiste, R.L.; Yahr, G.T.

    1995-08-01

    The thin aluminum-clad fuel plates proposed for the Advanced Neutron Source reactor are stressed by the high-velocity coolant flowing on each side of the plates and by the thermal gradients in the plates. The total stress, composed of the sum of the flow stress and the thermal stress at a point, could be reduced if the thermal loads tend to relax when the stress magnitude approaches the yield stress of the material. The potential of this occurring would be very significant in assessing the structural reliability of the fuel plates and has been investigated through experiment. The results of this investigation are given in this report.

  5. 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.

  6. 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.

  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. POD-Based Model Reduction toward Efficient Simulation of Flow in NuclearReactor Components

    NASA Astrophysics Data System (ADS)

    Ahmadpoor, Mohammad; Banyay, Greg; Mazumdar, Sagnik; Jana, Anirban; Kimber, Mark; Brigham, John

    2013-11-01

    The long-term objective of this research is reduced-order modeling (ROM) to simulate and understand the turbulent mixing inside the lower plenum of a Very High Temperature Reactor, while the present study focuses on confined isothermal jet flow. In general, two steps are required to generate a basis for a ROM: (1) acquisition of an ensemble of possible solution fields for the system; and (2) extracting key features of the ensemble to create the basis. Proper Orthogonal Decomposition (POD) is one approach for extracting features from an ensemble. For this work POD is used to capture the parametric variation of a flow with Reynolds (Re) number and time. Two approaches are considered for model reduction: (1) a regression-based approach, which does not keep the mathematical structure of the modeling, but rather uses interpolation and/or extrapolation to predict flow fields at different Re number or different times and (2) a Galerkin-projection approach in which the Navier-Stokes equations are projected onto the POD modes to obtain low-dimensional ordinary differential equations to represent the fluid flow under conditions outside of the original ensemble.

  10. 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.

  11. Use of principal components analysis and three-dimensional atmospheric-transport models for reactor-consequence evaluation

    SciTech Connect

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

    1982-01-01

    This work explores the use of principal components analysis coupled to three-dimensional atmospheric transport and dispersion models for evaluating the environmental consequences of reactor accidents. This permits the inclusion of meteorological data from multiple sites and the effects of topography in the consequence evaluation; features not normally included in such analyses. The technique identifies prevailing regional wind patterns and their frequencies for use in the transport and dispersion calculations. Analysis of a hypothetical accident scenario involving a release of radioactivity from a reactor situated in a river valley indicated the technique is quite useful whenever recurring wind patterns exist, as is often the case in complex terrain situations. Considerable differences were revealed in a comparison with results obtained from a more conventional Gaussian plume model using only the reactor site meteorology and no topographic effects.

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

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

    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.

  14. 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.

  15. Detection of Component Failures for Smart Structure Control Systems

    NASA Astrophysics Data System (ADS)

    Okubo, Hiroshi

    Uncertainties in the dynamics model of a smart structure are often of significance due to model errors caused by parameter identification errors and reduced-order modeling of the system. Design of a model-based Failure Detection and Isolation (FDI) system for smart structures, therefore, needs careful consideration regarding robustness with respect to such model uncertainties. In this paper, we proposes a new method of robust fault detection that is insensitive to the disturbances caused by unknown modeling errors while it is highly sensitive to the component failures. The capability of the robust detection algorithm is examined for the sensor failure of a flexible smart beam control system. It is shown by numerical simulations that the proposed method suppresses the disturbances due to model errors and markedly improves the detection performance.

  16. 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. PMID:27053524

  17. 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.

  18. VIPRE (Versatile Internals and Component Program for Reactors; EPRI)-01: A thermal-hydraulic code for reactor cores: Volume 4, Applications: Final report

    SciTech Connect

    Cuta, J.M.; Stewart, C.W.; Koontz, A.S.; Montgomery, S.D.

    1987-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 4: Applications) contains extensive comparisons of VIPRE calculations to experimental data. There are also sensitivity studies and evaluations of code numerical and computational performance. In addition, calculations performed by member utilities using VIPRE for comparisons with transient CHF data, and FSAR plant analyses are presented. Comparisons are also presented of plant thermal-hydraulic calculations with VIPRE and other COBRA codes. These calculations demonstrate the suitability of VIPRE for PWR core thermal-hydraulic analysis.

  19. 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.

  20. Catalytic reactor

    DOEpatents

    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.

  1. 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

  2. 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.

  3. 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.

  4. Induced radioactivity of LDEF materials and structural components.

    PubMed

    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

    1996-11-01

    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. PMID:11540519

  5. 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.

  6. 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.

  7. Fabrication and nondestructive examination development for advanced components and materials for the SP-100 space reactor

    NASA Astrophysics Data System (ADS)

    Ring, Peter J.; Dobrzynski, Walter J.

    1993-01-01

    Significant progress has now been made in the development of fabrication and Nondestructive Examination techniques for the SP-100 Space Reactor. All major fabrication challenges have been faced and overcome. Methods are in place for the fabrication and inspection of composite fuel cladding, the reactor honeycomb core, cold forging of the core support nozzle course, and electron beam welding of the auxiliary cooling loop system. Specifications and procedures have been developed and proven on actual hardware for electron beam welding, gas tungsten arc welding, heat treatment, solvent cleaning, chemical cleaning, ultrasonic inspection, helium leak testing, dye penetrant and microfocus rod anode radiography. Signicant work remains to be done but no problems have been identified which would prevent fabrication of the high temperature SP-100 Space Reactor.

  8. 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.

  9. 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.

  10. 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.

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

    DOE PAGESBeta

    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 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

  12. Recognizing genes and other components of genomic structure

    SciTech Connect

    Burks, C. ); Myers, E. . Dept. of Computer Science); Stormo, G.D. . Dept. of Molecular, Cellular and Developmental Biology)

    1991-01-01

    The Aspen Center for Physics (ACP) sponsored a three-week workshop, with 26 scientists participating, from 28 May to 15 June, 1990. The workshop, entitled Recognizing Genes and Other Components of Genomic Structure, focussed on discussion of current needs and future strategies for developing the ability to identify and predict the presence of complex functional units on sequenced, but otherwise uncharacterized, genomic DNA. We addressed the need for computationally-based, automatic tools for synthesizing available data about individual consensus sequences and local compositional patterns into the composite objects (e.g., genes) that are -- as composite entities -- the true object of interest when scanning DNA sequences. The workshop was structured to promote sustained informal contact and exchange of expertise between molecular biologists, computer scientists, and mathematicians. No participant stayed for less than one week, and most attended for two or three weeks. Computers, software, and databases were available for use as electronic blackboards'' and as the basis for collaborative exploration of ideas being discussed and developed at the workshop. 23 refs., 2 tabs.

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

    PubMed

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

    2015-10-28

    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 ≤ 10,000 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. 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. PMID:26520533

  14. 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.

  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. 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.

  17. Cyanobacterial Two-Component Proteins: Structure, Diversity, Distribution, and Evolution†

    PubMed Central

    Ashby, Mark K.; Houmard, Jean

    2006-01-01

    A survey of the already characterized and potential two-component protein sequences that exist in the nine complete and seven partially annotated cyanobacterial genome sequences available (as of May 2005) showed that the cyanobacteria possess a much larger repertoire of such proteins than most other bacteria. By analysis of the domain structure of the 1,171 potential histidine kinases, response regulators, and hybrid kinases, many various arrangements of about thirty different modules could be distinguished. The number of two-component proteins is related in part to genome size but also to the variety of physiological properties and ecophysiologies of the different strains. Groups of orthologues were defined, only a few of which have representatives with known physiological functions. Based on comparisons with the proposed phylogenetic relationships between the strains, the orthology groups show that (i) a few genes, some of them clustered on the genome, have been conserved by all species, suggesting their very ancient origin and an essential role for the corresponding proteins, and (ii) duplications, fusions, gene losses, insertions, and deletions, as well as domain shuffling, occurred during evolution, leading to the extant repertoire. These mechanisms are put in perspective with the different genetic properties that cyanobacteria have to achieve genome plasticity. This review is designed to serve as a basis for orienting further research aimed at defining the most ancient regulatory mechanisms and understanding how evolution worked to select and keep the most appropriate systems for cyanobacteria to develop in the quite different environments that they have successfully colonized. PMID:16760311

  18. Design Study of Small Lead-Cooled Fast Reactors Using SiC Cladding and Structure

    SciTech Connect

    Abu Khalid Rivai; Minoru Takahashi

    2006-07-01

    Effects of SiC cladding and structure on neutronics of reactor core for small lead-cooled fast reactors have been investigated analytically. The fuel of this reactor was uranium nitride with {sup 235}U enrichment of 11% in inner core and 13% in outer core. The reactors were designed by optimizing the use of natural uranium blanket and nitride fuel to prolong the fuel cycle. The fuels can be used without re-shuffling for 15 years. The coolant of this reactor was lead. A calculation was also conducted for steel cladding and structure type as comparison with SiC cladding and structure type. The results of calculation indicated that the neutron energy spectrum of the core using SiC was slightly softer than that using steel. The SiC type reactor was designed to have criticality at the beginning of cycle (BOC), although the steel type reactor could not have critical condition with the same size and geometry. In other words, the SiC type core can be designed smaller than the steel type core. The result of the design analysis showed that neutron flux distributions and power distribution was made flatter because the outer core enrichment was higher than inner core. The peak power densities could remain constant over the reactor operation. The consumption capability of uranium was quite high, i.e. 13% for 125 MWt reactor and 25% for 375 MWt reactor at EOC. (authors)

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. Intact and Degraded Component Criticality Calculations of N Reactors Spent Nuclear Fuel

    SciTech Connect

    L. Angers

    2001-01-31

    The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR.

  4. The nuclear data, A key component for reactor studies, Overview of AREVA NP needs and applications

    NASA Astrophysics Data System (ADS)

    Ravaux, Simon; Demy, Pierre-Marie; Rechatin, Clément

    2016-03-01

    The quality of the nuclear data is essential for AREVA NP. Indeed, many AREVA NP activities such as reactor design, safety studies or reactor instrumentation use them as input data. So, the nuclear data can be considered as a key element for AREVA NP. REVA NP's contribution in the improvement of the nuclear data consists in a joint effort with the CEA. It means a financing and a sharing of information which can give an orientation to the future research axis. The aim of this article is to present the industrial point of view from AREVA NP on the research on nuclear data. Several examples of collaborations with the CEA which have resulted in an improvement of the nuclear data are presented.

  5. 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)

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Experimental study of thermal crisis in connection with Tokamak reactor high heat flux components

    NASA Astrophysics Data System (ADS)

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

    2000-04-01

    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. 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

    ... whose expertise includes, at a minimum, PRA, safety analysis, plant operation, design engineering, and..., 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...

  12. 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

    ... whose expertise includes, at a minimum, PRA, safety analysis, plant operation, design engineering, and..., 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...

  13. Spectral structure of electron antineutrinos from nuclear reactors.

    PubMed

    Dwyer, D A; Langford, T J

    2015-01-01

    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. PMID:25615462

  14. Environmentally friendly efficient coupling of n-heptane by sulfated tri-component metal oxides in slurry bubble column reactor.

    PubMed

    Ma, Hongzhu; Xiao, Jing; Wang, Bo

    2009-07-30

    SO(4)(2-)/M(x)O(y) is of the greatest interest in solid catalysts and green catalysts. Slurry bubble column reactors are of considerable interest in industrial processes and various biochemical processes. The cetane number (CN) has widely used diesel fuel quality parameter related to the ignition delay time (and combustion quality) of a fuel. CN improvement of diesel fuels is a difficult task that refiners will face in the near future. For that purpose, the tests were designed in which n-heptane is used as the reactant in the air or ozone atmosphere at room temperature (RT) and local atmospheric pressure (LAP) using different catalysts of sulfated tri-component metal oxides SO(4)(2-)/Fe(2)O(3)-TiO2-SnO(2) (SFTSn) and SO(4)(2-)/MnO(2)-TiO2-SnO(2) (SMTSn) in slurry bubble column reactor. The products distribution was analyzed by gas chromatography-mass spectrometry (GC-MS) method and the results show that the relative selectivity of long linear alkane (C(12)-C(28)) reaches the maximum (87.330%) when SMTSn is used as catalyst in flow air at 60 min. Diesel fuel components with higher cetane numbers can be easily obtained from this study. PMID:19124196

  15. 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.

  16. 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)

  17. Lumped-parameter models for transient conduction in nuclear reactor components

    SciTech Connect

    Wulff, W.

    1980-06-01

    Lumped-parameter models are presented for the prediction of transient conduction in cylinders, plane-parallel slabs, tubes and fuel elements under conditions of small-break loss of coolant accidents and planned transients in nuclear reactor systems. The model accounts for heat generation by fission and decay and by gamma absorption and water reaction. The models consist of ordinary differential equations, one each for the solid cylinder, the slab and the tube, and two for fuel pellet and clad in the fuel element. The differential equations are derived from the partial differential equation of energy conservation by volume averaging. The models can be applied to each axial node of a flow passage. The model formulation is particularly suitable for explicit integration over time with high-order integration schemes for ordinary, first-order differential equations.

  18. 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.

  19. 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.

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

    SciTech Connect

    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 alignment 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.

  1. 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.

  2. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  3. 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

  4. 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.

  5. Structure and compositional studies of multi-component nanoparticles

    NASA Astrophysics Data System (ADS)

    Malyavanatham, Gokul

    The laser ablation of microparticle (LAM) process was used to study nanoparticles of multi-component materials. The production process utilized laser ablation of a continuously flowing aerosol of micron-sized particles under a gas ambient. An aerosol generator entrained microparticles into a gas flow and directed them through a nozzle into a laser interaction cell. After plasma breakdown, the shock wave propagated through the microparticles and the nanoparticles condensed behind this shockwave. Two methods were developed to collect nanoparticles; the first method used supersonic impaction on substrates at room temperature to enable direct writing of thick films and the second method used electric fields to deflect and collect charged, individual nanoparticles. Two methods for generating multi-component nanostructured materials were studied. The first method involved feeding single-phase microparticles containing the desired composition. Lead Zirconate Titanate (PZT) microparticles were used to generate nanoparticles that were then impacted onto substrates to produce thick films. Quality, morphology, crystallization and composition variations of these thick films were analyzed using material characterization techniques. Segregation of elements and an overall deficiency in Zr and Ti were observed in the deposited thick films as a result of the agglomerated state of the PZT microparticles. However, the analysis for this material system was complicated by the presence of multiple compounds. To develop a further understanding of how segregation occurs in multi-component systems during the LAM process, a second method for generating multi-component nanoparticles by feeding mixtures of single component microparticles was studied. Nanoparticles generated by ablation of Cu and Au microparticle mixtures were collected electrostatically and analyzed. Interactions between exploding microparticles resulted in condensation of nanoparticles that were non-equilibrium solid

  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. 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

  8. Flaw assessment guide for high-temperature reactor components subject to creep-fatigue loading

    SciTech Connect

    Ainsworth, R.A. . Berkeley Nuclear Labs.); Ruggles, M.B. ); Takahashi, Y. . Komae Research Lab.)

    1990-10-01

    A high-temperature flaw assessment procedure is described. This procedure is a result of a collaborative effort between Electric Power Research Institute in the United States, Central Research Institute of Electric Power Industry in Japan, and Nuclear Electric plc in the United Kingdom. The procedure addresses preexisting defects subject to creep-fatigue loading conditions. Laws employed to calculate the crack growth per cycle are defined in terms of fracture mechanics parameters and constants related to the component material. The crack-growth laws can be integrated to calculate the remaining life of a component or to predict the amount of crack extension in a given period. Fatigue and creep crack growth per cycle are calculated separately, and the total crack extension is taken as the simple sum of the two contributions. An interaction between the two propagation modes is accounted for in the material properties in the separate calculations. In producing the procedure, limitations of the approach have been identified. 25 refs., 1 fig.

  9. 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.

  10. Proportional-hazards models for improving the analysis of light-water-reactor-component failure data

    SciTech Connect

    Booker, J.B.; Johnson, M.E.; Easterling, R.G.

    1981-01-01

    The reliability of a power plant component may depend on a variety of factors (or covariates). If a single regression model can be specified to relate these factors to the failure rate, then all available data can be used to estimate and test for the effects of these covariates. One such model is a proportional hazards function that is specified as a product of two terms: a nominal hazard rate that is a function of time and a second term that is a function of the covariates. The purpose of this paper is to adapt two such models to LWR valve failure rate analysis, to compare the results, and to discuss the strengths and weaknesses of these applications.