Preliminary Analysis of SiC BWR Channel Box Performance under Normal Operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wirth, Brian; Singh, Gyanender P.; Gorton, Jacob

SiC-SiC composites are being considered for applications in the core components, including BWR channel box and fuel rod cladding, of light water reactors to improve accident tolerance. In the extreme nuclear reactor environment, core components like the BWR channel box will be exposed to neutron damage and a corrosive environment. To ensure reliable and safe operation of a SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperaturesmore » on the deformation behavior of the channel box over the course of one cycle of irradiation. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions have been calculated using the neutronics and thermal-hydraulics codes Serpent2 and COBRA-TF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5mm. The channel box bowing behavior is time dependent, and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.« less

Posttest data analysis and assessment of TRAC-BD1/MOD1 with data from a Full Integral Simulation Test (FIST) power transient experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheatley, P.D.; Wagner, K.C.

The FIST power transient test 6PMC2 was analyzed to further the understanding of the FIST facility and provide an assessment of TRAC-BD1/MOD1. FIST power transient 6PMC2 investigated the thermal-hydraulic response following inadvertent closure of the main steam isolation valve and the subsequent failure of the reactor to scram. Failure of the high pressure core spray system was also assumed, resulting on only the reactor core isolation cooling flow for inventory makeup during the transient. The experiment was a sensitivity study with relatively high core power and low makeup rates. This study provides one of the first opportunities to assess TRAC-BD1/MOD1more » under power transient and natural circulation conditions with data from a facility with prototypical BWR geometry. The power transient test was analyzed with emphasis on the following phenomena; (a) the system pressure response, (b) the natural circulation flows and rates, and (c) the heater rod cladding temperature response. Based on the results of this study, TRAC-BD1/MOD1 can be expected to calculate the thermal-hydraulic behavior of a BWR during a power transient.« less

Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

DOEpatents

Tokarz, R.D.

1981-10-27

This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

An efficient modeling method for thermal stratification simulation in a BWR suppression pool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haihua Zhao; Ling Zou; Hongbin Zhang

2012-09-01

The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safetymore » analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.« less

Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian J; Bowman, Stephen M; Gauld, Ian C

2015-01-01

[Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, andmore » it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades are inserted in various locations and at varying degrees during BWR operation based on the reload design. The presence of control blades during depletion hardens the neutron spectrum locally due to both moderator displacement and introduction of a thermal neutron absorber. The reactivity impact of control blade presence is investigated herein, as well as the effect of multiple (continuous and intermittent) exposure periods. The coupled effects of control blade presence on power density, void profile, or burnup profile have not been considered to date but will be addressed in future work.« less

Systematic void fraction studies with RELAP5, FRANCESCA and HECHAN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stosic, Z.; Preusche, G.

1996-08-01

In enhancing the scope of standard thermal-hydraulic codes applications beyond its capabilities, i.e. coupling with a one and/or three-dimensional kinetics core model, the void fraction, transferred from thermal-hydraulics to the core model, plays a determining role in normal operating range and high core flow, as the generated heat and axial power profiles are direct functions of void distribution in the core. Hence, it is very important to know if the void quality models in the programs which have to be coupled are compatible to allow the interactive exchange of data which are based on these constitutive void-quality relations. The presentedmore » void fraction study is performed in order to give the basis for the conclusion whether a transient core simulation using the RELAP5 void fractions can calculate the axial power shapes adequately. Because of that, the void fractions calculated with RELAP5 are compared with those calculated by BWR safety code for licensing--FRANCESCA and the best estimate model for pre- and post-dryout calculation in BWR heated channel--HECHAN. In addition, a comparison with standard experimental void-quality benchmark tube data is performed for the HECHAN code.« less

Post-test analysis of PIPER-ONE PO-IC-2 experiment by RELAP5/MOD3 codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bovalini, R.; D`Auria, F.; Galassi, G.M.

1996-11-01

RELAP5/MOD3.1 was applied to the PO-IC-2 experiment performed in PIPER-ONE facility, which has been modified to reproduce typical isolation condenser thermal-hydraulic conditions. RELAP5 is a well known code widely used at the University of Pisa during the past seven years. RELAP5/MOD3.1 was the latest version of the code made available by the Idaho National Engineering Laboratory at the time of the reported study. PIPER-ONE is an experimental facility simulating a General Electric BWR-6 with volume and height scaling ratios of 1/2,200 and 1./1, respectively. In the frame of the present activity a once-through heat exchanger immersed in a pool ofmore » ambient temperature water, installed approximately 10 m above the core, was utilized to reproduce qualitatively the phenomenologies expected for the Isolation Condenser in the simplified BWR (SBWR). The PO-IC-2 experiment is the flood up of the PO-SD-8 and has been designed to solve some of the problems encountered in the analysis of the PO-SD-8 experiment. A very wide analysis is presented hereafter including the use of different code versions.« less

Use of eddy current mixes to solve a weld examination application

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ward, R.C.; LaBoissonniere, A.

1995-12-31

The augmentation of typical nondestructive (i.e., ultrasound) weld inspection techniques by the use of eddy current tools may significantly enhance the quality and reliability of weld inspections. One recent example is the development of an eddy current technique for use in the examination of BWR core shroud welds, where multi-frequency mixes are used to eliminate signals coming from the weld material so that the examination of the heat affected zone is enhanced. An analysis tool most commonly associated with ultrasound examinations, the C-Scan based on gated information, may be implemented with eddy current data to enhance analysis.

Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, William BJ J; Ade, Brian J; Bowman, Stephen M

2015-01-01

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k eff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of latticemore » design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented.« less

Core design of a direct-cycle, supercritical-water-cooled fast breeder reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jevremovic, T.; Oka, Yoshiaki; Koshizuka, Seiichi

1994-10-01

The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without muchmore » deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.« less

BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

DOE Office of Scientific and Technical Information (OSTI.GOV)

T.L. Lotz

1997-02-15

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercialmore » spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.« less

Experimental and analytical study of stability characteristics of natural circulation boiling water reactors during startup transient

NASA Astrophysics Data System (ADS)

Woo, Kyoungsuk

Two-phase natural circulation loops are unstable at low pressure operating conditions. New reactor design relying on natural circulation for both normal and abnormal core cooling is susceptible to different types of flow instabilities. In contrast to forced circulation boiling water reactor (BWR), natural circulation BWR is started up without recirculation pumps. The tall chimney placed on the top of the core makes the system susceptible to flashing during low pressure start-up. In addition, the considerable saturation temperature variation may induce complicated dynamic behavior driven by thermal non-equilibrium between the liquid and steam. The thermal-hydraulic problems in two-phase natural circulation systems at low pressure and low power conditions are investigated through experimental methods. Fuel heat conduction, neutron kinetics, flow kinematics, energetics and dynamics that govern the flow behavior at low pressure, are formulated. A dimensionless analysis is introduced to obtain governing dimensionless groups which are groundwork of the system scaling. Based on the robust scaling method and start-up procedures of a typical natural circulation BWR, the simulation strategies for the transient with and without void reactivity feedback is developed. Three different heat-up rates are applied to the transient simulations to study characteristics of the stability during the start-up. Reducing heat-up rate leads to increase in the period of flashing-induced density wave oscillation and decrease in the system pressurization rate. However, reducing the heat-up rate is unable to completely prevent flashing-induced oscillations. Five characteristic regions of stability are discovered at low pressure conditions. They are stable single-phase, flashing near the separator, intermittent oscillation, sinusoidal oscillation and low subcooling stable regions. Stability maps were acquired for system pressures ranging 100 kPa to 400 kPa. According to experimental investigation, the flow becomes stable below a certain heat flux regardless of the inlet subcooling at the core and system pressure. At higher heat flux, unstable phenomena were indentified within a certain range of inlet subcooling. The unstable region diminishes as the system pressure increases. In natural circulation BWRs, the significant gravitational pressure drop over the tall chimney section induces a Type-I instability. The Type-I instability becomes especially important during low power and pressure conditions during reactor start-up. Under these circumstances the effect of pressure variations on the saturation enthalpy becomes significant. An experimental study shows that the flashing phenomenon in the adiabatic chimney section is dominant during the start-up of a natural circulation BWR. Since flashing occurs outside the core, nuclear feedback effects on the stability are small. Furthermore, the thermal-hydraulic oscillation period is much longer than power fluctuation period caused by void reactivity feedback. In the natural circulation system increasing the inlet restriction reduces the natural circulation flow rate, shifting the unstable region to higher inlet subcooling.

Radiation chemistry related to nuclear power technology

NASA Astrophysics Data System (ADS)

Ishigure, Kenkichi

A brief review is given to the radiation chemical problems, especially with the emphasis on water radiolysis, in the nuclear power technology. Radiation chemistry in aqueous system is pointed out to be closely related to the problems such as corrosion of Zircaloy, the formation of insoluble corrosion products or crud, stress corrosion cracking of stainless steel in BWR and the radioactive waste managements. The results of the constant extention rate tests on sensitized 304 stainless steel under irradiation are shown, and the computer calculations were carried out to simulate the model experiments on the release of crud from the corroding surface under irradiation and also the water radiolysis in core of BWR.

Numerical Simulation of the Emergency Condenser of the SWR-1000

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krepper, Eckhard; Schaffrath, Andreas; Aszodi, Attila

The SWR-1000 is a new innovative boiling water reactor (BWR) concept, which was developed by Siemens AG. This concept is characterized in particular by passive safety systems (e.g., four emergency condensers, four building condensers, eight passive pressure pulse transmitters, and six gravity-driven core-flooding lines). In the framework of the BWR Physics and Thermohydraulic Complementary Action to the European Union BWR Research and Development Cluster, emergency condenser tests were performed by Forschungszentrum Juelich at the NOKO test facility. Posttest calculations with ATHLET are presented, which aim at the determination of the removable power of the emergency condenser and its operation mode.more » The one-dimensional thermal-hydraulic code ATHLET was extended by the module KONWAR for the calculation of the heat transfer coefficient during condensation in horizontal tubes. In addition, results of conventional finite difference calculations using the code CFX-4 are presented, which investigate the natural convection during the heatup process at the secondary side of the NOKO test facility.« less

Optimization of a Boiling Water Reactor Loading Pattern Using an Improved Genetic Algorithm

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobayashi, Yoko; Aiyoshi, Eitaro

2003-08-15

A search method based on genetic algorithms (GA) using deterministic operators has been developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). The search method uses an Improved GA operator, that is, crossover, mutation, and selection. The handling of the encoding technique and constraint conditions is designed so that the GA reflects the peculiar characteristics of the BWR. In addition, some strategies such as elitism and self-reproduction are effectively used to improve the search speed. LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and three-dimensional-dependent constraints have alwaysmore » necessitated the use of three-dimensional core simulators for BWRs, so that an optimization method is required for computational efficiency. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant applying the Haling technique. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.« less

Development and Assessment of CTF for Pin-resolved BWR Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Wysocki, Aaron J; Collins, Benjamin S

2017-01-01

CTF is the modernized and improved version of the subchannel code, COBRA-TF. It has been adopted by the Consortium for Advanced Simulation for Light Water Reactors (CASL) for subchannel analysis applications and thermal hydraulic feedback calculations in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). CTF is now jointly developed by Oak Ridge National Laboratory and North Carolina State University. Until now, CTF has been used for pressurized water reactor modeling and simulation in CASL, but in the future it will be extended to boiling water reactor designs. This required development activities to integrate the code into the VERA-CSmore » workflow and to make it more ecient for full-core, pin resolved simulations. Additionally, there is a significant emphasis on producing high quality tools that follow a regimented software quality assurance plan in CASL. Part of this plan involves performing validation and verification assessments on the code that are easily repeatable and tied to specific code versions. This work has resulted in the CTF validation and verification matrix being expanded to include several two-phase flow experiments, including the General Electric 3 3 facility and the BWR Full-Size Fine Mesh Bundle Tests (BFBT). Comparisons with both experimental databases is reasonable, but the BFBT analysis reveals a tendency of CTF to overpredict void, especially in the slug flow regime. The execution of these tests is fully automated, analysis is documented in the CTF Validation and Verification manual, and the tests have become part of CASL continuous regression testing system. This paper will summarize these recent developments and some of the two-phase assessments that have been performed on CTF.« less

Approach to numerical safety guidelines based on a core melt criterion. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Azarm, M.A.; Hall, R.E.

1982-01-01

A plausible approach is proposed for translating a single level criterion to a set of numerical guidelines. The criterion for core melt probability is used to set numerical guidelines for various core melt sequences, systems and component unavailabilities. These guidelines can be used as a means for making decisions regarding the necessity for replacing a component or improving part of a safety system. This approach is applied to estimate a set of numerical guidelines for various sequences of core melts that are analyzed in Reactor Safety Study for the Peach Bottom Nuclear Power Plant.

Experimental Study of Two Phase Flow Behavior Past BWR Spacer Grids

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ratnayake, Ruwan K.; Hochreiter, L.E.; Ivanov, K.N.

2002-07-01

Performance of best estimate codes used in the nuclear industry can be significantly improved by reducing the empiricism embedded in their constitutive models. Spacer grids have been found to have an important impact on the maximum allowable Critical Heat Flux within the fuel assembly of a nuclear reactor core. Therefore, incorporation of suitable spacer grids models can improve the critical heat flux prediction capability of best estimate codes. Realistic modeling of entrainment behavior of spacer grids requires understanding the different mechanisms that are involved. Since visual information pertaining to the entrainment behavior of spacer grids cannot possibly be obtained frommore » operating nuclear reactors, experiments have to be designed and conducted for this specific purpose. Most of the spacer grid experiments available in literature have been designed in view of obtaining quantitative data for the purpose of developing or modifying empirical formulations for heat transfer, critical heat flux or pressure drop. Very few experiments have been designed to provide fundamental information which can be used to understand spacer grid effects and phenomena involved in two phase flow. Air-water experiments were conducted to obtain visual information on the two-phase flow behavior both upstream and downstream of Boiling Water Reactor (BWR) spacer grids. The test section was designed and constructed using prototypic dimensions such as the channel cross-section, rod diameter and other spacer grid configurations of a typical BWR fuel assembly. The test section models the flow behavior in two adjacent sub channels in the BWR core. A portion of a prototypic BWR spacer grid accounting for two adjacent channels was used with industrial mild steel rods for the purpose of representing the channel internals. Symmetry was preserved in this practice, so that the channel walls could effectively be considered as the channel boundaries. Thin films were established on the rod surfaces by injecting water through a set of perforations at the bottom ends of the rods, ensuring that the flow upstream of the bottom-most spacer grid is predominantly annular. The flow conditions were regulated such that they represent typical BWR operating conditions. Photographs taken during experiments show that the film entrainment increases significantly at the spacer grids, since the points of contact between the rods and the grids result in a peeling off of large portions of the liquid film from the rod surfaces. Decreasing the water flow resulted in eventual drying out, beginning at positions immediately upstream of the spacer grids. (authors)« less

Rapid depressurization event analysis in BWR/6 using RELAP5 and contain

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mueftueoglu, A.K.; Feltus, M.A.

1995-09-01

Noncondensable gases may become dissolved in Boiling Water Reactor (BWR) water level instrumentation during normal operations. Any dissolved noncondensable gases inside these water columns may come out of solution during rapid depressurization events, and displace water from the reference leg piping resulting in a false high level. These water level errors may cause a delay or failure in actuation, or premature shutdown of the Emergency Core Cooling System. (ECCS). If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response and othermore » signals for automatic actuation such as high drywell pressure. It is also important to determine the effect of the level signal on ECCS operation after it is being actuated. The objective of this study is to determine the detailed coupled containment/NSSS response during this rapid depressurization events in BWR/6. The selected scenarios involve: (a) inadvertent opening of all ADS valves, (b) design basis (DB) large break loss of coolant accident (LOCA), and (c) main steam line break (MSLB). The transient behaviors are evaluated in terms of: (a) vessel pressure and collapsed water level response, (b) specific transient boundary conditions, (e.g., scram, MSIV closure timing, feedwater flow, and break blowdown rates), (c) ECCS initiation timing, (d) impact of operator actions, (e) whether indications besides low-low water level were available. The results of the analysis had shown that there would be signals to actuate ECCS other than low reactor level, such as high drywell pressure, low vessel pressure, high suppression pool temperature, and that the plant operators would have significant indications to actuate ECCS.« less

Stability analysis of BWR nuclear-coupled thermal-hyraulics using a simple model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Karve, A.A.; Rizwan-uddin; Dorning, J.J.

1995-09-01

A simple mathematical model is developed to describe the dynamics of the nuclear-coupled thermal-hydraulics in a boiling water reactor (BWR) core. The model, which incorporates the essential features of neutron kinetics, and single-phase and two-phase thermal-hydraulics, leads to simple dynamical system comprised of a set of nonlinear ordinary differential equations (ODEs). The stability boundary is determined and plotted in the inlet-subcooling-number (enthalpy)/external-reactivity operating parameter plane. The eigenvalues of the Jacobian matrix of the dynamical system also are calculated at various steady-states (fixed points); the results are consistent with those of the direct stability analysis and indicate that a Hopf bifurcationmore » occurs as the stability boundary in the operating parameter plane is crossed. Numerical simulations of the time-dependent, nonlinear ODEs are carried out for selected points in the operating parameter plane to obtain the actual damped and growing oscillations in the neutron number density, the channel inlet flow velocity, and the other phase variables. These indicate that the Hopf bifurcation is subcritical, hence, density wave oscillations with growing amplitude could result from a finite perturbation of the system even where the steady-state is stable. The power-flow map, frequently used by reactor operators during start-up and shut-down operation of a BWR, is mapped to the inlet-subcooling-number/neutron-density (operating-parameter/phase-variable) plane, and then related to the stability boundaries for different fixed inlet velocities corresponding to selected points on the flow-control line. The stability boundaries for different fixed inlet subcooling numbers corresponding to those selected points, are plotted in the neutron-density/inlet-velocity phase variable plane and then the points on the flow-control line are related to their respective stability boundaries in this plane.« less

Application Of The Iberdrola Licensing Methodology To The Cofrentes BWR-6 110% Extended Power Up-rate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mata, Pedro; Fuente, Rafael de la; Iglesias, Javier

Iberdrola (spanish utility) and Iberdrola Ingenieria (engineering branch) have been developing during the last two years the 110% Extended Power Up-rate Project (EPU 110%) for Cofrentes BWR-6. IBERDROLA has available an in-house design and licensing reload methodology that has been approved by the Spanish Nuclear Regulatory Authority. This methodology has been already used to perform the nuclear design and the reload licensing analysis for Cofrentes cycles 12 to 14. The methodology has been also applied to develop a significant number of safety analysis of the Cofrentes Extended Power Up-rate including: Reactor Heat Balance, Core and Fuel performance, Thermal Hydraulic Stability,more » ECCS LOCA Evaluation, Transient Analysis, Anticipated Transient Without Scram (ATWS) and Station Blackout (SBO) Since the scope of the licensing process of the Cofrentes Extended Power Up-rate exceeds the range of analysis included in the Cofrentes generic reload licensing process, it has been required to extend the applicability of the Cofrentes licensing methodology to the analysis of new transients. This is the case of the TLFW transient. The content of this paper shows the benefits of having an in-house design and licensing methodology, and describes the process to extend the applicability of the methodology to the analysis of new transients. The case of analysis of Total Loss of Feedwater with the Cofrentes Retran Model is included as an example of this process. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trianti, Nuri, E-mail: nuri.trianti@gmail.com; Nurjanah,; Su’ud, Zaki

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 andmore » 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.« less

An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hirano, Masashi

1997-07-01

This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clayton, Dwight A.; Poore, III, Willis P.

2015-01-01

The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Markmore » I plant for those instrumentation systems considered most important for accident management purposes.« less

Method for depleting BWRs using optimal control rod patterns

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

1991-01-01

Control rod (CR) programming is an essential core management activity for boiling water reactors (BWRs). After establishing a core reload design for a BWR, CR programming is performed to develop a sequence of exposure-dependent CR patterns that assure the safe and effective depletion of the core through a reactor cycle. A time-variant target power distribution approach has been assumed in this study. The authors have developed OCTOPUS to implement a new two-step method for designing semioptimal CR programs for BWRs. The optimization procedure of OCTOPUS is based on the method of approximation programming and uses the SIMULATE-E code for nucleonicsmore » calculations.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Epiney, A.; Canepa, S.; Zerkak, O.

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R

2015-01-01

Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditionalmore » Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.« less

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.

This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARCmore » and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.« less

Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Was, Gary

This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

Computational Analysis of Splash Occurring in the Deposition Process in Annular-Mist Flow

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xie, Heng; Koshizuka, Seiichi; Oka, Yoshiaki

2004-07-01

The deposition process of a single droplet on the film is numerically simulated by the Moving Particle Semi-implicit (MPS) method to analyze the possibility and effect of splash occurring in the deposition process in BWR condition. The model accounts for the presence of inertial, gravitation, viscous and surface tension and is validated by comparison with experiment results. A simple one-dimensional mixture model is developed to calculate the necessary parameters for the simulation of deposition in BWR condition. The deposition process of a single droplet in BWR condition is simulated. The effect of impact angle of droplet and the velocity ofmore » liquid film are analyzed. A film buffer model is developed to fit the simulation results of critical value for splash. A correlation of critical Weber number for splash in BWR condition is obtained and used to analyze the effect of splash. It is found that the splash play important role in the deposition and re-entrainment process in high quality condition in BWR. The mass fraction of re-entrainment caused by splash in different quality condition is also calculated. (authors)« less

A study of the effect of space-dependent neutronics on stochastically-induced bifurcations in BWR dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Analytis, G.T.

1995-09-01

A non-linear one-group space-dependent neutronic model for a finite one-dimensional core is coupled with a simple BWR feed-back model. In agreement with results obtained by the authors who originally developed the point-kinetics version of this model, we shall show numerically that stochastic reactivity excitations may result in limit-cycles and eventually in a chaotic behaviour, depending on the magnitude of the feed-back coefficient K. In the framework of this simple space-dependent model, the effect of the non-linearities on the different spatial harmonics is studied and the importance of the space-dependent effects is exemplified and assessed in terms of the importance ofmore » the higher harmonics. It is shown that under certain conditions, when the limit-cycle-type develop, the neutron spectra may exhibit strong space-dependent effects.« less

Application of reliability-centered-maintenance to BWR ECCS motor operator valve performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Choi, Y.A.

1993-01-01

This paper describes the application of reliability-centered maintenance (RCM) methods to plant probabilistic risk assessment (PRA) and safety analyses for four boiling water reactor emergency core cooling systems (ECCSs): (1) high-pressure coolant injection (HPCI); (2) reactor core isolation cooling (RCIC); (3) residual heat removal (RHR); and (4) core spray systems. Reliability-centered maintenance is a system function-based technique for improving a preventive maintenance program that is applied on a component basis. Those components that truly affect plant function are identified, and maintenance tasks are focused on preventing their failures. The RCM evaluation establishes the relevant criteria that preserve system function somore » that an RCM-focused approach can be flexible and dynamic.« less

Optimization of Boiling Water Reactor Loading Pattern Using Two-Stage Genetic Algorithm

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobayashi, Yoko; Aiyoshi, Eitaro

2002-10-15

A new two-stage optimization method based on genetic algorithms (GAs) using an if-then heuristic rule was developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). In the first stage, the LP is optimized using an improved GA operator. In the second stage, an exposure-dependent control rod pattern (CRP) is sought using GA with an if-then heuristic rule. The procedure of the improved GA is based on deterministic operators that consist of crossover, mutation, and selection. The handling of the encoding technique and constraint conditions by that GA reflects the peculiar characteristics of the BWR. In addition, strategies suchmore » as elitism and self-reproduction are effectively used in order to improve the search speed. The LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and constraints dependent on three dimensions have always necessitated the use of three-dimensional core simulators for BWRs, so that optimization of computational efficiency is required. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant in two phases. One phase is only LP optimization applying the Haling technique. The other phase is an LP optimization that considers the CRP during reactor operation. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.« less

The Japanese utilities` expectations for subchannel analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toba, Akio; Omoto, Akira

1995-12-01

Boiling water reactor (BWR) utilities in Japan began to consider the development of a mechanistic model to describe the critical heat transfer conditions in the BWR fuel subchannel. Such a mechanistic model will not only decrease the necessity of tests, but will also help by removing some overly conservative safety margins in thermal hydraulics. With the use of a postdryout heat transfer correlation, new acceptance criteria may be applicable to evaluate the fuel integrity. Mechanistic subchannel analysis models will certainly back up this approach. This model will also be applicable to the analysis of large-size fuel bundles and examination ofmore » corrosion behavior.« less

Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dallman, R J; Gottula, R C; Holcomb, E E

1987-05-01

An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented.

Towards a Consolidated Approach for the Assessment of Evaluation Models of Nuclear Power Reactors

DOE PAGES

Epiney, A.; Canepa, S.; Zerkak, O.; ...

2016-11-02

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian J; Marshall, William BJ J; Martinez-Gonzalez, Jesus S

Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents themore » analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.« less

BWR Anticipated Transients Without Scram Leading to Instability

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng L. Y.; Baek J.; Cuadra, A.

2013-11-10

Anticipated transients without scram (ATWS) in aboiling water reactor (BWR) were simulated in order to understand reactor response and determine the effectiveness of automatic and operator actions to mitigate this beyond-design-basis accident. The events of interest herein are initiated by a turbine trip when the reactor is operating in the expanded operating domainMELLLA+ [maximum extended load line limit plus]. In these events the reactor may initially be at up to 120% of the original licensed thermal power (OLTP) and at flow rates as low as 80% of rated.For these (and similar) ATWS events the concern isthat when the reactor powermore » decreases in response to a dual recirculation pump trip, the core will become unstable and large amplitude oscillations will begin. The occurrence of these power oscillations, if left unmitigated, may result in fuel damage, and the amplitude of the poweroscillations may hamper the effectiveness of the injection of dissolved neutron absorber through the standby liquid control system (SLCS).« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Espinosa-Paredes, Gilberto; Prieto-Guerrero, Alfonso; Nunez-Carrera, Alejandro

This paper introduces a wavelet-based method to analyze instability events in a boiling water reactor (BWR) during transient phenomena. The methodology to analyze BWR signals includes the following: (a) the short-time Fourier transform (STFT) analysis, (b) decomposition using the continuous wavelet transform (CWT), and (c) application of multiresolution analysis (MRA) using discrete wavelet transform (DWT). STFT analysis permits the study, in time, of the spectral content of analyzed signals. The CWT provides information about ruptures, discontinuities, and fractal behavior. To detect these important features in the signal, a mother wavelet has to be chosen and applied at several scales tomore » obtain optimum results. MRA allows fast implementation of the DWT. Features like important frequencies, discontinuities, and transients can be detected with analysis at different levels of detail coefficients. The STFT was used to provide a comparison between a classic method and the wavelet-based method. The damping ratio, which is an important stability parameter, was calculated as a function of time. The transient behavior can be detected by analyzing the maximum contained in detail coefficients at different levels in the signal decomposition. This method allows analysis of both stationary signals and highly nonstationary signals in the timescale plane. This methodology has been tested with the benchmark power instability event of Laguna Verde nuclear power plant (NPP) Unit 1, which is a BWR-5 NPP.« less

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorzel, A.

2006-07-01

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a Germanmore » BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)« less

Investigation of two and three parameter equations of state for cryogenic fluids

NASA Technical Reports Server (NTRS)

Jenkins, Susan L.; Majumdar, Alok K.; Hendricks, Robert C.

1990-01-01

Two-phase flows are a common occurrence in cryogenic engines and an accurate evaluation of the heat-transfer coefficient in two-phase flow is of significant importance in their analysis and design. The thermodynamic equation of state plays a key role in calculating the heat transfer coefficient which is a function of thermodynamic and thermophysical properties. An investigation has been performed to study the performance of two- and three-parameter equations of state to calculate the compressibility factor of cryogenic fluids along the saturation loci. The two-parameter equations considered here are van der Waals and Redlich-Kwong equations of state. The three-parameter equation represented here is the generalized Benedict-Webb-Rubin (BWR) equation of Lee and Kesler. Results have been compared with the modified BWR equation of Bender and the extended BWR equations of Stewart. Seven cryogenic fluids have been tested; oxygen, hydrogen, helium, nitrogen, argon, neon, and air. The performance of the generalized BWR equation is poor for hydrogen and helium. The van der Waals equation is found to be inaccurate for air near the critical point. For helium, all three equations of state become inaccurate near the critical point.

New core-reflector boundary conditions for transient nodal reactor calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, E.K.; Kim, C.H.; Joo, H.K.

1995-09-01

New core-reflector boundary conditions designed for the exclusion of the reflector region in transient nodal reactor calculations are formulated. Spatially flat frequency approximations for the temporal neutron behavior and two types of transverse leakage approximations in the reflector region are introduced to solve the transverse-integrated time-dependent one-dimensional diffusion equation and then to obtain relationships between net current and flux at the core-reflector interfaces. To examine the effectiveness of new core-reflector boundary conditions in transient nodal reactor computations, nodal expansion method (NEM) computations with and without explicit representation of the reflector are performed for Laboratorium fuer Reaktorregelung und Anlagen (LRA) boilingmore » water reactor (BWR) and Nuclear Energy Agency Committee on Reactor Physics (NEACRP) pressurized water reactor (PWR) rod ejection kinetics benchmark problems. Good agreement between two NEM computations is demonstrated in all the important transient parameters of two benchmark problems. A significant amount of CPU time saving is also demonstrated with the boundary condition model with transverse leakage (BCMTL) approximations in the reflector region. In the three-dimensional LRA BWR, the BCMTL and the explicit reflector model computations differ by {approximately}4% in transient peak power density while the BCMTL results in >40% of CPU time saving by excluding both the axial and the radial reflector regions from explicit computational nodes. In the NEACRP PWR problem, which includes six different transient cases, the largest difference is 24.4% in the transient maximum power in the one-node-per-assembly B1 transient results. This difference in the transient maximum power of the B1 case is shown to reduce to 11.7% in the four-node-per-assembly computations. As for the computing time, BCMTL is shown to reduce the CPU time >20% in all six transient cases of the NEACRP PWR.« less

The effect of rider weight and additional weight in Icelandic horses in tölt: part II. Stride parameters responses.

PubMed

Gunnarsson, V; Stefánsdóttir, G J; Jansson, A; Roepstorff, L

2017-09-01

This study investigated the effects of rider weight in the BW ratio (BWR) range common for Icelandic horses (20% to 35%), on stride parameters in tölt in Icelandic horses. The kinematics of eight experienced Icelandic school horses were measured during an incremental exercise test using a high-speed camera (300 frames/s). Each horse performed five phases (642 m each) in tölt at a BWR between rider (including saddle) and horse starting at 20% (BWR20) and increasing to 25% (BWR25), 30% (BWR30), 35% (BWR35) and finally 20% (BWR20b) was repeated. One professional rider rode all horses and weight (lead) was added to saddle and rider as needed. For each phase, eight strides at speed of 5.5 m/s were analyzed for stride duration, stride frequency, stride length, duty factor (DF), lateral advanced placement, lateral advanced liftoff, unipedal support (UPS), bipedal support (BPS) and height of front leg action. Stride length became shorter (Y=2.73-0.004x; P0.05). In conclusion, increased BWR decreased stride length and increased DF proportionally to the same extent in all limbs, whereas BPS increased at the expense of decreased UPS. These changes can be expected to decrease tölt quality when subjectively evaluated according to the breeding goals for the Icelandic horse. However, beat, symmetry and height of front leg lifting were not affected by BWR.

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC)more » system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.« less

Best-estimate coupled RELAP/CONTAIN analysis of inadvertent BWR ADS valve opening transient

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Muftuoglu, A.K.

1993-01-01

Noncondensible gases may become dissolved in boiling water reactor (BWR) water-level instrumentation during normal operations. Any dissolved noncondensible gases inside these water columns may come out of solution during rapid depressurization events and displace water from the reference leg piping, resulting in a false high level. Significant errors in water-level indication are not expected to occur until the reactor pressure vessel (RPV) pressure has dropped below [approximately]450 psig. These water level errors may cause a delay or failure in emergency core cooling system (ECCS) actuation. The RPV water level is monitored using the pressure of a water column having amore » varying height (reactor water level) that is compared to the pressure of a water column maintained at a constant height (reference level). The reference legs have small-diameter pipes with varying lengths that provide a constant head of water and are located outside the drywell. The amount of noncondensible gases dissolved in each reference leg is very dependent on the amount of leakage from the reference leg and its geometry and interaction of the reactor coolant system with the containment, i.e., torus or suppression pool, and reactor building. If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response. In the postulated inadvertent opening of all seven automatic depressurization system (ADS) valves, the ECCS signal on high drywell pressure would be circumvented because the ADS valves discharge directly into the suppression pool. A best-estimate analysis of such an inadvertent opening of all ADS valves would have to consider the thermal-hydraulic coupling between the pool, drywell, reactor building, and RPV.« less

QUAD+ BWR Fuel Assembly demonstration program at Browns Ferry plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doshi, P.K.; Mayhue, L.T.; Robert, J.T.

1984-04-01

The QUAD+ fuel assembly is an improved BWR fuel assembly designed and manufactured by Westinghouse Electric Corporation. The design features a water cross separating four fuel minibundles in an integral channel. A demonstration program for this fuel design is planned for late 1984 in cycle 6 of Browns Ferry 2, a TVA plant. Objectives for the design of the QUAD+ demonstration assemblies are compatibility in performance and transparency in safety analysis with the feed fuel. These objectives are met. Inspections of the QUAD+ demonstration assemblies are planned at each refueling outage.

Trace Assessment for BWR ATWS Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, L.Y.; Diamond, D.; Arantxa Cuadra, Gilad Raitses, Arnold Aronson

2010-04-22

A TRACE/PARCS input model has been developed in order to be able to analyze anticipated transients without scram (ATWS) in a boiling water reactor. The model is based on one developed previously for the Browns Ferry reactor for doing loss-of-coolant accident analysis. This model was updated by adding the control systems needed for ATWS and a core model using PARCS. The control systems were based on models previously developed for the TRAC-B code. The PARCS model is based on information (e.g., exposure and moderator density (void) history distributions) obtained from General Electric Hitachi and cross sections for GE14 fuel obtainedmore » from an independent source. The model is able to calculate an ATWS, initiated by the closure of main steam isolation valves, with recirculation pump trip, water level control, injection of borated water from the standby liquid control system and actuation of the automatic depres-surization system. The model is not considered complete and recommendations are made on how it should be improved.« less

Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trianti, Nuri, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Su'ud, Zaki, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Arif, Idam, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id

2014-09-30

Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tightmore » concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.« less

TRACE/PARCS analysis of the OECD/NEA Oskarshamn-2 BWR stability benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kozlowski, T.; Downar, T.; Xu, Y.

2012-07-01

On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations with a decay ratio of about 1.4. The event was successfully modeled by the TRACE/PARCS coupled code system, and further analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations (decay ratio) and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validationmore » for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK. coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (authors)« less

Recent developments in BWR fuel design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Congdon, S.P.; Noble, L.D.; Wood, J.E.

1991-11-01

Substantial increases in the cost effectiveness and performance capability of boiling water reactor (BWR) fuel designs have been implemented in the past 5 to 7 yr. This increase has been driven by (a) utility desires to lower fuel and operating costs and (b) design innovations that have lowered enrichment requirements, improved thermal-hydraulic performance, and increased discharge exposure. Higher discharge exposures reduce disposal costs for European and Asian utilities and enable US utilities to lengthen operating cycles. A typical BWR reload fuel bundle fabricated today has 25% higher {sup 235}U enrichment and a factor of 2 higher gadolinium loading than onemore » made several years ago. Today's BWR fuel bundles also contain more unheated water reduces the axial water density variation, lowers the void coefficient, and enhances the neutron efficiency of the bundle, reducing both the gadolinium poison and the enrichment requirements. In addition to these general trends, the following unique design innovations have further enhanced the fuel cost efficiency and performance characteristics of BWR fuel: ferrule spacer, part length rods, interactive channel, and bundle enhanced spectral shift. GE's fuel designs offer the flexibility for modern BWR fuel requirements and contain unique design features that enhance flexibility for modern BWR fuel requirements and contain unique design features that enhance flexibility and fuel cycle economics.« less

Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arai, Kenji; Ebata, Shigeo

1997-07-01

This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding ofmore » the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.« less

The effect of rider weight and additional weight in Icelandic horses in tölt: part I. Physiological responses.

PubMed

Stefánsdóttir, G J; Gunnarsson, V; Roepstorff, L; Ragnarsson, S; Jansson, A

2017-09-01

This study examined the effect of increasing BW ratio (BWR) between rider and horse, in the BWR range common for Icelandic horses (20% to 35%), on heart rate (HR), plasma lactate concentration (Lac), BWR at Lac 4 mmol/l (W4), breathing frequency (BF), rectal temperature (RT) and hematocrit (Hct) in Icelandic horses. In total, eight experienced school-horses were used in an incremental exercise test performed outdoors on an oval riding track and one rider rode all horses. The exercise test consisted of five phases (each 642 m) in tölt, a four-beat symmetrical gait, at a speed of 5.4±0.1 m/s (mean±SD), where BWR between rider (including saddle) and horse started at 20% (BWR20), was increased to 25% (BWR25), 30% (BWR30), and 35% (BWR35) and finally decreased to 20% (BWR20b). Between phases, the horses were stopped (~5.5 min) to add lead weights to specially adjusted saddle bags and a vest on the rider. Heart rate was measured during warm-up, the exercise test and after 5, 15 and 30 min of recovery and blood samples were taken and BF recorded at rest, and at end of each of these aforementioned occasions. Rectal temperature was measured at rest, at end of the exercise test and after a 30-min recovery period. Body size and body condition score (BCS) were registered and a clinical examination performed on the day before the test and for 2 days after. Heart rate and BF increased linearly (P0.05), but negative correlations (P<0.05) existed between body size measurements and Hct. While HR, Hct and BF recovered to values at rest within 30 min, Lac and RT did not. All horses had no clinical remarks on palpation and at walk 1 and 2 days after the test. In conclusion, increasing BWR from 20% to 35% resulted in increased HR, Lac, RT and BF responses in the test group of experienced adult Icelandic riding horses. The horses mainly worked aerobically until BWR reached 22.7%, but considerable individual differences (17.0% to 27.5%) existed that were not linked to horse size, but to back BCS.

Analysis of dose rates received around the storage pool for irradiated control rods in a BWR nuclear power plant.

PubMed

Ródenas, J; Abarca, A; Gallardo, S

2011-08-01

BWR control rods are activated by neutron reactions in the reactor. The dose produced by this activity can affect workers in the area surrounding the storage pool, where activated rods are stored. Monte Carlo (MC) models for neutron activation and dose assessment around the storage pool have been developed and validated. In this work, the MC models are applied to verify the expected reduction of dose when the irradiated control rod is hanged in an inverted position into the pool. 2010 Elsevier Ltd. All rights reserved.

High Fidelity BWR Fuel Simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoon, Su Jong

This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fractionmore » and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.« less

Noble gas, iodine, and cesium transport in a postulated loss of decay heat removal accident at Browns Ferry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichner, R.P.; Hodge, S.A.; Weber, C.F.

1984-08-01

This report presents an analysis of the movement of noble gas, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto themore » drywell floor. The analysis of fission product transport presented in this report is based on the no-operator-action sequence and provides an estimate of fission product inventories, as a function of time, within 14 control volumes outside the core, with the atmosphere considered as the final control volume in the transport sequence. As in the case of accident sequences previously studied, we find small barrier for noble gas ejection to air, these gases being effectively purged from the drywell and reactor building by steam and concrete degradation gases. However, significant decay of krypton isotopes occurs during the long delay times involved in this sequence. In contrast, large degrees of holdup for iodine and cesium are projected due to the chemical reactivity of these elements. Only about 2 x 10/sup -4/% of the initial iodine and cesium activity are predicted to be released to the atmosphere. Principal barriers for release are deposition on reactor vessel and containment walls. A significant amount of iodine is captured in the water pool formed in the reactor building basement after actuation of the fire protection system.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

BWR Steam Dryer Alternating Stress Assessment Procedures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morante, R. J.; Hambric, S. A.; Ziada, S.

2016-12-01

This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

Logistics Modeling of Emplacement Rate and Duration of Operations for Generic Geologic Repository Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalinina, Elena Arkadievna; Hardin, Ernest

This study identified potential geologic repository concepts for disposal of spent nuclear fuel (SNF) and (2) evaluated the achievable repository waste emplacement rate and the time required to complete the disposal for these concepts. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel. The results of this study provide an important input for the rough-order-of-magnitude (ROM) disposal cost analysis. The disposal concepts cover three major categories of host geologic media: crystalline or hard rock, salt, and argillaceous rock. Four waste package sizes are considered: 4PWR/9BWR; 12PWR/21BWR; 21PWR/44BWR, and dual purpose canisters (DPCs). The DPC concepts assumemore » that the existing canisters will be sealed into disposal overpacks for direct disposal. Each concept assumes one of the following emplacement power limits for either emplacement or repository closure: 1.7 kW; 2.2 kW; 5.5 kW; 10 kW; 11.5 kW, and 18 kW.« less

Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

NASA Astrophysics Data System (ADS)

Artnak, Edward Joseph, III

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

Characterization of 14C in Swedish light water reactors.

PubMed

Magnusson, Asa; Aronsson, Per-Olof; Lundgren, Klas; Stenström, Kristina

2008-08-01

This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

Report on the BWR owners group radiation protection/ALARA Committee

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aldrich, L.R.

1995-03-01

Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements inmore » relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming {open_quotes}World Class{close_quotes} performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance.« less

ARCADIA{sup R} - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas

2007-07-01

Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code systemmore » ARCADIA{sup R} and concludes on customer benefits. ARCADIA{sup R} is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA{sup R} system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)« less

Coupling Schemes for Multiphysics Reactor Simulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vijay Mahadeven; Jean Ragusa

2007-11-01

This report documents the progress of the student Vijay S. Mahadevan from the Nuclear Engineering Department of Texas A&M University over the summer of 2007 during his visit to the INL. The purpose of his visit was to investigate the physics-based preconditioned Jacobian-free Newton-Krylov method applied to physics relevant to nuclear reactor simulation. To this end he studied two test problems that represented reaction-diffusion and advection-reaction. These two test problems will provide the basis for future work in which neutron diffusion, nonlinear heat conduction, and a twophase flow model will be tightly coupled to provide an accurate model of amore » BWR core.« less

Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, William BJ J

2016-01-01

A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blademore » histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.« less

Water chemistry control and decontamination experience with TEPCO BWR`s and the measures planned for the future

DOE Office of Scientific and Technical Information (OSTI.GOV)

Suzuki, N.; Miyamaru, K.

1995-03-01

The new TEPCO BWR`s are capable of having the occupational radiation exposure controlled successfully at a low level by selecting low cobalt steel, using corrosion-resistant steel, employing dual condensate polishing systems, and controlling Ni/Fe ratio during operation. The occupational radiation exposure of the old BWR`s, on the other hand, remains high though reduced substantially through the use of low cobalt replacement steel and the partial addition of a filter in the condensate polishing system. Currently under review is the overall decontamination procedure for the old BWR`s to find out to measures needed to reduce the amount of crud that ismore » and has been carried over into the nuclear reactor. The current status of decontamination is reported below.« less

Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R.

2015-08-01

Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramicmore » microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, FeCrAl would tend to generate heat and hydrogen from oxidation at a slower rate compared to the zirconium-based alloys in use today. The previous study, [2], of the FeCrAl ATF concept during station blackout (SBO) severe accident scenarios in BWRs was based on simulating short term SBO (STSBO), long term SBO (LTSBO), and modified SBO scenarios occurring in a BWR-4 reactor with MARK-I containment. The analysis indicated that FeCrAl had the potential to delay the onset of fuel failure by a few hours depending on the scenario, and it could delay lower head failure by several hours. The analysis demonstrated reduced in-vessel hydrogen production. However, the work was preliminary and was based on limited knowledge of material properties for FeCrAl. Limitations of the MELCOR code were identified for direct use in modeling ATF concepts. This effort used an older version of MELCOR (1.8.5). Since these analyses, the BWR model has been updated for use in MELCOR 1.8.6 [10], and more representative material properties for FeCrAl have been modeled. Sections 2 4 present updated analyses for the FeCrAl ATF concept response during severe accidents in a BWR. The purpose of the study is to estimate the potential gains afforded by the FeCrAl ATF concept during BWR SBO scenarios.« less

Investigation of Containment Flooding Strategy for Mark-III Nuclear Power Plant with MAAP4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su Weinian; Wang, S.-J.; Chiang, S.-C

2005-06-15

Containment flooding is an important strategy for severe accident management of a conventional boiling water reactor (BWR) system. The purpose of this work is to investigate the containment flooding strategy of the Mark-III system after a reactor pressure vessel (RPV) breach. The Kuosheng Power Plant is a typical BWR-6 nuclear power plant (NPP) with Mark-III containment. The Severe Accident Management Guideline (SAMG) of the Kuosheng NPP has been developed based on the BWR Owners Group (BWROG) Emergency Procedure and Severe Accident Guidelines, Rev. 2. Therefore, the Kuosheng NPP is selected as the plant for study, and the MAAP4 code ismore » chosen as the tool for analysis. A postulated specific station blackout sequence for the Kuosheng NPP is cited as a reference case for this analysis. Because of the design features of Mark-III containment, the debris in the reactor cavity may not be submerged after an RPV breach when one follows the containment flooding strategy as suggested in the BWROG generic guideline, and the containment integrity could be challenged eventually. A more specific containment flooding strategy with drywell venting after an RPV breach is investigated, and a more stable plant condition is achieved with this strategy. Accordingly, the containment flooding strategy after an RPV breach will be modified for the Kuosheng SAMG, and these results are applicable to typical Mark-III plants with drywell vent path.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Downar, Thomas

This report summarizes the current status of VERA-CS Verification and Validation for PWR Core Follow operation and proposes a multi-phase plan for continuing VERA-CS V&V in FY17 and FY18. The proposed plan recognizes the hierarchical nature of a multi-physics code system such as VERA-CS and the importance of first achieving an acceptable level of V&V on each of the single physics codes before focusing on the V&V of the coupled physics solution. The report summarizes the V&V of each of the single physics codes systems currently used for core follow analysis (ie MPACT, CTF, Multigroup Cross Section Generation, and BISONmore » / Fuel Temperature Tables) and proposes specific actions to achieve a uniformly acceptable level of V&V in FY17. The report also recognizes the ongoing development of other codes important for PWR Core Follow (e.g. TIAMAT, MAMBA3D) and proposes Phase II (FY18) VERA-CS V&V activities in which those codes will also reach an acceptable level of V&V. The report then summarizes the current status of VERA-CS multi-physics V&V for PWR Core Follow and the ongoing PWR Core Follow V&V activities for FY17. An automated procedure and output data format is proposed for standardizing the output for core follow calculations and automatically generating tables and figures for the VERA-CS Latex file. A set of acceptance metrics is also proposed for the evaluation and assessment of core follow results that would be used within the script to automatically flag any results which require further analysis or more detailed explanation prior to being added to the VERA-CS validation base. After the Automation Scripts have been completed and tested using BEAVRS, the VERA-CS plan proposes the Watts Bar cycle depletion cases should be performed with the new cross section library and be included in the first draft of the new VERA-CS manual for release at the end of PoR15. Also, within the constraints imposed by the proprietary nature of plant data, as many as possible of the FY17 AMA Plant Core Follow cases should also be included in the VERA-CS manual at the end of PoR15. After completion of the ongoing development of TIAMAT for fully coupled, full core calculations with VERA-CS / BISON 1.5D, and after the completion of the refactoring of MAMBA3D for CIPS analysis in FY17, selected cases from the VERA-CS validation based should be performed, beginning with the legacy cases of Watts Bar and BEAVRS in PoR16. Finally, as potential Phase III future work some additional considerations are identified for extending the VERA-CS V&V to other reactor types such as the BWR.« less

Recent GE BWR fuel experience and design evolution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, J.E.; Potts, G.A.; Proebstle, R.A.

1992-01-01

Reliable fuel operation is essential to the safe, reliable, and economic power production by today's commercial nuclear reactors. GE Nuclear Energy is committed to maximize fuel reliability through the progressive development of improved fuel design features and dedication to provide the maximum quality of the design features and dedication to provide the maximum quality of the design, fabrication, and operation of GE BWR fuel. Over the last 35 years, GE has designed, fabricated, and placed in operation over 82,000 BWR fuel bundles containing over 5 million fuel rods. This experience includes successful commercial reactor operation of fuel assemblies to greatermore » than 45000 MWd/MTU bundle average exposure. This paper reports that this extensive experience base has enabled clear identification and characterization of the active failure mechanisms. With this failure mechanism characterization, mitigating actions have been developed and implemented by GE to provide the highest reliability BWR fuel bundles possible.« less

Standard-less analysis of Zircaloy clad samples by an instrumental neutron activation method

NASA Astrophysics Data System (ADS)

Acharya, R.; Nair, A. G. C.; Reddy, A. V. R.; Goswami, A.

2004-03-01

A non-destructive method for analysis of irregular shape and size samples of Zircaloy has been developed using the recently standardized k0-based internal mono standard instrumental neutron activation analysis (INAA). The samples of Zircaloy-2 and -4 tubes, used as fuel cladding in Indian boiling water reactors (BWR) and pressurized heavy water reactors (PHWR), respectively, have been analyzed. Samples weighing in the range of a few tens of grams were irradiated in the thermal column of Apsara reactor to minimize neutron flux perturbations and high radiation dose. The method utilizes in situ relative detection efficiency using the γ-rays of selected activation products in the sample for overcoming γ-ray self-attenuation. Since the major and minor constituents (Zr, Sn, Fe, Cr and/or Ni) in these samples were amenable to NAA, the absolute concentrations of all the elements were determined using mass balance instead of using the concentration of the internal mono standard. Concentrations were also determined in a smaller size Zircaloy-4 sample by irradiating in the core position of the reactor to validate the present methodology. The results were compared with literature specifications and were found to be satisfactory. Values of sensitivities and detection limits have been evaluated for the elements analyzed.

Air ingression calculations for selected plant transients using MELCOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kmetyk, L.N.

1994-01-01

Two sets of MELCOR calculations have been completed studying the effects of air ingression on the consequences of various severe accident scenarios. One set of calculations analyzed a station blackout with surge line failure prior to vessel breach, starting from nominal operating conditions; the other set of calculations analyzed a station blackout occurring during shutdown (refueling) conditions. Both sets of analyses were for the Surry plant, a three-loop Westinghouse PWR. For both accident scenarios, a basecase calculation was done, and then repeated with air ingression from containment into the core region following core degradation and vessel failure. In addition tomore » the two sets of analyses done for this program, a similar air-ingression sensitivity study was done as part of a low-power/shutdown PRA, with results summarized here; that PRA study also analyzed a station blackout occurring during shutdown (refueling) conditions, but for the Grand Gulf plant, a BWR/6 with Mark III containment. These studies help quantify the amount of air that would have to enter the core region to have a significant impact on the severe accident scenario, and demonstrate that one effect, of air ingression is substantial enhancement of ruthenium release. These calculations also show that, while the core clad temperatures rise more quickly due to oxidation with air rather than steam, the core also degrades and relocates more quickly, so that no sustained, enhanced core heatup is predicted to occur with air ingression.« less

Application of the IBERDROLA RETRAN Licensing Methodology to the Confrentes BWR-6 110% Extended Power Uprate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fuente, Rafael de la; Iglesias, Javier; Sedano, Pablo G.

IBERDROLA (Spanish utility) and IBERDROLA INGENIERIA (engineering branch) have been developing during the last 2 yr the 110% Extended Power Uprate Project for Cofrentes BWR-6. IBERDROLA has available an in-house design and licensing reload methodology that has been approved in advance by the Spanish Nuclear Regulatory Authority. This methodology has been applied to perform the nuclear design and the reload licensing analysis for Cofrentes cycles 12 and 13 and to develop a significant number of safety analyses of the Cofrentes Extended Power.Because the scope of the licensing process of the Cofrentes Extended Power Uprate exceeds the range of analysis includedmore » in the Cofrentes generic reload licensing process, it has been required to extend the applicability of the Cofrentes RETRAN model to the analysis of new transients. This is the case of the total loss of feedwater (TLFW) transient.The content of this paper shows the benefits of having an in-house design and licensing methodology and describes the process to extend the applicability of the Cofrentes RETRAN model to the analysis of new transients, particularly in this paper the TLFW transient.« less

A two-step method for developing a control rod program for boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

1992-01-01

This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in amore » computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.

Severe accident management can be defined as the use of existing and/or altemative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable ofmore » propagating both data and model uncertainty. It is applied to several potential strategies including PWR cavity flooding, BWR drywell flooding, PWR depressurization and PWR feed and bleed.« less

Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

NASA Astrophysics Data System (ADS)

Tanaka, Ken-ichi

2016-06-01

We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV) of a Boiling Water Reactor (BWR) by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au) and Nickel (Ni) at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE PAGES

Ilas, Germina; Liljenfeldt, Henrik

2017-05-19

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Liljenfeldt, Henrik

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

FY2012 summary of tasks completed on PROTEUS-thermal work.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C.H.; Smith, M.A.

2012-06-06

PROTEUS is a suite of the neutronics codes, both old and new, that can be used within the SHARP codes being developed under the NEAMS program. Discussion here is focused on updates and verification and validation activities of the SHARP neutronics code, DeCART, for application to thermal reactor analysis. As part of the development of SHARP tools, the different versions of the DeCART code created for PWR, BWR, and VHTR analysis were integrated. Verification and validation tests for the integrated version were started, and the generation of cross section libraries based on the subgroup method was revisited for the targetedmore » reactor types. The DeCART code has been reorganized in preparation for an efficient integration of the different versions for PWR, BWR, and VHTR analysis. In DeCART, the old-fashioned common blocks and header files have been replaced by advanced memory structures. However, the changing of variable names was minimized in order to limit problems with the code integration. Since the remaining stability problems of DeCART were mostly caused by the CMFD methodology and modules, significant work was performed to determine whether they could be replaced by more stable methods and routines. The cross section library is a key element to obtain accurate solutions. Thus, the procedure for generating cross section libraries was revisited to provide libraries tailored for the targeted reactor types. To improve accuracy in the cross section library, an attempt was made to replace the CENTRM code by the MCNP Monte Carlo code as a tool obtaining reference resonance integrals. The use of the Monte Carlo code allows us to minimize problems or approximations that CENTRM introduces since the accuracy of the subgroup data is limited by that of the reference solutions. The use of MCNP requires an additional set of libraries without resonance cross sections so that reference calculations can be performed for a unit cell in which only one isotope of interest includes resonance cross sections, among the isotopes in the composition. The OECD MHTGR-350 benchmark core was simulated using DeCART as initial focus of the verification/validation efforts. Among the benchmark problems, Exercise 1 of Phase 1 is a steady-state benchmark case for the neutronics calculation for which block-wise cross sections were provided in 26 energy groups. This type of problem was designed for a homogenized geometry solver like DIF3D rather than the high-fidelity code DeCART. Instead of the homogenized block cross sections given in the benchmark, the VHTR-specific 238-group ENDF/B-VII.0 library of DeCART was directly used for preliminary calculations. Initial results showed that the multiplication factors of a fuel pin and a fuel block with or without a control rod hole were off by 6, -362, and -183 pcm Dk from comparable MCNP solutions, respectively. The 2-D and 3-D one-third core calculations were also conducted for the all-rods-out (ARO) and all-rods-in (ARI) configurations, producing reasonable results. Figure 1 illustrates the intermediate (1.5 eV - 17 keV) and thermal (below 1.5 eV) group flux distributions. As seen from VHTR cores with annular fuels, the intermediate group fluxes are relatively high in the fuel region, but the thermal group fluxes are higher in the inner and outer graphite reflector regions than in the fuel region. To support the current project, a new three-year I-NERI collaboration involving ANL and KAERI was started in November 2011, focused on performing in-depth verification and validation of high-fidelity multi-physics simulation codes for LWR and VHTR. The work scope includes generating improved cross section libraries for the targeted reactor types, developing benchmark models for verification and validation of the neutronics code with or without thermo-fluid feedback, and performing detailed comparisons of predicted reactor parameters against both Monte Carlo solutions and experimental measurements. The following list summarizes the work conducted so far for PROTEUS-Thermal Tasks: Unification of different versions of DeCART was initiated, and at the same time code modernization was conducted to make code unification efficient; (2) Regeneration of cross section libraries was attempted for the targeted reactor types, and the procedure for generating cross section libraries was updated by replacing CENTRM with MCNP for reference resonance integrals; (3) The MHTGR-350 benchmark core was simulated using DeCART with VHTR-specific 238-group ENDF/B-VII.0 library, and MCNP calculations were performed for comparison; and (4) Benchmark problems for PWR and BWR analysis were prepared for the DeCART verification/validation effort. In the coming months, the work listed above will be completed. Cross section libraries will be generated with optimized group structures for specific reactor types.« less

Sediment Budgeting in Dam-Affected Rivers: Assessing the Influence of Damming, Tributaries, and Alluvial Valley Sediment Storage on Sediment Regimes

NASA Astrophysics Data System (ADS)

Wilcox, A. C.; Dekker, F. J.; Riebe, C. S.

2014-12-01

Although sediment supply is recognized as a fundamental driver of fluvial processes, measuring how dams affect sediment regimes and incorporating such knowledge into management strategies remains challenging. To determine the influences of damming, tributary supply, and valley morphology and sediment storage on downstream sediment supply in a dryland river, the Bill Williams River (BWR) in western Arizona, we measured basin erosion rates using cosmogenic nuclide analysis of beryllium-10 (10Be) at sites upstream and downstream of a dam along the BWR, as well as from tributaries downstream of the dam. Riverbed sediment mixing calculations were used to test if the dam, which blocks sediment supply from the upper 85% of the basin's drainage area, increases the proportion of tributary sediment to residual upstream sediment in mainstem samples downstream of the dam. Erosion rates in the BWR watershed are more than twice as large in the upper catchment (136 t km-2 yr-1) than in tributaries downstream of Alamo Dam (61 t km-2 yr-1). Tributaries downstream of the dam have little influence on mainstem sediment dynamics. The effect of the dam on reducing sediment supply is limited, however, because of the presence of large alluvial valleys along the mainstem BWR downstream of the dam that store substantial sediment and mitigate supply reductions from the upper watershed. These inferences, from our 10Be derived erosion rates and mixing calculations, are consistent with field observations of downstream changes in bed material size, which suggest that sediment-deficit conditions are restricted to a 10 km reach downstream of the dam, and limited reservoir bathymetry data. Many studies have suggested that tributary sediment inputs downstream of dams play a key role in mitigating dam-induced sediment deficits, but here we show that in a dryland river with ephemeral tributaries, sediment stored in alluvial valleys can also play a key role and in some cases trumps the role of tributaries.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as thesemore » installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.« less

A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2015-03-01

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoidmore » overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety system components such as the safety relief valve (SRV), the RCIC system, the wet well, and the dry well. The results show reasonable system behaviors while exhibiting rich dynamics such as variable flow rates through RCIC turbine and pump during the SBO transient. The model has the potential to resolve the Fukushima RCIC mystery after adding the off-design two-phase turbine operation model and other additional improvements.« less

Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morikawa, Yoshitake

1995-03-01

The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolantmore » system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.« less

Status Report on Ex-Vessel Coolability and Water Management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, M. T.; Robb, K. R.

Specific to BWR plants, current accident management guidance calls for flooding the drywell to a level of approximately 1.2 m (4 feet) above the drywell floor once vessel breach has been determined. While this action can help to submerge ex-vessel core debris, it can also result in flooding the wetwell and thereby rendering the wetwell vent path unavailable. An alternate strategy is being developed in the industry guidance for responding to the severe accident capable vent Order, EA-13-109. The alternate strategy being proposed would throttle the flooding rate to achieve a stable wetwell water level while preserving the wetwell ventmore » path. The overall objective of this work is to upgrade existing analytical tools (i.e. MELTSPREAD and CORQUENCH - which have been used as part of the DOE-sponsored Fukushima accident analyses) in order to provide flexible, analytically capable, and validated models to support the development of water throttling strategies for BWRs that are aimed at keeping ex-vessel core debris covered with water while preserving the wetwell vent path.« less

Optimum Water Chemistry in radiation field buildup control

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lin, Chien, C.

1995-03-01

Nuclear utilities continue to face the challenGE of reducing exposure of plant maintenance personnel. GE Nuclear Energy has developed the concept of Optimum Water Chemistry (OWC) to reduce the radiation field buildup and minimize the radioactive waste production. It is believed that reduction of radioactive sources and improvement of the water chemistry quality should significantly reduce both the radiation exposure and radwaste production. The most important source of radioactivity is cobalt and replacement of cobalt containing alloy in the core region as well as in the entire primary system is considered the first priority to achieve the goal of lowmore » exposure and minimized waste production. A plant specific computerized cobalt transport model has been developed to evaluate various options in a BWR system under specific conditions. Reduction of iron input and maintaining low ionic impurities in the coolant have been identified as two major tasks for operators. Addition of depleted zinc is a proven technique to reduce Co-60 in reactor water and on out-of-core piping surfaces. The effect of HWC on Co-60 transport in the primary system will also be discussed.« less

Fuel Performance Calculations for FeCrAl Cladding in BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

George, Nathan; Sweet, Ryan; Maldonado, G. Ivan

2015-01-01

This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behaviormore » of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.« less

RAMONA-3B application to Browns Ferry ATWS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Slovik, G.C.; Neymotin, L.; Cazzoli, E.

1984-01-01

This paper discusses two preliminary MSIV clsoure ATWS calculations done using the RAMONA-3B code and the work being done to create the necessary cross section sets for the Browns Ferry Unit 1 reactor. The RAMONA-3B code employs a three-dimensional neutron kinetics model coupled with one-dimensional, four equation, nonhomogeneous, nonequilibrium thermal hydraulics. To be compatible with 3-D neutron kinetics, the code uses parallel coolant channels in the core. It also includes a boron transport model and all necessary BWR components such as jet pump, recirculation pump, steam separator, steamline with safety and relief valves, main steam isolation valve, turbine stop valve,more » and turbine bypass valve. A summary of RAMONA-3B neutron kinetics and thermal hydraulics models is presented in the Appendix.« less

Heat up and failure of BWR upper internals during a severe accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R.

In boiling water reactors, the shroud dome, separators, and dryers above the core are made of approximately 100,000 kg of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. In this scenario, the upper internals can also be heated by thermal radiation from the hot degrading core. Historically, models of the upper internals have been relatively simple in severe accident codes. The upper internals are typically modeled in MELCOR as two lumped volumes with simplifiedmore » heat transfer characteristics and no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. The MELCOR modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. Consistent with a previous study, the results indicate that the upper internals can reach high temperatures during a severe accident sufficient to lose their structural integrity and relocate. Finally, the additional 100 metric tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.« less

Heat up and failure of BWR upper internals during a severe accident

DOE PAGES

Robb, Kevin R.

2017-02-21

In boiling water reactors, the shroud dome, separators, and dryers above the core are made of approximately 100,000 kg of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. In this scenario, the upper internals can also be heated by thermal radiation from the hot degrading core. Historically, models of the upper internals have been relatively simple in severe accident codes. The upper internals are typically modeled in MELCOR as two lumped volumes with simplifiedmore » heat transfer characteristics and no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. The MELCOR modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. Consistent with a previous study, the results indicate that the upper internals can reach high temperatures during a severe accident sufficient to lose their structural integrity and relocate. Finally, the additional 100 metric tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.« less

SiC Composite for Fuel Structure Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yueh, Ken

Extensive evaluation was performed to determine the suitability of using SiC composite as a boiling water reactor (BWR) fuel channel material. A thin walled SiC composite box, 10 cm in dimension by approximately 1.5 mm wall thickness was fabricated using chemical vapor deposition (CVD) for testing. Mechanical test results and performance evaluations indicate the material could meet BWR channel mechanical design requirement. However, large mass loss of up to 21% was measured in in-pile corrosion test under BWR-like conditions in under 3 months of irradiation. A fresh sister sample irradiated in a follow-up cycle under PWR conditions showed no measureablemore » weight loss and thus supports the hypothesis that the oxidizing condition of the BWR-like coolant chemistry was responsible for the high corrosion rate. A thermodynamic evaluation showed SiC is not stable and the material may oxidize to form SiO 2 and CO 2. Silica has demonstrated stability in high temperature steam environment and form a protective oxide layer under severe accident conditions. However, it does not form a protective layer in water under normal BWR operational conditions due to its high solubility. Corrosion product stabilization by modifying the SiC CVD surface is an approach evaluated in this study to mitigate the high corrosion rate. Titanium and zirconium have been selected as stabilizing elements since both TiSiO 4 and ZrSiO 4 are insoluble in water. Corrosion test results in oxygenated water autoclave indicate TiSiO4 does not form a protective layer. However, zirconium doped test samples appear to form a stable continuous layer of ZrSiO 4 during the corrosion process. Additional process development is needed to produce a good ZrSiC coating to verify functionality of the mitigation concept.« less

Heat up and potential failure of BWR upper internals during a severe accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R

2015-01-01

In boiling water reactors, the steam dome, steam separators, and dryers above the core are comprised of approximately 100 tons of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. Historically, the upper internals have been modeled using severe accident codes with relatively simple approximations. The upper internals are typically modeled in MELCOR as two lumped volumes with simplified heat transfer characteristics, with no structural integrity considerations, and with limited ability to oxidize, melt, andmore » relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. This modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. The results indicate that the upper internals can reach high temperatures during a severe accident; they are predicted to reach a high enough temperature such that they lose their structural integrity and relocate. The additional 100 tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.« less

Interim reliability evaluation program, Browns Ferry 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mays, S.E.; Poloski, J.P.; Sullivan, W.H.

1981-01-01

Probabilistic risk analysis techniques, i.e., event tree and fault tree analysis, were utilized to provide a risk assessment of the Browns Ferry Nuclear Plant Unit 1. Browns Ferry 1 is a General Electric boiling water reactor of the BWR 4 product line with a Mark 1 (drywell and torus) containment. Within the guidelines of the IREP Procedure and Schedule Guide, dominant accident sequences that contribute to public health and safety risks were identified and grouped according to release categories.

Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay,more » and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernander, O.; Haga, I.; Segerberg, F.

BS>From international nuclear industries fair; Basel, Switzerland (16 Oct 1972). Although the present status of the boiling water reactor is one of proven technology, design refinements and technical innovations are still being made to further improve reliability, economy and safety. The new standard ASEA- ATOM BWR features a number of such refinements and design improvements involving main circulation punips, containment design, refuelling system and off-gas treatment plant. In some respects the nuclear and hydraulic design of the ASEA- ATOM BWR differs from that adopted by other BWR manufacturers. Since the Oskarshamn I plant was the first nuclear power station havingmore » these features an extensive physics and hydraulics test program was made during the reactor start- up. The results of these tests have fully confirmed the ability of calculation methods to predict the behavior of the reactor. (auth)« less

Real time monitoring of environmental crack growth in BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hale, D.; Diehl, C.G.

1988-01-01

A comprehensive field test program was recently completed at several Boiling Water Reactors (BWR) to quantify the effect of coolant impurities on the initiation and growth of stress corrosion cracks. A new technology was utilized which allows for real time monitoring of stress corrosion crack growth rates. The BWR environments were characterized using Ion Chromatography and Electro Chemical Potential (ECP) measurements. The effects of typical water chemistry transients and startups were quantified.

Advanced Neutronics Tools for BWR Design Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Santamarina, A.; Hfaiedh, N.; Letellier, R.

2006-07-01

This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007more » BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy. (authors)« less

Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martinez-Gonzalez, Jesus S.; Ade, Brian J.; Bowman, Stephen M.

2015-01-01

Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational datamore » available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luk, V.K.; Hessheimer, M.F.; Matsumoto, T.

A high pressure test of a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of a steel containment vessel (SCV), representing an improved boiling water reactor (BWR) Mark II containment, was conducted on December 11--12, 1996 at Sandia National Laboratories. This paper describes the preliminary results of the high pressure test. In addition, the preliminary post-test measurement data and the preliminary comparison of test data with pretest analysis predictions are also presented.

Statistical evaluation of the metallurgical test data in the ORR-PSF-PVS irradiation experiment. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stallmann, F.W.

1984-08-01

A statistical analysis of Charpy test results of the two-year Pressure Vessel Simulation metallurgical irradiation experiment was performed. Determination of transition temperature and upper shelf energy derived from computer fits compare well with eyeball fits. Uncertainties for all results can be obtained with computer fits. The results were compared with predictions in Regulatory Guide 1.99 and other irradiation damage models.

Analyzing simulation-based PRA data through traditional and topological clustering: A BWR station blackout case study

DOE PAGES

Maljovec, D.; Liu, S.; Wang, B.; ...

2015-07-14

Here, dynamic probabilistic risk assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP and MELCOR) with simulation controller codes (e.g., RAVEN and ADAPT). Whereas system simulator codes model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic and operating procedures) and stochastic (e.g., component failures and parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by sampling values of a set of parameters and simulating the system behavior for that specific set of parameter values. For complex systems, a major challenge in using DPRA methodologies is to analyze the large number of scenarios generated,more » where clustering techniques are typically employed to better organize and interpret the data. In this paper, we focus on the analysis of two nuclear simulation datasets that are part of the risk-informed safety margin characterization (RISMC) boiling water reactor (BWR) station blackout (SBO) case study. We provide the domain experts a software tool that encodes traditional and topological clustering techniques within an interactive analysis and visualization environment, for understanding the structures of such high-dimensional nuclear simulation datasets. We demonstrate through our case study that both types of clustering techniques complement each other for enhanced structural understanding of the data.« less

PWR and BWR spent fuel assembly gamma spectra measurements

NASA Astrophysics Data System (ADS)

Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

2016-10-01

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

PWR and BWR spent fuel assembly gamma spectra measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

PWR and BWR spent fuel assembly gamma spectra measurements

DOE PAGES

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

2016-07-17

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

Computer program for optimal BWR congtrol rod programming

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taner, M.S.; Levine, S.H.; Carmody, J.M.

1995-12-31

A fully automated computer program has been developed for designing optimal control rod (CR) patterns for boiling water reactors (BWRs). The new program, called OCTOPUS-3, is based on the OCTOPUS code and employs SIMULATE-3 (Ref. 2) for the analysis. There are three aspects of OCTOPUS-3 that make it successful for use at PECO Energy. It incorporates a new feasibility algorithm that makes the CR design meet all constraints, it has been coupled to a Bourne Shell program 3 to allow the user to run the code interactively without the need for a manual, and it develops a low axial peakmore » to extend the cycle. For PECO Energy Co.`s limericks it increased the energy output by 1 to 2% over the traditional PECO Energy design. The objective of the optimization in OCTOPUS-3 is to approximate a very low axial peaked target power distribution while maintaining criticality, keeping the nodal and assembly peaks below the allowed maximum, and meeting the other constraints. The user-specified input for each exposure point includes: CR groups allowed-to-move, target k{sub eff}, and amount of core flow. The OCTOPUS-3 code uses the CR pattern from the previous step as the initial guess unless indicated otherwise.« less

Irradiation-induced sensitization and stress corrosion cracking of Type 304 stainless steel core-internal components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, H.M.; Ruther, W.E.; Sanecki, J.E.

1991-08-01

High- and commercial-purity heats of Type 304 stainless steel, obtained from neutron absorber tubes after irradiation to fluence levels of up to 2 {times} 10{sup 21} n{center dot}cm{sup {minus}2} (E > 1 MeV) in two boiling water reactors, were examined by Auger electron spectroscopy to characterize irradiation-induced grain- boundary segregation and depletion of alloying and impurity elements. Segregation of Si, P, Ni, and an unidentified element or compound that gives rise to an Auger energy peak at 59 eV was observed in the commercial-purity heat. Such segregation was negligible in high-purity material, except for Ni. No evidence of S segregationmore » was observed in either material. Cr depletion was more pronounced in the high-purity material than in the commercial-purity material. These observations suggest a synergism between the significant level of impurities and Cr depletion in the commercial-purity heat. In the absence of such synergism, Cr depletion appears more pronounced in the high-purity heat. Initial results of constant-extension-rate tests conducted on the two heats in air an in simulated BWR water were correlated with the results from analysis by Auger electron spectroscopy. 15 refs., 10 figs.« less

Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data

NASA Astrophysics Data System (ADS)

Rochman, Dimitri A.; Vasiliev, Alexander; Dokhane, Abdelhamid; Ferroukhi, Hakim

2018-05-01

This paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal.

Radiation field control at the latest BWR plants -- design principle, operational experience and future subjects

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uchida, Shunsuke; Ohsumi, Katsumi; Takashima, Yoshie

1995-03-01

Improvements of operational procedures to control water chemistry, e.g., nickel/iron control, as well as application of hardware improvements for reducing radioactive corrosion products resulted in an extremely low occupational exposure of less than 0.5 man.Sv/yr without any serious impact on the radwaste system, for BWR plants involved in the Japanese Improvement and Standardization Program. Recently, {sup 60}C radioactively in the reactor water has been increasing due to less crud fixation on the two smooth surfaces of new type high performance fuels and to the pH drop caused by chromium oxide anions released from stainless steel structures and pipings. This increasemore » must be limited by changes in water chemistry, e.g., applications of modified nickel/iron ratio control and weak alkali control. Controlled water chemistry to optimize three points, the plant radiation level and integrities of fuel and structural materials, is the primary future subject for BWR water chemistry.« less

TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng L. Y.; Baek J.; Cuadra,A.

2013-11-10

A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic tomore » initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.« less

Heuristic rules embedded genetic algorithm for in-core fuel management optimization

NASA Astrophysics Data System (ADS)

Alim, Fatih

The objective of this study was to develop a unique methodology and a practical tool for designing loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and BP placement in order to achieve maximum cycle length while satisfying the safety constraints. Genetic Algorithms (GA) have been already used to solve this problem for LP optimization for both PWR and Boiling Water Reactor (BWR). The GA, which is a stochastic method works with a group of solutions and uses random variables to make decisions. Based on the theories of evaluation, the GA involves natural selection and reproduction of the individuals in the population for the next generation. The GA works by creating an initial population, evaluating it, and then improving the population by using the evaluation operators. To solve this optimization problem, a LP optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed in the framework of this thesis. This code is applicable for all types of PWR cores having different geometries and structures with an unlimited number of FA types in the inventory. To reach this goal, an innovative GA is developed by modifying the classical representation of the genotype. To obtain the best result in a shorter time, not only the representation is changed but also the algorithm is changed to use in-core fuel management heuristics rules. The improved GA code was tested to demonstrate and verify the advantages of the new enhancements. The developed methodology is explained in this thesis and preliminary results are shown for the VVER-1000 reactor hexagonal geometry core and the TMI-1 PWR. The improved GA code was tested to verify the advantages of new enhancements. The core physics code used for VVER in this research is Moby-Dick, which was developed to analyze the VVER by SKODA Inc. The SIMULATE-3 code, which is an advanced two-group nodal code, is used to analyze the TMI-1.

Impact of nonabsorbing control rod tips on kinetics feedback for BWR turbine trip without bypass RETRAN-03 analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Knerr, R.; Shoop, U.

1993-01-01

RETRAN-03 studies were performed for the boiling water reactor (BWR) turbine trip without bypass (TTWOB) event to investigate how the non-neutron-absorbing material on control rod tips affect scram delay timing and reactivity feedback. Scram delay, Doppler temperature, and moderator void (density) feedback were varied to assess their relative impact on kinetics behavior. Although a generic point-kinetics RETRAN-03 TTWOB model 2 was employed, actual plant information was used to develop the basic and parametric cases.

Performance of iron-chromium-aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

NASA Astrophysics Data System (ADS)

Zhong, Weicheng; Mouche, Peter A.; Han, Xiaochun; Heuser, Brent J.; Mandapaka, Kiran K.; Was, Gary S.

2016-03-01

Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe-Zr is addressed with the FeCrAl-YSZ system.

Load Variation Influences on Joint Work During Squat Exercise in Reduced Gravity

NASA Technical Reports Server (NTRS)

DeWitt, John K.; Fincke, Renita S.; Logan, Rachel L.; Guilliams, Mark E.; Ploutz-Snyder, Lori L.

2011-01-01

Resistance exercises that load the axial skeleton, such as the parallel squat, are incorporated as a critical component of a space exercise program designed to maximize the stimuli for bone remodeling and muscle loading. Astronauts on the International Space Station perform regular resistance exercise using the Advanced Resistive Exercise Device (ARED). Squat exercises on Earth entail moving a portion of the body weight plus the added bar load, whereas in microgravity the body weight is 0, so all load must be applied via the bar. Crewmembers exercising in microgravity currently add approx.70% of their body weight to the bar load as compensation for the absence of the body weight. This level of body weight replacement (BWR) was determined by crewmember feedback and personal experience without any quantitative data. The purpose of this evaluation was to utilize computational simulation to determine the appropriate level of BWR in microgravity necessary to replicate lower extremity joint work during squat exercise in normal gravity based on joint work. We hypothesized that joint work would be positively related to BWR load.

Phenomenology of BWR fuel assembly degradation

NASA Astrophysics Data System (ADS)

Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

2018-03-01

Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

SINGLE PHASE ANALYTICAL MODELS FOR TERRY TURBINE NOZZLE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

All BWR RCIC (Reactor Core Isolation Cooling) systems and PWR AFW (Auxiliary Feed Water) systems use Terry turbine, which is composed of the wheel with turbine buckets and several groups of fixed nozzles and reversing chambers inside the turbine casing. The inlet steam is accelerated through the turbine nozzle and impacts on the wheel buckets, generating work to drive the RCIC pump. As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC systems in Fukushima accidents and extend BWR RCIC and PWR AFW operational range and flexibility, mechanistic models for the Terry turbine, based on Sandiamore » National Laboratories’ original work, has been developed and implemented in the RELAP-7 code to simulate the RCIC system. RELAP-7 is a new reactor system code currently under development with the funding support from U.S. Department of Energy. The RELAP-7 code is a fully implicit code and the preconditioned Jacobian-free Newton-Krylov (JFNK) method is used to solve the discretized nonlinear system. This paper presents a set of analytical models for simulating the flow through the Terry turbine nozzles when inlet fluid is pure steam. The implementation of the models into RELAP-7 will be briefly discussed. In the Sandia model, the turbine bucket inlet velocity is provided according to a reduced-order model, which was obtained from a large number of CFD simulations. In this work, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine bucket inlet. The models include both adiabatic expansion process inside the nozzle and free expansion process out of the nozzle to reach the ambient pressure. The combined models are able to predict the steam mass flow rate and supersonic velocity to the Terry turbine bucket entrance, which are the necessary input conditions for the Terry Turbine rotor model. The nozzle analytical models were validated with experimental data and benchmarked with CFD simulations. The analytical models generally agree well with the experimental data and CFD simulations. The analytical models are suitable for implementation into a reactor system analysis code or severe accident code as part of mechanistic and dynamical models to understand the RCIC behaviors. The cases with two-phase flow at the turbine inlet will be pursued in future work.« less

Verification of BWR Turbine Skyshine Dose with the MCNP5 Code Based on an Experiment Made at SHIMANE Nuclear Power Station

NASA Astrophysics Data System (ADS)

Tayama, Ryuichi; Wakasugi, Kenichi; Kawanaka, Ikunori; Kadota, Yoshinobu; Murakami, Yasuhiro

We measured the skyshine dose from turbine buildings at Shimane Nuclear Power Station Unit 1 (NS-1) and Unit 2 (NS-2), and then compared it with the dose calculated with the Monte Carlo transport code MCNP5. The skyshine dose values calculated with the MCNP5 code agreed with the experimental data within a factor of 2.8, when the roof of the turbine building was precisely modeled. We concluded that our MCNP5 calculation was valid for BWR turbine skyshine dose evaluation.

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE PAGES

Coble, Jamie; Orton, Christopher; Schwantes, Jon

2017-01-17

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coble, Jamie; Orton, Christopher; Schwantes, Jon

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Passive gamma analysis of the boiling-water-reactor assemblies

NASA Astrophysics Data System (ADS)

Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

2016-09-01

This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

Passive gamma analysis of the boiling-water-reactor assemblies

DOE PAGES

Vo, D.; Favalli, A.; Grogan, B.; ...

2016-09-01

This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in themore » past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.« less

TEM/STEM study of Zircaloy-2 with protective FeAl(Cr) layers under simulated BWR environment and high-temperature steam exposure

NASA Astrophysics Data System (ADS)

Park, Donghee; Mouche, Peter A.; Zhong, Weicheng; Mandapaka, Kiran K.; Was, Gary S.; Heuser, Brent J.

2018-04-01

FeAl(Cr) thin-film depositions on Zircaloy-2 were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) with respect to oxidation behavior under simulated boiling water reactor (BWR) conditions and high-temperature steam. Columnar grains of FeAl with Cr in solid solution were formed on Zircaloy-2 coupons using magnetron sputtering. NiFe2O4 precipitates on the surface of the FeAl(Cr) coatings were observed after the sample was exposed to the simulated BWR environment. High-temperature steam exposure resulted in grain growth and consumption of the FeAl(Cr) layer, but no delamination at the interface. Outward Al diffusion from the FeAl(Cr) layer occurred during high-temperature steam exposure (700 °C for 3.6 h) to form a 100-nm-thick alumina oxide layer, which was effective in mitigating oxidation of the Zircaloy-2 coupons. Zr intermetallic precipitates formed near the FeAl(Cr) layer due to the inward diffusion of Fe and Al. The counterflow of vacancies in response to the Al and Fe diffusion led to porosity within the FeAl(Cr) layer.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

Recent plant studies using Victoria 2.0

DOE Office of Scientific and Technical Information (OSTI.GOV)

BIXLER,NATHAN E.; GASSER,RONALD D.

2000-03-08

VICTORIA 2.0 is a mechanistic computer code designed to analyze fission product behavior within the reactor coolant system (RCS) during a severe nuclear reactor accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS and secondary circuits. These predictions account for the chemical and aerosol processes that affect radionuclide behavior. VICTORIA 2.0 was released in early 1999; a new version VICTORIA 2.1, is now under development. The largest improvements in VICTORIA 2.1 are connected with the thermochemical database, which is being revised andmore » expanded following the recommendations of a peer review. Three risk-significant severe accident sequences have recently been investigated using the VICTORIA 2.0 code. The focus here is on how various chemistry options affect the predictions. Additionally, the VICTORIA predictions are compared with ones made using the MELCOR code. The three sequences are a station blackout in a GE BWR and steam generator tube rupture (SGTR) and pump-seal LOCA sequences in a 3-loop Westinghouse PWR. These sequences cover a range of system pressures, from fully depressurized to full system pressure. The chief results of this study are the fission product fractions that are retained in the core, RCS, secondary, and containment and the fractions that are released into the environment.« less

Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

NASA Astrophysics Data System (ADS)

Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

2017-01-01

The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ott, Larry J.; Howell, Michael; Robb, Kevin R.

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristicsmore » are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate constant (3 times and 10 times that of the rate constant for APMT) had a negligible impact on the early stages of the accident and minor impacts on the accident progression after the first relocation of the fuel. At temperatures below 1,500°C, increasing the rate constant for APMT by a factor of 10 still resulted in only minor FeCrAl oxidation. In general, the gains afforded by the FeCrAl enhanced ATF concept with respect to accident sequence timing and combustible gas generation are consistent with previous efforts. Compared with the traditional Zircaloy-based cladding and channel box system, the FeCrAl concept could provide a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. For example, a station blackout was simulated in which cooling water injection was lost 36 hours after shutdown. The timing to first fuel relocation was delayed by approximately 5 h for the FeCrAl ATF concept compared with that of the traditional Zircaloy-based cladding and channel box system.« less

BWR Servicing and Refueling Improvement Program: Phase I summary report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perry, D.R.

1978-09-01

Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was tomore » identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.« less

A Pumping Algorithm for Ergodic Stochastic Mean Payoff Games with Perfect Information

NASA Astrophysics Data System (ADS)

Boros, Endre; Elbassioni, Khaled; Gurvich, Vladimir; Makino, Kazuhisa

In this paper, we consider two-person zero-sum stochastic mean payoff games with perfect information, or BWR-games, given by a digraph G = (V = V B ∪ V W ∪ V R , E), with local rewards r: E to { R}, and three types of vertices: black V B , white V W , and random V R . The game is played by two players, White and Black: When the play is at a white (black) vertex v, White (Black) selects an outgoing arc (v,u). When the play is at a random vertex v, a vertex u is picked with the given probability p(v,u). In all cases, Black pays White the value r(v,u). The play continues forever, and White aims to maximize (Black aims to minimize) the limiting mean (that is, average) payoff. It was recently shown in [7] that BWR-games are polynomially equivalent with the classical Gillette games, which include many well-known subclasses, such as cyclic games, simple stochastic games (SSG's), stochastic parity games, and Markov decision processes. In this paper, we give a new algorithm for solving BWR-games in the ergodic case, that is when the optimal values do not depend on the initial position. Our algorithm solves a BWR-game by reducing it, using a potential transformation, to a canonical form in which the optimal strategies of both players and the value for every initial position are obvious, since a locally optimal move in it is optimal in the whole game. We show that this algorithm is pseudo-polynomial when the number of random nodes is constant. We also provide an almost matching lower bound on its running time, and show that this bound holds for a wider class of algorithms. Let us add that the general (non-ergodic) case is at least as hard as SSG's, for which no pseudo-polynomial algorithm is known.

Ecosystem effects of environmental flows: Modelling and experimental floods in a dryland river

USGS Publications Warehouse

Shafroth, P.B.; Wilcox, A.C.; Lytle, D.A.; Hickey, J.T.; Andersen, D.C.; Beauchamp, Vanessa B.; Hautzinger, A.; McMullen, L.E.; Warner, A.

2010-01-01

Successful environmental flow prescriptions require an accurate understanding of the linkages among flow events, geomorphic processes and biotic responses. We describe models and results from experimental flow releases associated with an environmental flow program on the Bill Williams River (BWR), Arizona, in arid to semiarid western U.S.A. Two general approaches for improving knowledge and predictions of ecological responses to environmental flows are: (1) coupling physical system models to ecological responses and (2) clarifying empirical relationships between flow and ecological responses through implementation and monitoring of experimental flow releases. We modelled the BWR physical system using: (1) a reservoir operations model to simulate reservoir releases and reservoir water levels and estimate flow through the river system under a range of scenarios, (2) one- and two-dimensional river hydraulics models to estimate stage-discharge relationships at the whole-river and local scales, respectively, and (3) a groundwater model to estimate surface- and groundwater interactions in a large, alluvial valley on the BWR where surface flow is frequently absent. An example of a coupled, hydrology-ecology model is the Ecosystems Function Model, which we used to link a one-dimensional hydraulic model with riparian tree seedling establishment requirements to produce spatially explicit predictions of seedling recruitment locations in a Geographic Information System. We also quantified the effects of small experimental floods on the differential mortality of native and exotic riparian trees, on beaver dam integrity and distribution, and on the dynamics of differentially flow-adapted benthic macroinvertebrate groups. Results of model applications and experimental flow releases are contributing to adaptive flow management on the BWR and to the development of regional environmental flow standards. General themes that emerged from our work include the importance of response thresholds, which are commonly driven by geomorphic thresholds or mediated by geomorphic processes, and the importance of spatial and temporal variation in the effects of flows on ecosystems, which can result from factors such as longitudinal complexity and ecohydrological feedbacks. ?? Published 2009.

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sweet, Ryan; George, Nathan M.; Terrani, Kurt A.

2016-08-30

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling themore » integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.« less

WHEN MODEL MEETS REALITY – A REVIEW OF SPAR LEVEL 2 MODEL AGAINST FUKUSHIMA ACCIDENT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhegang Ma

The Standardized Plant Analysis Risk (SPAR) models are a set of probabilistic risk assessment (PRA) models used by the Nuclear Regulatory Commission (NRC) to evaluate the risk of operations at U.S. nuclear power plants and provide inputs to risk informed regulatory process. A small number of SPAR Level 2 models have been developed mostly for feasibility study purpose. They extend the Level 1 models to include containment systems, group plant damage states, and model containment phenomenology and accident progression in containment event trees. A severe earthquake and tsunami hit the eastern coast of Japan in March 2011 and caused significantmore » damages on the reactors in Fukushima Daiichi site. Station blackout (SBO), core damage, containment damage, hydrogen explosion, and intensive radioactivity release, which have been previous analyzed and assumed as postulated accident progression in PRA models, now occurred with various degrees in the multi-units Fukushima Daiichi site. This paper reviews and compares a typical BWR SPAR Level 2 model with the “real” accident progressions and sequences occurred in Fukushima Daiichi Units 1, 2, and 3. It shows that the SPAR Level 2 model is a robust PRA model that could very reasonably describe the accident progression for a real and complicated nuclear accident in the world. On the other hand, the comparison shows that the SPAR model could be enhanced by incorporating some accident characteristics for better representation of severe accident progression.« less

The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, M. T.

MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input thatmore » describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of improvements that are documented in this report have been specifically implemented to support industry in developing Severe Accident Water Management (SAWM) strategies for Boiling Water Reactors.« less

Status update of the BWR cask simulator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindgren, Eric R.; Durbin, Samuel G.

2015-09-01

The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximummore » thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on heat load and the effect of simulated wind on a simplified below ground vent configuration.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hznnera, K.; Hetzler, F.; Hyden, L.

From international nuclear industries fair; Basel, Switzerland (16 Oct 1972). Some features of ASEA-ATOM's BWR fuel design and fabrication processes are given. The in pile fuel performance experience to date is reviewed. (auth)

Round Robin Analyses of the Steel Containment Vessel Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Costello, J.F.; Hashimote, T.; Klamerus, E.W.

A high pressure test of the steel containment vessel (SCV) model was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. Several organizations from the US, Europe, and Asia were invited to participate in a Round Robin analysis to perform independent pretest predictions and posttest evaluations of the behavior of the SCV model during the high pressure test. Both pretest and posttest analysis results from all Round Robin participants were compared tomore » the high pressure test data. This paper summarizes the Round Robin analysis activities and discusses the lessons learned from the collective effort.« less

Characterization of cracking behavior using posttest fractographic analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobayashi, T.; Shockey, D.A.

A determination of time to initiation of stress corrosion cracking in structures and test specimens is important for performing structural failure analysis and for setting inspection intervals. Yet it is seldom possible to establish how much of a component's lifetime represents the time to initiation of fracture and how much represents postinitiation crack growth. This exploratory research project was undertaken to examine the feasibility of determining crack initiation times and crack growth rates from posttest examination of fracture surfaces of constant-extension-rate-test (CERT) specimens by using the fracture reconstruction applying surface topography analysis (FRASTA) technique. The specimens used in this studymore » were Type 304 stainless steel fractured in several boiling water reactor (BWR) aqueous environments. 2 refs., 25 figs., 2 tabs.« less

Spent fuel burnup estimation by Cerenkov glow intensity measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuribara, Masayuki

1994-10-01

The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to themore » declared burnups for the tested spent fuel assemblies for cooling times ranging from 900--2.000 d.« less

In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless Steel under Simulated BWR Condition in JMTR

NASA Astrophysics Data System (ADS)

Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka

The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.

Condensate polisher prefiltration study for Laguna Verde Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garcia, A.; Oyen, L.C.; Nelson, R.A.

1995-05-01

This paper describes an analysis of the iron and copper in the condensate and the technical and economic assessment of the installation of condensate polisher prefilters in Comision Federal de Electricidad`s Laguna Verde Nuclear Generating Station (LVNGS) north of Veracruz, Mexico. LVNGS is a 654 MWe General Electric BWR plant; Unit 1 has been in commercial operation since July, 1990, and Unit 2 is scheduled to become operational in June, 1995. The primary purpose of this study was to (1) analyze the high iron and copper concentrations in the condensate and feedwater, (2) identify, assess, and evaluate techniques to reducemore » the iron and copper concentrations, and (3) perform a cost-benefit analysis of the installation of implementing the appropriate techniques.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adamov, E.O.; Kuklin, A.N.; Mityaev, Yu.I.

The nuclear power plants with boiling water reactors of improved safety are being developed. There is 26 years of operating experience with the plant VK-50 in Dimitrovgrad. The design and operation of the BWR reactors are described.

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

NASA Astrophysics Data System (ADS)

González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

2015-10-01

The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

Enhancing BWR proliferation resistance fuel with minor actinides

NASA Astrophysics Data System (ADS)

Chang, Gray S.

2009-03-01

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm 3) to the top (0.35 g/cm 3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides ( 237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate-term of nuclear energy reconnaissance.

Flow-induced vibration and fretting-wear damage in a moisture separator reheater

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pettigrew, M.J.; Taylor, C.E.; Fisher, N.J.

1996-12-01

Tube failures due to excessive flow-induced vibration were experienced in the tube bundles of moisture separator reheaters in a BWR nuclear station. This paper presents the results of a root cause analysis and covers recommendations for continued operation and for replacement tube bundles. The following tasks are discussed: tube failure analysis; flow velocity distribution calculations; flow-induced vibration analysis with particular emphasis on finned-tubes; fretting-wear testing of a tube and tube-support material combination under simulated operating conditions; field measurements of flow-induced vibration; and development of vibration specifications for replacement tube bundles. The effect of transient operating conditions and of other operationalmore » changes such as tube fouling were considered in the analysis. This paper outlines a typical field problem and illustrates the application of flow-induced vibration technology for the solution of a practical problem.« less

Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Majed, M.; Morback, G.; Wiman, P.

1995-09-01

The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give largemore » advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.« less

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1995-08-01

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized intomore » six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.« less

RELAP-7 Development Updates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

Generation of the V4.2m5 and AMPX and MPACT 51 and 252-Group Libraries with ENDF/B-VII.0 and VII.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kang Seog

The evaluated nuclear data file (ENDF)/B-7.0 v4.1m3 MPACT 47-group library has been used as a main library for the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronics simulator in simulating pressurized water reactor (PWR) problems. Recent analysis for the high void boiling water reactor (BWR) fuels and burnt fuels indicates that the 47-group library introduces relatively large reactivity bias. Since the 47- group structure does not match with the SCALE 6.2 252-group boundaries, the CASL Virtual Environment for Reactor Applications Core Simulator (VERA-CS) MPACT library must be maintained independently, which causes quality assurance concerns. In order to addressmore » this issue, a new 51-group structure has been proposed based on the MPACT 47- g and SCALE 252-g structures. In addition, the new CASL library will include a 19-group structure for gamma production and interaction cross section data based on the SCALE 19- group structure. New AMPX and MPACT 51-group libraries have been developed with the ENDF/B-7.0 and 7.1 evaluated nuclear data. The 19-group gamma data also have been generated for future use, but they are only available on the AMPX 51-g library. In addition, ENDF/B-7.0 and 7.1 MPACT 252-g libraries have been generated for verification purposes. Various benchmark calculations have been performed to verify and validate the newly developed libraries.« less

Job analysis of maintenance-mechanic position for the nuclear power plant maintenance personnel reliability model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Siegel, A.I.; Bartter, W.D.; Kopstein, F.F.

1982-06-01

The task list method of job survey was used. In collaboration with BWR and PWR personnel, a list of 107 tasks performed by maintenance mechanics was developed, grouped into: remove and install, test and repair, inspect and perform preventive maintenance, miscellaneous, communication, and report preparation. For each listed task, the questionnaire form inquired into: frequency of performance, task completion time, safety consequences of improper performance, and the amount of training required to perform the task proficiently. Scaled information was requested about seven abilities: (1) visual speed, accuracy, and recognition; (2) gross motor coordination; (3) fine manual dexterity; (4) strength andmore » stamina; (5) cognition; (6) memory; and (7) problem solving required for function completion. Survey forms were distributed to 27 nuclear power plants. Thirty-one maintenance mechanics representing 17 plants returned the completed forms. Frequency of performing tasks was bimodally distributed: (1) between once a year and once every six months, and (2) about once a week. More than half of the tasks have potential risk consequences if improperly performed. The five tasks with the greatest risk implications in the case of inadequate performance were: (1) remove and install reactor and dry-well heads, (2) test and repair reactor system components, (3) remove and install pressurizer mechanical relief valves, (4) test and repair pressurizer relief valves, (5) remove and install core spray pumps, seals, and valves. Hierarchically, the public risk associated with the various functions was: (1) remove and install, (2) test and repair, (3) preventive maintenance, (4) miscellaneous tasks, (5) communication, and (6) report preparation.« less

CPR methodology with new steady-state criterion and more accurate statistical treatment of channel bow

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baumgartner, S.; Bieli, R.; Bergmann, U. C.

2012-07-01

An overview is given of existing CPR design criteria and the methods used in BWR reload analysis to evaluate the impact of channel bow on CPR margins. Potential weaknesses in today's methodologies are discussed. Westinghouse in collaboration with KKL and Axpo - operator and owner of the Leibstadt NPP - has developed an optimized CPR methodology based on a new criterion to protect against dryout during normal operation and with a more rigorous treatment of channel bow. The new steady-state criterion is expressed in terms of an upper limit of 0.01 for the dryout failure probability per year. This ismore » considered a meaningful and appropriate criterion that can be directly related to the probabilistic criteria set-up for the analyses of Anticipated Operation Occurrences (AOOs) and accidents. In the Monte Carlo approach a statistical modeling of channel bow and an accurate evaluation of CPR response functions allow the associated CPR penalties to be included directly in the plant SLMCPR and OLMCPR in a best-estimate manner. In this way, the treatment of channel bow is equivalent to all other uncertainties affecting CPR. Emphasis is put on quantifying the statistical distribution of channel bow throughout the core using measurement data. The optimized CPR methodology has been implemented in the Westinghouse Monte Carlo code, McSLAP. The methodology improves the quality of dryout safety assessments by supplying more valuable information and better control of conservatisms in establishing operational limits for CPR. The methodology is demonstrated with application examples from the introduction at KKL. (authors)« less

Fatigue crack growth in SA508-CL2 steel in a high temperature, high purity water environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerber, T.L.; Heald, J.D.; Kiss, E.

1974-10-01

Fatigue crack growth tests were conducted with 1 in. plate specimens of SA508-CL 2 steel in room temperature air, 550$sup 0$F air and in a 550$sup 0$F, high purity, water environment. Zero-tension load controlled tests were run at cyclic frequencies as low as 0.037 CPM. Results show that growth rates in the simulated Boiling Water Reactor (BWR) water environment are faster than growth rates observed in 550$sup 0$F air and these rates are faster than the room temperature rate. In the BWR water environment, lowering the cyclic frequency from 0.37 to 0.037 CPM caused only a slight increase in themore » fatigue crack growth rate. All growth rates measured in these tests were below the upper bound design curve presented in Section XI of the ASME Code. (auth)« less

Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors

NASA Astrophysics Data System (ADS)

Gordon, Barry; Garcia, Susan

Boiling water reactors (BWRs) operate with very high purity water. However, even the utilization of near theoretical conductivity water cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel, wrought nickel alloys and nickel weld metals under oxygenated conditions. IGSCC can be further accelerated by the presence of certain impurities dissolved in the coolant. The goal of this paper is to present the technical basis for controlling various impurities under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions for mitigation of IGSCC. More specifically, the effects of typical BWR ionic impurities (e.g., sulfate, chloride, nitrate, borate, phosphate, etc.) on IGSCC propensities in both NWC and HWC environments will be discussed. The technical basis for zinc addition to the BWR coolant will also provided along with an in-plant example of the most severe water chemistry transient to date.

Analysis of loss of decay-heat-removal sequences at Browns Ferry Unit One

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrington, R.M.

1983-01-01

This paper summarizes the Oak Ridge National Laboratory (ORNL) report Loss of DHR Sequences at Browns Ferry Unit One - Accident Sequence Analysis (NUREG/CR-2973). The Loss of DHR investigation is the third in a series of accident studies concerning the BWR 4 - MK I containment plant design. These studies, sponsored by the Nuclear Regulatory Commission Severe Accident Sequence Analysis (SASA) program, have been conducted at ORNL with the full cooperation of the Tennessee Valley Authority (TVA). The purpose of the SASA studies is to predetermine the probable course of postulated severe accidents so as to establish the timing andmore » the sequence of events. The SASA studies also produce recommendations concerning the implementation of better system design and better emergency operating instructions and operator training. The ORNL studies also include a detailed, best-estimate calculation of the release and transport of radioactive fission products following postulated severe accidents.« less

ALARA efforts in nordic BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ingemansson, T.; Lundgren, K.; Elkert, J.

1995-03-01

Some ALARA-related ABB Atom projects are currently under investigation. One of the projects has been ordered by the Swedish Radiation Protection Institute, and two others by the Nordic BWR utilities. The ultimate objective of the projects is to identify and develop methods to significantly decrease the future exposure levels in the Nordic BWRS. As 85% to 90% of the gamma radiation field in the Nordic BWRs originates from Co-60, the only way to significantly decrease the radiation doses is to effect Co and Co-60. The strategy to do this is to map the Co sources and estimate the source strengthmore » of Co from these sources, and to study the possibility to affect the release of Co-60 from the core surfaces and the uptake on system surfaces. Preliminary results indicate that corrosion/erosion of a relatively small number of Stellite-coated valves and/or dust from grinding of Stellite valves may significantly contribute to the Co input to the reactors. This can be seen from a high measured Co/Ni ratio in the feedwater and in the reactor water. If stainless steel is the only source of Co, the Co/Ni ratio would be less than 0.02 as the Co content in the steel is less than 0.2%. The Co/Ni ratio in the reactor water, however, is higher than 0.1, indicating that the major fraction of the Co originates from Stellite-coated valves. There are also other possible explanations for an increase of the radiation fields. The Co-60 inventory on the core surfaces increases approximately as the square of the burn-up level. If the burn-up is increased from 35 to 5 MWd/kgU, the Co-60 inventory on the core surfaces will be doubled. Also the effect on the behavior of Co-60 of different water chemistry and materials conditions is being investigated. Examples of areas studied are Fe and Zn injection, pH-control, and different forms of surface pre-treatments.« less

Impulsivity-based thrifty eating phenotype and the protective role of n-3 PUFAs intake in adolescents.

PubMed

Reis, R S; Dalle Molle, R; Machado, T D; Mucellini, A B; Rodrigues, D M; Bortoluzzi, A; Bigonha, S M; Toazza, R; Salum, G A; Minuzzi, L; Buchweitz, A; Franco, A R; Pelúzio, M C G; Manfro, G G; Silveira, P P

2016-03-15

The goal of the present study was to investigate whether intrauterine growth restriction (IUGR) affects brain responses to palatable foods and whether docosahexaenoic acid (DHA, an omega-3 fatty acid that is a primary structural component of the human brain) serum levels moderate the association between IUGR and brain and behavioral responses to palatable foods. Brain responses to palatable foods were investigated using a functional magnetic resonance imaging task in which participants were shown palatable foods, neutral foods and non-food items. Serum DHA was quantified in blood samples, and birth weight ratio (BWR) was used as a proxy for IUGR. The Dutch Eating Behavior Questionnaire (DEBQ) was used to evaluate eating behaviors. In the contrast palatable food > neutral items, we found an activation in the right superior frontal gyrus with BWR as the most important predictor; the lower the BWR (indicative of IUGR), the greater the activation of this region involved in impulse control/decision making facing the viewing of palatable food pictures versus neutral items. At the behavioral level, a general linear model predicting external eating using the DEBQ showed a significant interaction between DHA and IUGR status; in IUGR individuals, the higher the serum DHA, the lower is external eating. In conclusion, we suggest that IUGR moderates brain responses when facing stimuli related to palatable foods, activating an area related to impulse control. Moreover, higher intake of n-3 PUFAs can protect IUGR individuals from developing inappropriate eating behaviors, the putative mechanism of protection would involve decreasing intake in response to external food cues in adolescents/young adults.

Impulsivity-based thrifty eating phenotype and the protective role of n-3 PUFAs intake in adolescents

PubMed Central

Reis, R S; Dalle Molle, R; Machado, T D; Mucellini, A B; Rodrigues, D M; Bortoluzzi, A; Bigonha, S M; Toazza, R; Salum, G A; Minuzzi, L; Buchweitz, A; Franco, A R; Pelúzio, M C G; Manfro, G G; Silveira, P P

2016-01-01

The goal of the present study was to investigate whether intrauterine growth restriction (IUGR) affects brain responses to palatable foods and whether docosahexaenoic acid (DHA, an omega-3 fatty acid that is a primary structural component of the human brain) serum levels moderate the association between IUGR and brain and behavioral responses to palatable foods. Brain responses to palatable foods were investigated using a functional magnetic resonance imaging task in which participants were shown palatable foods, neutral foods and non-food items. Serum DHA was quantified in blood samples, and birth weight ratio (BWR) was used as a proxy for IUGR. The Dutch Eating Behavior Questionnaire (DEBQ) was used to evaluate eating behaviors. In the contrast palatable food > neutral items, we found an activation in the right superior frontal gyrus with BWR as the most important predictor; the lower the BWR (indicative of IUGR), the greater the activation of this region involved in impulse control/decision making facing the viewing of palatable food pictures versus neutral items. At the behavioral level, a general linear model predicting external eating using the DEBQ showed a significant interaction between DHA and IUGR status; in IUGR individuals, the higher the serum DHA, the lower is external eating. In conclusion, we suggest that IUGR moderates brain responses when facing stimuli related to palatable foods, activating an area related to impulse control. Moreover, higher intake of n-3 PUFAs can protect IUGR individuals from developing inappropriate eating behaviors, the putative mechanism of protection would involve decreasing intake in response to external food cues in adolescents/young adults. PMID:26978737

Reactor shutdown experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cletcher, J.W.

1995-10-01

This is a regular report of summary statistics relating to recent reactor shutdown experience. The information includes both number of events and rates of occurence. It was compiled from data about operating events that were entered into the SCSS data system by the Nuclear Operations Analysis Center at the Oak ridge National Laboratory and covers the six mont period of July 1 to December 31, 1994. Cumulative information, starting from May 1, 1994, is also reported. Updates on shutdown events included in earlier reports is excluded. Information on shutdowns as a function of reactor power at the time of themore » shutdown for both BWR and PWR reactors is given. Data is also discerned by shutdown type and reactor age.« less

Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, R.E.

Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments.

Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hadgu, Teklu; Hardin, Ernest; Matteo, Edward N.

2015-12-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are keptmore » open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

2016-01-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement formore » extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

75 FR 21046 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards In accordance with the... on Reactor Safeguards (ACRS) will hold a meeting on May 6-8, 2010, 11545 Rockville Pike, Rockville....: Boiling Water Reactor (BWR) Owners Group (BWROG) Topical Report NEDC-33347P, ``Containment Overpressure...

ALARA Council: Sharing our resources and experiences to reduce doses at Commonwealth Edison Facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rescek, F.

1995-03-01

Commonwealth Edison Company is an investor-owned utility company supplying electricity to over three million customers (eight million people) in Chicago and northern Illinois, USA. The company operates 16 generating stations which have the capacity to produce 22,522 megawatts of electricity. Six of these generating stations, containing 12 nuclear units, supply 51% of this capacity. The 12 nuclear units are comprised of four General Electric boiling water (BWR-3) reactors, two General Electric BWR-5 reactors, and six Westinghouse four-loop pressurized water reactors (PWR). In August 1993, Commonwealth Edison created an ALARA Council with the responsibility to provide leadership and guidance that resultsmore » in an effective ALARA Culture within the Nuclear Operations Division. Unlike its predecessor, the Corporate ALARA Committee, the ALARA Council is designed to bring together senior managers from the six nuclear stations and corporate to create a collaborative effort to reduce occupational doses at Commonwealth Edison`s stations.« less

The startup of the Dodewaard natural circulation boiling water reactor -- Experiences

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nissen, W.H.M.; Van Der Voet, J.; Karuza, J.

1994-07-01

Because of its similarity to the simplified boiling water reactor (SBWR), the Dodewaard natural circulation boiling water reactor (BWR) is of special interest to further development of the SBWR design. It has become especially important to gain more insight into the Dodewaard BWR behavior during startup, paying special attention to its stability. Therefore, special instrumentation was used by means of which a series of measurements were taken during the two startups in February and June 1992. The results obtained from these measurements are used to deepen insight into the recirculation flow and the stability of the reactor during startup undermore » conditions with a normal pressure/power trajectory. They have already shown a very early recirculation flow onset during low-power operation and no indication of reactor instability. Furthermore, they will be used as a basis for the research program investigating the reactor behavior under different pressure/power conditions, which is scheduled for next year.« less

Removal of Sb-125 and Tc-99 from Liquid Radwaste by Novel Adsorbents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harjula, R.O.; Koivula, R.; Paajanen, A.

2006-07-01

Novel proprietary metal oxide materials (MOM) have been tested for the removal of Sb-125 from simulated Floor Drain Waters of BWR. Antimony was present in the solutions as oxidized anionic form. Long term column experiment with simulated liquid that showed high Sb-125 removal at least up to 8000 bed volumes. One column experiments was carried out using nonradioactive Sb to exhaust the column. Leaching tests with 1000 ppm boric acid showed that 100 % of absorbed Sb remains in the sorbent material. Column experiments with real Fuel Pond Water from Olkiluoto NPP (BWR) showed reduction of Sb-125 (feed level 400more » Bq/L, 1.10{sup -5} {mu}Ci/mL) below detection limit (MDA = 1.7 Bq/L, 5.10{sup -8},{mu}Ci/mL). Additional experiments have also been carried out with pertechnetate (Tc-99) ions. Results indicate that MOM materials are efficient also for the removal of Tc-99 from concentrated NaNO{sub 3} solution. (authors)« less

Emergency cooling system and method

DOEpatents

Oosterkamp, W.J.; Cheung, Y.K.

1994-01-04

An improved emergency cooling system and method are disclosed that may be adapted for incorporation into or use with a nuclear BWR wherein a reactor pressure vessel (RPV) containing a nuclear core and a heat transfer fluid for circulation in a heat transfer relationship with the core is housed within an annular sealed drywell and is fluid communicable therewith for passage thereto in an emergency situation the heat transfer fluid in a gaseous phase and any noncondensibles present in the RPV, an annular sealed wetwell houses the drywell, and a pressure suppression pool of liquid is disposed in the wetwell and is connected to the drywell by submerged vents. The improved emergency cooling system and method has a containment condenser for receiving condensible heat transfer fluid in a gaseous phase and noncondensibles for condensing at least a portion of the heat transfer fluid. The containment condenser has an inlet in fluid communication with the drywell for receiving heat transfer fluid and noncondensibles, a first outlet in fluid communication with the RPV for the return to the RPV of the condensed portion of the heat transfer fluid and a second outlet in fluid communication with the drywell for passage of the noncondensed balance of the heat transfer fluid and the noncondensibles. The noncondensed balance of the heat transfer fluid and the noncondensibles passed to the drywell from the containment condenser are mixed with the heat transfer fluid and the noncondensibles from the RPV for passage into the containment condenser. A water pool is provided in heat transfer relationship with the containment condenser and is thermally communicable in an emergency situation with an environment outside of the drywell and the wetwell for conducting heat transferred from the containment condenser away from the wetwell and the drywell. 5 figs.

Posttest Analyses of the Steel Containment Vessel Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Costello, J.F.; Hessheimer, M.F.; Ludwigsen, J.S.

A high pressure test of a scale model of a steel containment vessel (SCV) was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. This testis part of a program to investigate the response of representative models of nuclear containment structures to pressure loads beyond the design basis accident. The posttest analyses of this test focused on three areas where the pretest analysis effort did not adequately predict the model behavior duringmore » the test. These areas are the onset of global yielding, the strain concentrations around the equipment hatch and the strain concentrations that led to a small tear near a weld relief opening that was not modeled in the pretest analysis.« less

Effects of experimental floods on riparian and aquatic ecosystems: Bill Williams River, Arizona

NASA Astrophysics Data System (ADS)

Shafroth, P. B.; Andersen, D. C.; Wilcox, A. C.; Kui, L.; Stella, J. C.

2013-12-01

Development of flow prescriptions for environmental purposes along rivers is relatively common, but implementation of these 'environmental flows' occurs infrequently. Implementation is critical for testing hypotheses relating flow regime to biotic response, which ultimately can inform adaptive flow management. We describe the development of flow prescriptions and evaluate responses of riparian vegetation, beaver dams, and associated aquatic habitat to experimental floods and intervening base flows associated with an environmental flow program on the Bill Williams River (BWR), in semiarid Arizona. First, we assessed effects of flow releases between 1993 and 2009 designed to favor the establishment and maintenance of native riparian trees (Populus and Salix) and disfavor an invasive, nonnative shrub (Tamarix spp.) downstream of Alamo Dam on the BWR. Our data are multi-scaled and include a several-decade assessment of changes to major vegetation types based on a time series of aerial photography, an assessment of species composition and abundance sampled in permanent vegetation quadrats, and targeted seedling surveys following experimental floods. Between 1993 and 2009, we observed significant increases in Populus and Salix forests and essentially no change in Tamarix. Experimental floods in 2006 and 2007 resulted in higher mortality of Tamarix seedlings than Salix. These results illustrate the potential for managing streamflow to influence riparian vegetation dynamics, including management of nonnative species. Second, we examined the role of beaver as ecosystem engineers in the BWR and linkages to flow releases between 2004 and 2013. Beaver convert lotic stream habitat to lentic through dam construction and maintenance during low flow periods, and the process is reversed when a flood or other event causes dam failure. We estimated the extent of lotic and beaver-created lentic (beaver pond) habitat along the BWR and related the likelihood of damage or destruction of beaver dams to the magnitude and duration of experimental floods. We obtained counts of beaver dams at various times from aerial photographs, aerial videography, and ground surveys. The ratio of lotic to lentic stream length was approximately 6 times greater following a large flood versus a 7 year period with no significant flood releases. Floods of different magnitudes and durations resulted in notably different levels of damage or removal of beaver dams. Finally, we sampled woody vegetation adjacent to the channel to estimate the effect of beaver herbivory, and noted high levels of mature tree mortality in one of our study reaches. Results of our previous and ongoing investigations are reported to land and water managers as part of an adaptive streamflow management process.

Level-2 IPE for the Laguna Verde NPS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arellano, J.; De Loera, M.A.; Rea, R.

1996-12-31

In response to generic letter GL 88-20, Comision Federal de Electricidad and Instituto de Investigaciones Electricas have jointly developed the individual plant examination (IPE) for the Laguna Verde nuclear power station unit I (LVNPS). This plant is a 675-MW(electric) boiling water reactor (BWR/5) with a reinforced concrete Mark-II containment. The approach used to fulfill the IPE requirements was to make a level-1 probabilistic risk assessment (IPE level 1) plus a containment performance analysis including the behavior and release of the fission products to the environment (IPE level 2). This paper describes the level-2 portion of the LVNPS IPE, paying specialmore » attention to both some improvements to the traditional analytical methods and to the main results.« less

Two-phase pressure drop reduction BWR assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dix, G.E.; Crowther, R.L.; Colby, M.J.

1991-05-21

This patent describes an improved fuel assembly for a boiling water reactor. It comprises: a fuel channel; a lower tie plate; an upper tie plate; the lower tie plate and the upper tie plate defining a two-dimensional matrix; at least one water rod the fuel rods being partial length rods.

Recent developments in chemical decontamination technology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, C.J.

1995-03-01

Chemical decontamination of parts of reactor coolant systems is a mature technology, used routinely in many BWR plants, but less frequently in PWRs. This paper reviews recent developments in the technology - corrosion minimization, waste processing and full system decontamination, including the fuel. Earlier work was described in an extensive review published in 1990.

Comparison of Models for Spacer Grid Pressure Loss in Nuclear Fuel Bundles for One and Two-Phase Flows

NASA Astrophysics Data System (ADS)

Maskal, Alan B.

Spacer grids maintain the structural integrity of the fuel rods within fuel bundles of nuclear power plants. They can also improve flow characteristics within the nuclear reactor core. However, spacer grids add reactor coolant pressure losses, which require estimation and engineering into the design. Several mathematical models and computer codes were developed over decades to predict spacer grid pressure loss. Most models use generalized characteristics, measured by older, less precise equipment. The study of OECD/US-NRC BWR Full-Size Fine Mesh Bundle Tests (BFBT) provides updated and detailed experimental single and two-phase results, using technically advanced flow measurements for a wide range of boundary conditions. This thesis compares the predictions from the mathematical models to the BFBT experimental data by utilizing statistical formulae for accuracy and precision. This thesis also analyzes the effects of BFBT flow characteristics on spacer grids. No single model has been identified as valid for all flow conditions. However, some models' predictions perform better than others within a range of flow conditions, based on the accuracy and precision of the models' predictions. This study also demonstrates that pressure and flow quality have a significant effect on two-phase flow spacer grid models' biases.

Development of a fully-consistent reduced order model to study instabilities in boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dykin, V.; Demaziere, C.

2012-07-01

A simple nonlinear Reduced Order Model to study global, regional and local instabilities in Boiling Water Reactors is described. The ROM consists of three submodels: neutron-kinetic, thermal-hydraulic and heat-transfer models. The neutron-kinetic model allows representing the time evolution of the three first neutron kinetic modes: the fundamental, the first and the second azimuthal modes. The thermal-hydraulic model describes four heated channels in order to correctly simulate out-of-phase behavior. The coupling between the different submodels is performed via both void and Doppler feedback mechanisms. After proper spatial homogenization, the governing equations are discretized in the time-domain. Several modifications, compared to othermore » existing ROMs, have been implemented, and are reported in this paper. One novelty of the ROM is the inclusion of both azimuthal modes, which allows to study combined instabilities (in-phase and out-of-phase), as well as to investigate the corresponding interference effects between them. The second modification concerns the precise estimation of so-called reactivity coefficients or C{sub mn}{sup *V,D} - coefficients by using direct cross-section data from SIMULATE-3 combined with the CORE SIM core simulator in order to calculate Eigenmodes. Furthermore, a non-uniform two-step axial power profile is introduced to simulate the separate heat production in the single and two-phase regions, respectively. An iterative procedure was developed to calculate the solution to the coupled neutron-kinetic/thermal-hydraulic static problem prior to solving the time-dependent problem. Besides, the possibility of taking into account the effect of local instabilities is demonstrated in a simplified manner. The present ROM is applied to the investigation of an actual instability that occurred at the Swedish Forsmark-1 BWR in 1996/1997. The results generated by the ROM are compared with real power plant measurements performed during stability tests and show a good qualitative agreement. The present study provides some insight in a deeper understanding of the physical principles which drive both core-wide and local instabilities. (authors)« less

77 FR 69507 - Proposed Model Safety Evaluation for Plant-Specific Adoption of Technical Specifications Task...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-19

..., ``Revise Shutdown Margin Definition To Address Advanced Fuel Designs'' AGENCY: Nuclear Regulatory... Shutdown Margin Definition to Address Advanced Fuel Designs.'' DATES: Comment period expires on December 19... address newer BWR fuel designs, which may be more reactive at shutdown temperatures above 68[emsp14][deg]F...

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).

On the Solidification and Structure Formation during Casting of Large Inserts in Ferritic Nodular Cast Iron

NASA Astrophysics Data System (ADS)

Tadesse, Abel; Fredriksson, Hasse

2018-06-01

The graphite nodule count and size distributions for boiling water reactor (BWR) and pressurized water reactor (PWR) inserts were investigated by taking samples at heights of 2160 and 1150 mm, respectively. In each cross section, two locations were taken into consideration for both the microstructural and solidification modeling. The numerical solidification modeling was performed in a two-dimensional model by considering the nucleation and growth in eutectic ductile cast iron. The microstructural results reveal that the nodule size and count distribution along the cross sections are different in each location for both inserts. Finer graphite nodules appear in the thinner sections and close to the mold walls. The coarser nodules are distributed mostly in the last solidified location. The simulation result indicates that the finer nodules are related to a higher cooling rate and a lower degree of microsegregation, whereas the coarser nodules are related to a lower cooling rate and a higher degree of microsegregation. The solidification time interval and the last solidifying locations in the BWR and PWR are also different.

Investigation of natural circulation instability and transients in passively safe novel modular reactor

NASA Astrophysics Data System (ADS)

Shi, Shanbin

The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal-hydraulic and nuclear coupled startup transients are performed to investigate the flow instabilities at low pressure and low power conditions. Two different power ramps are chosen to study the effect of power density on the flow instability. The experimental startup transient tests show the existence of three different flow instability mechanisms during the low pressure startup transients, i.e., flashing instability, condensation induced instability, and density wave oscillations. Flashing instability in the chimney section of the test loop and density wave oscillation are the main flow instabilities observed when the system pressure is below 0.5 MPa. They show completely different type of oscillations, i.e., intermittent oscillation and sinusoidal oscillation, in void fraction profile during the startup transients. In order to perform nuclear-coupled startup transients with void reactivity feedback, the Point Kinetics model is utilized to calculate the transient power during the startup transients. In addition, the differences between the electric resistance heaters and typical fuel element are taken into account. The reactor power calculated shows some oscillations due to flashing instability during the transients. However, the void reactivity feedback does not have significant influence on the flow instability during the startup procedure for the NMR-50. Further investigation of very small power ramp on the startup transients is carried out for the thermal-hydraulic startup transients. It is found that very small power density can eliminate the flashing oscillation in the single phase natural circulation and stabilize the flow oscillations in the phase of net vapor generation. Furthermore, initially pressurized startup procedure is investigated to eliminate the main flow instabilities. The results show that the pressurized startup procedure can suppress the flashing instability at low pressure and low power conditions. In order to have a deep understanding of natural circulation flow instability, the quasi-steady tests are performed using the test facility installed with preheater and subcooler. The effects of system pressure, core inlet subcooling, core power density, inlet flow resistance coefficient, and void reactivity feedback are investigated in the quasi-steady state tests. The stability boundaries are determined between unstable and stable flow conditions in the dimensionless stability plane of inlet subcooling number and Zuber number. In order to predict the stability boundary theoretically, linear stability analysis in the frequency domain is performed at four sections of the loop. The flashing in the chimney is considered as an axially uniform heat source. The dimensionless characteristic equation of the pressure drop perturbation is obtained by considering the void fraction effect and outlet flow resistance in the chimney section. The flashing boundary shows some discrepancies with previous experimental data from the quasi-steady state tests. In the future, thermal non-equilibrium is recommended to improve the accuracy of flashing instability boundary.

Aging Management Guideline for commercial nuclear power plants: Motor control centers; Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toman, G.; Gazdzinski, R.; O`Hearn, E.

1994-02-01

This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specificmore » aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.« less

Scalable Methods for Uncertainty Quantification, Data Assimilation and Target Accuracy Assessment for Multi-Physics Advanced Simulation of Light Water Reactors

NASA Astrophysics Data System (ADS)

Khuwaileh, Bassam

High fidelity simulation of nuclear reactors entails large scale applications characterized with high dimensionality and tremendous complexity where various physics models are integrated in the form of coupled models (e.g. neutronic with thermal-hydraulic feedback). Each of the coupled modules represents a high fidelity formulation of the first principles governing the physics of interest. Therefore, new developments in high fidelity multi-physics simulation and the corresponding sensitivity/uncertainty quantification analysis are paramount to the development and competitiveness of reactors achieved through enhanced understanding of the design and safety margins. Accordingly, this dissertation introduces efficient and scalable algorithms for performing efficient Uncertainty Quantification (UQ), Data Assimilation (DA) and Target Accuracy Assessment (TAA) for large scale, multi-physics reactor design and safety problems. This dissertation builds upon previous efforts for adaptive core simulation and reduced order modeling algorithms and extends these efforts towards coupled multi-physics models with feedback. The core idea is to recast the reactor physics analysis in terms of reduced order models. This can be achieved via identifying the important/influential degrees of freedom (DoF) via the subspace analysis, such that the required analysis can be recast by considering the important DoF only. In this dissertation, efficient algorithms for lower dimensional subspace construction have been developed for single physics and multi-physics applications with feedback. Then the reduced subspace is used to solve realistic, large scale forward (UQ) and inverse problems (DA and TAA). Once the elite set of DoF is determined, the uncertainty/sensitivity/target accuracy assessment and data assimilation analysis can be performed accurately and efficiently for large scale, high dimensional multi-physics nuclear engineering applications. Hence, in this work a Karhunen-Loeve (KL) based algorithm previously developed to quantify the uncertainty for single physics models is extended for large scale multi-physics coupled problems with feedback effect. Moreover, a non-linear surrogate based UQ approach is developed, used and compared to performance of the KL approach and brute force Monte Carlo (MC) approach. On the other hand, an efficient Data Assimilation (DA) algorithm is developed to assess information about model's parameters: nuclear data cross-sections and thermal-hydraulics parameters. Two improvements are introduced in order to perform DA on the high dimensional problems. First, a goal-oriented surrogate model can be used to replace the original models in the depletion sequence (MPACT -- COBRA-TF - ORIGEN). Second, approximating the complex and high dimensional solution space with a lower dimensional subspace makes the sampling process necessary for DA possible for high dimensional problems. Moreover, safety analysis and design optimization depend on the accurate prediction of various reactor attributes. Predictions can be enhanced by reducing the uncertainty associated with the attributes of interest. Accordingly, an inverse problem can be defined and solved to assess the contributions from sources of uncertainty; and experimental effort can be subsequently directed to further improve the uncertainty associated with these sources. In this dissertation a subspace-based gradient-free and nonlinear algorithm for inverse uncertainty quantification namely the Target Accuracy Assessment (TAA) has been developed and tested. The ideas proposed in this dissertation were first validated using lattice physics applications simulated using SCALE6.1 package (Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lattice models). Ultimately, the algorithms proposed her were applied to perform UQ and DA for assembly level (CASL progression problem number 6) and core wide problems representing Watts Bar Nuclear 1 (WBN1) for cycle 1 of depletion (CASL Progression Problem Number 9) modeled via simulated using VERA-CS which consists of several multi-physics coupled models. The analysis and algorithms developed in this dissertation were encoded and implemented in a newly developed tool kit algorithms for Reduced Order Modeling based Uncertainty/Sensitivity Estimator (ROMUSE).

Effects of Thermal Aging on Material Properties, Stress Corrosion Cracking, and Fracture Toughness of AISI 316L Weld Metal

NASA Astrophysics Data System (ADS)

Lucas, Timothy; Forsström, Antti; Saukkonen, Tapio; Ballinger, Ronald; Hänninen, Hannu

2016-08-01

Thermal aging and consequent embrittlement of materials are ongoing issues in cast stainless steels, as well as duplex, and high-Cr ferritic stainless steels. Spinodal decomposition is largely responsible for the well-known "748 K (475 °C) embrittlement" that results in drastic reductions in ductility and toughness in these materials. This process is also operative in welds of either cast or wrought stainless steels where δ-ferrite is present. While the embrittlement can occur after several hundred hours of aging at 748 K (475 °C), the process is also operative at lower temperatures, at the 561 K (288 °C) operating temperature of a boiling water reactor (BWR), for example, where ductility reductions have been observed after several tens of thousands of hours of exposure. An experimental program was carried out in order to understand how spinodal decomposition may affect changes in material properties in Type 316L BWR piping weld metals. The study included material characterization, nanoindentation hardness, double-loop electrochemical potentiokinetic reactivation (DL-EPR), Charpy-V, tensile, SCC crack growth, and in situ fracture toughness testing as a function of δ-ferrite content, aging time, and temperature. SCC crack growth rates of Type 316L stainless steel weld metal under simulated BWR conditions showed an approximate 2 times increase in crack growth rate over that of the unaged as-welded material. In situ fracture toughness measurements indicate that environmental exposure can result in a reduction of toughness by up to 40 pct over the corresponding at-temperature air-tested values. Material characterization results suggest that spinodal decomposition is responsible for the degradation of material properties measured in air, and that degradation of the in situ properties may be a result of hydrogen absorbed during exposure to the high-temperature water environment.

Neutron Collar Evolution and Fresh PWR Assembly Measurements with a New Fast Neutron Passive Collar

DOE Office of Scientific and Technical Information (OSTI.GOV)

Menlove, Howard Olsen; Geist, William H.; Root, Margaret A.

The passive neutron collar approach removes the effect of poison rods when using a 1mm Gd liner. This project sets out to solve the following challenges: BWR fuel assemblies have less mass and less neutron multiplication than PWR; and effective removal of cosmic ray spallation neutron bursts needed via QC tests.

Summary of BISON Development and Validation Activities - NEAMS FY16 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, R. L.; Pastore, G.; Gamble, K. A.

This summary report contains an overview of work performed under the work package en- titled “FY2016 NEAMS INL-Engineering Scale Fuel Performance (BISON)” A first chapter identifies the specific FY-16 milestones, providing a basic description of the associated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes major additional accomplishments, which in- clude: 1) publication of a journal article on solution verification and validation of BISON for LWR fuel, 2) publication of a journal article on 3D Missing Pellet Surface (MPS) analysis of BWR fuel, 3) use of BISON to designmore » a unique 3D MPS validation experiment for future in- stallation in the Halden research reactor, 4) participation in an OECD benchmark on Pellet Clad Mechanical Interaction (PCMI), 5) participation in an OECD benchmark on Reactivity Insertion Accident (RIA) analysis, 6) participation in an OECD activity on uncertainity quantification and sensitivity analysis in nuclear fuel modeling and 7) major improvements to BISON’s fission gas behavior models. A final chapter outlines FY-17 future work.« less

BWR station blackout: A RISMC analysis using RAVEN and RELAP5-3D

DOE PAGES

Mandelli, D.; Smith, C.; Riley, T.; ...

2016-01-01

The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates and improved operations. In order to evaluate the impact of these factors on the safety of the plant, the Risk-Informed Safety Margin Characterization (RISMC) project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions and accident scenarios. This paper presents a case study in order to show the capabilities of the RISMC methodology to assess impact of power uprate of a Boiling Watermore » Reactor system during a Station Black-Out accident scenario. We employ a system simulator code, RELAP5-3D, coupled with RAVEN which perform the stochastic analysis. Furthermore, our analysis is performed by: 1) sampling values from a set of parameters from the uncertainty space of interest, 2) simulating the system behavior for that specific set of parameter values and 3) analyzing the outcomes from the set of simulation runs.« less

Efficiency of static core turn-off in a system-on-a-chip with variation

DOEpatents

Cher, Chen-Yong; Coteus, Paul W; Gara, Alan; Kursun, Eren; Paulsen, David P; Schuelke, Brian A; Sheets, II, John E; Tian, Shurong

2013-10-29

A processor-implemented method for improving efficiency of a static core turn-off in a multi-core processor with variation, the method comprising: conducting via a simulation a turn-off analysis of the multi-core processor at the multi-core processor's design stage, wherein the turn-off analysis of the multi-core processor at the multi-core processor's design stage includes a first output corresponding to a first multi-core processor core to turn off; conducting a turn-off analysis of the multi-core processor at the multi-core processor's testing stage, wherein the turn-off analysis of the multi-core processor at the multi-core processor's testing stage includes a second output corresponding to a second multi-core processor core to turn off; comparing the first output and the second output to determine if the first output is referring to the same core to turn off as the second output; outputting a third output corresponding to the first multi-core processor core if the first output and the second output are both referring to the same core to turn off.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Masuda, Y.; Chiba, N.; Matsuo, Y.

This research proposes to investigate the impact behavior of the steel plate of BWR containment vessels against missiles, caused by the postulated catastrophic failure of components with a high kinetic energy. Although the probability of the occurrence of missiles inside and outside of containment vessels is extremely low, the following items are required to maintain the integrity of containment vessels: the probability of the occurrence of missiles, the weight and energy of missiles, and the impact behavior of containment vessel steel plate against postulated missiles. In connection with the third item, an actualscale missile test was conducted. In addition, amore » computation analysis was performed to confirm the impact behavior against the missiles, in order to search for wide applicability to the various kinds of postulated missiles. This research tries to derive a new empirical formula which carries out the assessment of the integrity of containment vessels.« less

KERENA safety concept in the context of the Fukushima accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zacharias, T.; Novotny, C.; Bielor, E.

Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basicmore » physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)« less

Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2008-01-21

In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heat-affected-zone materials, that were irradiated to fluence levels as high as {approx} 2x 10{sup 21} n/cm{sup 2} (E > 1more » MeV) ({approx} 3 dpa) in a light water reactor at 288-300 C. The results are compared with the data available in the literature. The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined. A superposition model was used to represent the cyclic CGRs of austenitic SSs. The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that bounds the existing data has been defined. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated.« less

Silicon carbide composite for light water reactor fuel assembly applications

NASA Astrophysics Data System (ADS)

Yueh, Ken; Terrani, Kurt A.

2014-05-01

The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

Development of new UV-I. I. Cerenkov Viewing Device

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuribara, Masayuki; Nemoto, Koshichi

1994-02-01

The Cerenkov glow images from boiling-water reactors (BWR) and pressurized-water reactors (PWR) irradiated fuel assemblies are generally used for inspections. However, sometimes it is difficult or impossible to identify the image by the conventional Cerenkov Viewing Device (CVD), because of the long cooling time and/or low burnup. Now a new UV-I.I. (Ultra-Violet light Image Intensifier) CVD has been developed, which can detect the very weak Cerenkov glow from spent fuel assemblies. As this new device uses the newly developed proximity focused type UV-I.I., Cerenkov photons are used efficiently, producing better quality Cerenkov glow images. Moreover, since the image is convertedmore » to a video signal, it is easy to improve the signal to noise ratio (S/N) by an image processor. The new CVD was tested at BWR and PWR power plants in Japan, with fuel burnups ranging from 6,200--33,000 MWD/MTU (megawatt days per metric ton of uranium) and cooling times ranging from 370 to 6,200 d. The tests showed that the new CVD is superior to the conventional STA/CRIEPI CVD, and could detect very feeble Cerenkov glow images using an image processor.« less

Annual progress report on the NSRR experiments, (21)

NASA Astrophysics Data System (ADS)

1992-05-01

Fuel behavior studies under simulated reactivity-initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor (NSRR) since 1975. This report gives the results of experiments performed from April, 1989 through March, 1990 and discussions of them. A total of 41 tests were carried out during this period. The tests are distinguished into pre-irradiated fuel tests and fresh fuel tests; the former includes 2 JMTR pre-irradiated fuel tests, 2 PWR pre-irradiated fuel tests, and 2 BWR pre-irradiated fuel tests, and the latter includes 6 standard fuel tests (6 SP(center dot)CP scoping tests), 7 power and cooling condition parameter tests (4 flow shrouded fuel tests, 1 bundle simulation test, 1 fully water-filled vessel test, 1 high pressure/high temperature loop test), 12 special fuel tests (3 stainless steel clad fuel tests, 3 improved PWR fuel tests, 6 improved BWR fuel tests), 3 severe fuel damage tests (1 high temperature flooding test, 1 flooding behavior observation test, 1 debris coolability test), 3 fast breeder reactor fuel tests (2 moderator material characteristic measurement tests, 1 fuel behavior observation test), and 2 miscellaneous tests (2 preliminary tests for pre-irradiated fuel tests).

Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Douglas W. Akers; Edwin A. Harvego

2012-08-01

This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and datamore » on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.« less

Comparison of new generation low-complexity flood inundation mapping tools with a hydrodynamic model

NASA Astrophysics Data System (ADS)

Afshari, Shahab; Tavakoly, Ahmad A.; Rajib, Mohammad Adnan; Zheng, Xing; Follum, Michael L.; Omranian, Ehsan; Fekete, Balázs M.

2018-01-01

The objective of this study is to compare two new generation low-complexity tools, AutoRoute and Height Above the Nearest Drainage (HAND), with a two-dimensional hydrodynamic model (Hydrologic Engineering Center-River Analysis System, HEC-RAS 2D). The assessment was conducted on two hydrologically different and geographically distant test-cases in the United States, including the 16,900 km2 Cedar River (CR) watershed in Iowa and a 62 km2 domain along the Black Warrior River (BWR) in Alabama. For BWR, twelve different configurations were set up for each of the models, including four different terrain setups (e.g. with and without channel bathymetry and a levee), and three flooding conditions representing moderate to extreme hazards at 10-, 100-, and 500-year return periods. For the CR watershed, models were compared with a simplistic terrain setup (without bathymetry and any form of hydraulic controls) and one flooding condition (100-year return period). Input streamflow forcing data representing these hypothetical events were constructed by applying a new fusion approach on National Water Model outputs. Simulated inundation extent and depth from AutoRoute, HAND, and HEC-RAS 2D were compared with one another and with the corresponding FEMA reference estimates. Irrespective of the configurations, the low-complexity models were able to produce inundation extents similar to HEC-RAS 2D, with AutoRoute showing slightly higher accuracy than the HAND model. Among four terrain setups, the one including both levee and channel bathymetry showed lowest fitness score on the spatial agreement of inundation extent, due to the weak physical representation of low-complexity models compared to a hydrodynamic model. For inundation depth, the low-complexity models showed an overestimating tendency, especially in the deeper segments of the channel. Based on such reasonably good prediction skills, low-complexity flood models can be considered as a suitable alternative for fast predictions in large-scale hyper-resolution operational frameworks, without completely overriding hydrodynamic models' efficacy.

10 CFR 50.75 - Reporting and recordkeeping for decommissioning planning.

Code of Federal Regulations, 2010 CFR

2010-01-01

... investing or otherwise, that a prudent investor would use in the same circumstances. The term “prudent... than or equal to 3400 MWt $105 between 1200 MWt and 3400 MWt (For a PWR of less than 1200 MWt, use P... 3400 MWt (For a BWR of less than 1200 MWt, use P=1200 MWt) $(104+0.009P) (2) An adjustment factor at...

A Lattice Boltzmann Framework for the simulation of boiling hydrodynamics in BWRs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jain, P. K.; Tentner, A.; Uddin, R.

2008-01-01

Multi phase and multi component flows are ubiquitous in nature as well as in many man-made processes. A specific example is the Boiling Water Reactor (BWR) core, in which the coolant enters the core as liquid, undergoes a phase change as it traverses the core and exits as a high quality two-phase mixture. Two-phase flows in BWRs typically manifest a wide variety of geometrical patterns of the co-existing phases depending on the local system conditions. Modeling of such flows currently relies on empirical correlations (for example, in the simulation of bubble nucleation, bubble growth and coalescence, and inter-phase surface topologymore » transitions) that hinder the accurate simulation of two-phase phenomena using Computational Fluid Dynamics (CFD) approaches. The Lattice Boltzmann Method (LBM) is in rapid development as a modeling tool to understand these macro-phenomena by coupling them with their underlying micro-dynamics. This paper presents a consistent LBM formulation for the simulation of a two-phase water-steam system. Results of initial model validation in a range of thermodynamic conditions typical for BWRs are also shown. The interface between the two coexisting phases is captured from the dynamics of the model itself, i.e., no interface tracking is needed. The model is based on the Peng-Robinson (P-R) non-ideal equation of state and can quantitatively approximate the phase-coexistence curve for water at different temperatures ranging from 125 to 325 oC. Consequently, coexisting phases with large density ratios (up to {approx}1000) may be simulated. Two-phase models in the 200-300 C temperature range are of significant importance to nuclear engineers since most BWRs operate under similar thermodynamic conditions. Simulation of bubbles and droplets in a gravity-free environment of the corresponding coexisting phase until steady state is reached satisfies Laplace law at different temperatures and thus, yield the surface tension of the fluid. Comparing the LBM surface tension thus calculated using the LBM to the corresponding experimental values for water, the LBM lattice unit (lu) can be scaled to the physical units. Using this approach, spatial scaling of the LBM emerges from the model itself and is not imposed externally.« less

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

JPRS Report Science & Technology Japan

DTIC Science & Technology

1989-10-25

Testing Slated for New BWR Fuel Assemblies [GENSHIRYOKU SANGYO SHIMBUN, 25 May 89] .... 37 Nuclear Fuel Planning System Developed [GENSHIRYOKU... Development (Debt) 13,272 ((Debt) 3,839) 7,995 (3,610) In addition, the budget has guaranteed that the following programs will proceed according... develop a combined cycle engine that will be capable of attaining high reliability and good fuel consumption at a wide range of speeds from low speed to

TRIGA MARK-II source term

DOE Office of Scientific and Technical Information (OSTI.GOV)

Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Hamzah, N. S., E-mail: mark-dennis@nuclearmalaysia.gov.my; Abi, M. J. B., E-mail: mark-dennis@nuclearmalaysia.gov.my

ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences ofmore » results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.« less

Nuclear fuel performance: Trends, remedies and challenges

NASA Astrophysics Data System (ADS)

Rusch, C. A.

2008-12-01

It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. 'Fuel reliability' has a much broader definition than just maintaining mechanical integrity and being leaker free - fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems.

Development of a reliable estimation procedure of radioactivity inventory in a BWR plant due to neutron irradiation for decommissioning

NASA Astrophysics Data System (ADS)

Tanaka, Ken-ichi; Ueno, Jun

2017-09-01

Reliable information of radioactivity inventory resulted from the radiological characterization is important in order to plan decommissioning planning and is also crucial in order to promote decommissioning in effectiveness and in safe. The information is referred to by planning of decommissioning strategy and by an application to regulator. Reliable information of radioactivity inventory can be used to optimize the decommissioning processes. In order to perform the radiological characterization reliably, we improved a procedure of an evaluation of neutron-activated materials for a Boiling Water Reactor (BWR). Neutron-activated materials are calculated with calculation codes and their validity should be verified with measurements. The evaluation of neutron-activated materials can be divided into two processes. One is a distribution calculation of neutron-flux. Another is an activation calculation of materials. The distribution calculation of neutron-flux is performed with neutron transport calculation codes with appropriate cross section library to simulate neutron transport phenomena well. Using the distribution of neutron-flux, we perform distribution calculations of radioactivity concentration. We also estimate a time dependent distribution of radioactivity classification and a radioactive-waste classification. The information obtained from the evaluation is utilized by other tasks in the preparatory tasks to make the decommissioning plan and the activity safe and rational.

Probabilistic pipe fracture evaluations for leak-rate-detection applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahman, S.; Ghadiali, N.; Paul, D.

1995-04-01

Regulatory Guide 1.45, {open_quotes}Reactor Coolant Pressure Boundary Leakage Detection Systems,{close_quotes} was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, {open_quotes}Leak Before Break Evaluation Procedures{close_quotes} where a margin of 10 on the leak detection limit is used in determining the crackmore » size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

McElroy, W.N.

1985-08-01

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

Multi-Core Processor Memory Contention Benchmark Analysis Case Study

NASA Technical Reports Server (NTRS)

Simon, Tyler; McGalliard, James

2009-01-01

Multi-core processors dominate current mainframe, server, and high performance computing (HPC) systems. This paper provides synthetic kernel and natural benchmark results from an HPC system at the NASA Goddard Space Flight Center that illustrate the performance impacts of multi-core (dual- and quad-core) vs. single core processor systems. Analysis of processor design, application source code, and synthetic and natural test results all indicate that multi-core processors can suffer from significant memory subsystem contention compared to similar single-core processors.

Estimation of carbon 14 inventory in hull and end-piece wastes from Japanese commercial reprocessing operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tomofumi Sakuragi; Hiromi Tanabe; Emiko Hirose

2013-07-01

Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters. Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the {sup 14}N(n,p){sup 14}C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated bymore » ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts. In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types x 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the calculated activation. The total C-14 inventory was estimated as 4.46x10{sup 14} Bq, consisting of 2.58x10{sup 14} Bq for BWRs and 1.87x10{sup 14} Bq for PWRs, and is consistent with the safety assessment of 4.4x10{sup 14} Bq. However, the distribution of the C-14 inventory to hull oxide, which was estimated under the assumption of instantaneous radionuclide release in the safety assessment, decreased from 5.72x10{sup 13} Bq (13% of the total) in the previous assessment to 1.30x10{sup 13} Bq (2.9% of the total; consisting of 1.48x10{sup 12} for BWRs and 1.15x10{sup 13} for PWRs). In other words, the exposure dose peak is reduced to approximate 25% of its previous value due to the use of detailed oxide film data that the BWR cladding has a thin oxide film. Other instantaneous release components for C-14 such as the fuel residual were negligible. (authors)« less

Publications - GMC 418 | Alaska Division of Geological & Geophysical

Science.gov Websites

DGGS GMC 418 Publication Details Title: Porosity, permeability, grain density core analysis (CT scans , permeability, grain density core analysis (CT scans), and core photos from the ConocoPhillips N. Cook Inlet

Development and Application of Laser Peening System for PWR Power Plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Masaki Yoda; Itaru Chida; Satoshi Okada

2006-07-01

Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion crackingmore » (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it's too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J-groove welds after laser ultrasonic testing (LUT) as the remote inspection, and we peened the outer surface and the weld for Ikata Unit 2 supplementary. For core deluge line nozzles and primary water inlet nozzles, we peened the inner surface of the dissimilar metal welding, which is of nickel base alloy, joining a safe end and a low alloy metal nozzle. In this paper, the development and the actual application of the laser peening system for PWR power plants will be described. (authors)« less

The influence of geometric imperfections on the stability of three-layer beams with foam core

NASA Astrophysics Data System (ADS)

Wstawska, Iwona

2017-01-01

The main objective of this work is the numerical analysis (FE analysis) of stability of three-layer beams with metal foam core (alumina foam core). The beams were subjected to pure bending. The analysis of the local buckling was performed. Furthermore, the influence of geometric parameters of the beam and material properties of the core (linear and non-linear model) on critical loads values and buckling shape were also investigated. The calculations were made on a family of beams with different mechanical properties of the core (elastic and elastic-plastic material). In addition, the influence of geometric imperfections on deflection and normal stress values of the core and the faces has been evaluated.

A Finite Element Analysis for Predicting the Residual Compressive Strength of Impact-Damaged Sandwich Panels

NASA Technical Reports Server (NTRS)

Ratcliffe, James G.; Jackson, Wade C.

2008-01-01

A simple analysis method has been developed for predicting the residual compressive strength of impact-damaged sandwich panels. The method is tailored for honeycomb core-based sandwich specimens that exhibit an indentation growth failure mode under axial compressive loading, which is driven largely by the crushing behavior of the core material. The analysis method is in the form of a finite element model, where the impact-damaged facesheet is represented using shell elements and the core material is represented using spring elements, aligned in the thickness direction of the core. The nonlinear crush response of the core material used in the analysis is based on data from flatwise compression tests. A comparison with a previous analysis method and some experimental data shows good agreement with results from this new approach.

A Finite Element Analysis for Predicting the Residual Compression Strength of Impact-Damaged Sandwich Panels

NASA Technical Reports Server (NTRS)

Ratcliffe, James G.; Jackson, Wade C.

2008-01-01

A simple analysis method has been developed for predicting the residual compression strength of impact-damaged sandwich panels. The method is tailored for honeycomb core-based sandwich specimens that exhibit an indentation growth failure mode under axial compression loading, which is driven largely by the crushing behavior of the core material. The analysis method is in the form of a finite element model, where the impact-damaged facesheet is represented using shell elements and the core material is represented using spring elements, aligned in the thickness direction of the core. The nonlinear crush response of the core material used in the analysis is based on data from flatwise compression tests. A comparison with a previous analysis method and some experimental data shows good agreement with results from this new approach.

Disposition of feedwater nozzle UT indications in a BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leshnoff, S.D.; Orski, M.A.

A technical logic is developed, which justifies the disposition of feedwater nozzle ultrasonic testing (UT) indications in order to return to operation without visual inspection of the vessel inside surface. Present regulatory guidance is to inspect the inside surface from the inside if a reportable indication is found. A highly sensitive, tomographic UT technique, developed by Kraftwerk Union, is used to detect and size machined notches in the blend radius and bore regions of a full-sized feedwater nozzle mock-up.

Application of the TEMPEST computer code for simulating hydrogen distribution in model containment structures. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trent, D.S.; Eyler, L.L.

In this study several aspects of simulating hydrogen distribution in geometric configurations relevant to reactor containment structures were investigated using the TEMPEST computer code. Of particular interest was the performance of the TEMPEST turbulence model in a density-stratified environment. Computed results illustrated that the TEMPEST numerical procedures predicted the measured phenomena with good accuracy under a variety of conditions and that the turbulence model used is a viable approach in complex turbulent flow simulation.

Evaluation of Analysis Techniques for Fluted-Core Sandwich Cylinders

NASA Technical Reports Server (NTRS)

Lovejoy, Andrew E.; Schultz, Marc R.

2012-01-01

Buckling-critical launch-vehicle structures require structural concepts that have high bending stiffness and low mass. Fluted-core, also known as truss-core, sandwich construction is one such concept. In an effort to identify an analysis method appropriate for the preliminary design of fluted-core cylinders, the current paper presents and compares results from several analysis techniques applied to a specific composite fluted-core test article. The analysis techniques are evaluated in terms of their ease of use and for their appropriateness at certain stages throughout a design analysis cycle (DAC). Current analysis techniques that provide accurate determination of the global buckling load are not readily applicable early in the DAC, such as during preliminary design, because they are too costly to run. An analytical approach that neglects transverse-shear deformation is easily applied during preliminary design, but the lack of transverse-shear deformation results in global buckling load predictions that are significantly higher than those from more detailed analysis methods. The current state of the art is either too complex to be applied for preliminary design, or is incapable of the accuracy required to determine global buckling loads for fluted-core cylinders. Therefore, it is necessary to develop an analytical method for calculating global buckling loads of fluted-core cylinders that includes transverse-shear deformations, and that can be easily incorporated in preliminary design.

s-core network decomposition: A generalization of k-core analysis to weighted networks

NASA Astrophysics Data System (ADS)

Eidsaa, Marius; Almaas, Eivind

2013-12-01

A broad range of systems spanning biology, technology, and social phenomena may be represented and analyzed as complex networks. Recent studies of such networks using k-core decomposition have uncovered groups of nodes that play important roles. Here, we present s-core analysis, a generalization of k-core (or k-shell) analysis to complex networks where the links have different strengths or weights. We demonstrate the s-core decomposition approach on two random networks (ER and configuration model with scale-free degree distribution) where the link weights are (i) random, (ii) correlated, and (iii) anticorrelated with the node degrees. Finally, we apply the s-core decomposition approach to the protein-interaction network of the yeast Saccharomyces cerevisiae in the context of two gene-expression experiments: oxidative stress in response to cumene hydroperoxide (CHP), and fermentation stress response (FSR). We find that the innermost s-cores are (i) different from innermost k-cores, (ii) different for the two stress conditions CHP and FSR, and (iii) enriched with proteins whose biological functions give insight into how yeast manages these specific stresses.

Single Event Effects (SEE) Testing of Embedded DSP Cores within Microsemi RTAX4000D Field Programmable Gate Array (FPGA) Devices

NASA Technical Reports Server (NTRS)

Perez, Christopher E.; Berg, Melanie D.; Friendlich, Mark R.

2011-01-01

Motivation for this work is: (1) Accurately characterize digital signal processor (DSP) core single-event effect (SEE) behavior (2) Test DSP cores across a large frequency range and across various input conditions (3) Isolate SEE analysis to DSP cores alone (4) Interpret SEE analysis in terms of single-event upsets (SEUs) and single-event transients (SETs) (5) Provide flight missions with accurate estimate of DSP core error rates and error signatures.

An improved method for field extraction and laboratory analysis of large, intact soil cores

USGS Publications Warehouse

Tindall, J.A.; Hemmen, K.; Dowd, J.F.

1992-01-01

Various methods have been proposed for the extraction of large, undisturbed soil cores and for subsequent analysis of fluid movement within the cores. The major problems associated with these methods are expense, cumbersome field extraction, and inadequate simulation of unsaturated flow conditions. A field and laboratory procedure is presented that is economical, convenient, and simulates unsaturated and saturated flow without interface flow problems and can be used on a variety of soil types. In the field, a stainless steel core barrel is hydraulically pressed into the soil (30-cm diam. and 38 cm high), the barrel and core are extracted from the soil, and after the barrel is removed from the core, the core is then wrapped securely with flexible sheet metal and a stainless mesh screen is attached to the bottom of the core for support. In the laboratory the soil core is set atop a porous ceramic plate over which a soil-diatomaceous earth slurry has been poured to assure good contact between plate and core. A cardboard cylinder (mold) is fastened around the core and the empty space filled with paraffin wax. Soil cores were tested under saturated and unsaturated conditions using a hanging water column for potentials ???0. Breakthrough curves indicated that no interface flow occurred along the edge of the core. This procedure proved to be reliable for field extraction of large, intact soil cores and for laboratory analysis of solute transport.

Recovery and Lithologic Analysis of Sediment from Hole UT-GOM2-1-H002, Green Canyon 955, Northern Gulf of Mexico

NASA Astrophysics Data System (ADS)

Kinash, N.; Cook, A.; Sawyer, D.; Heber, R.

2017-12-01

In May 2017 the University of Texas led a drilling and pressure coring expedition in the northern Gulf of Mexico, UT-GOM2-01. The holes were located in Green Canyon Block 955, where the Gulf of Mexico Joint Industry Project Leg II identified an approximately 100m thick hydrate-filled course-grained levee unit in 2009. Two separate wells were drilled into this unit: Holes H002 and H005. In Hole H002, a cutting shoe drill bit was used to collect the pressure cores, and only 1 of the 8 cores collected was pressurized during recovery. The core recovery in Hole H002 was generally poor, about 34%, while the only pressurized core had 45% recovery. In Hole H005, a face bit was used during pressure coring where 13 cores were collected and 9 cores remained pressurized. Core recovery in Hole H005 was much higher, at about 75%. The type of bit was not the only difference between the holes, however. Drilling mud was used throughout the drilling and pressure coring of Hole H002, while only seawater was used during the first 80m of pressure cores collected in Hole H005. Herein we focus on lithologic analysis of Hole H002 with the goal of documenting and understanding core recovery in Hole H002 to compare with Hole H005. X-ray Computed Tomography (XCT) images were collected by Geotek on pressurized cores, mostly from Hole H005, and at Ohio State on unpressurized cores, mostly from Hole H002. The XCT images of unpressurized cores show minimal sedimentary structures and layering, unlike the XCT images acquired on the pressurized, hydrate-bearing cores. Only small sections of the unpressurized cores remained intact. The unpressurized cores appear to have two prominent facies: 1) silt that did not retain original sedimentary fabric and often was loose within the core barrel, and 2) dense mud sections with some sedimentary structures and layering present. On the XCT images, drilling mud appears to be concentrated on the sides of cores, but also appears in layers and fractures within intact core sections. On microscope images, the drilling mud also appears to saturate the pores in some silt intervals. Further analysis of the unpressurized cores is planned, including X-ray diffraction, grain size analysis and porosity measurements. These results will be compared to the pressurized cores to understand if further lithologic factors could have affected core recovery.

Improvement of Speckle Contrast Image Processing by an Efficient Algorithm.

PubMed

Steimers, A; Farnung, W; Kohl-Bareis, M

2016-01-01

We demonstrate an efficient algorithm for the temporal and spatial based calculation of speckle contrast for the imaging of blood flow by laser speckle contrast analysis (LASCA). It reduces the numerical complexity of necessary calculations, facilitates a multi-core and many-core implementation of the speckle analysis and enables an independence of temporal or spatial resolution and SNR. The new algorithm was evaluated for both spatial and temporal based analysis of speckle patterns with different image sizes and amounts of recruited pixels as sequential, multi-core and many-core code.

User’s guide to the collection and analysis of tree cores to assess the distribution of subsurface volatile organic compounds

USGS Publications Warehouse

Vroblesky, Don A.

2008-01-01

Analysis of the volatile organic compound content of tree cores is an inexpensive, rapid, simple approach to examining the distribution of subsurface volatile organic compound contaminants. The method has been shown to detect several volatile petroleum hydrocarbons and chlorinated aliphatic compounds associated with vapor intrusion and ground-water contamination. Tree cores, which are approximately 3 inches long, are obtained by using an increment borer. The cores are placed in vials and sealed. After a period of equilibration, the cores can be analyzed by headspace analysis gas chromatography. Because the roots are exposed to volatile organic compound contamination in the unsaturated zone or shallow ground water, the volatile organic compound concentrations in the tree cores are an indication of the presence of subsurface volatile organic compound contamination. Thus, tree coring can be used to detect and map subsurface volatile organic compound contamination. For comparison of tree-core data at a particular site, it is important to maintain consistent methods for all aspects of tree-core collection, handling, and analysis. Factors affecting the volatile organic compound concentrations in tree cores include the type of volatile organic compound, the tree species, the rooting depth, ground-water chemistry, the depth to the contaminated horizon, concentration differences around the trunk related to variations in the distribution of subsurface volatile organic compounds, concentration differences with depth of coring related to volatilization loss through the bark and possibly other unknown factors, dilution by rain, seasonal influences, sorption, vapor-exchange rates, and within-tree volatile organic compound degradation.

Vascularity as assessed by Doppler intraoral ultrasound around the invasion front of tongue cancer is a predictor of pathological grade of malignancy and cervical lymph node metastasis.

PubMed

Yamamoto, Chika; Yuasa, Kenji; Okamura, Kazuhiko; Shiraishi, Tomoko; Miwa, Kunihiro

2016-01-01

To quantitatively evaluate the relationship of vascularity of tongue cancer as demonstrated on intraoral ultrasonography images and tumour thickness with pathological grade of malignancy and the presence of cervical lymph node metastases. 18 patients with tongue cancer were enrolled in this retrospective study. Using Doppler ultrasonography images of the invasion front of the cancers along the length of their tumour boundaries, three vascular indexes were analysed quantitatively, namely ratio of blood flow signal area within the cancer to whole tumour area (BAR), blood flow signal number ratio (BNR) and blood flow signal width ratio (BWR). The associations between these three indexes and occurrence of cervical lymph node metastasis and pathological grade of malignancy [Yamamoto-Kohama (YK) classification] were assessed. Furthermore, the relationship between tumour thickness and occurrence of cervical lymph node metastasis was evaluated on B-mode intraoral ultrasonography images. There was no significant association between BAR and tumour thickness or occurrence of cervical lymph node metastasis. The BNRs and BWRs of patients with cervical lymph node metastasis were significantly higher than those of patients without nodal involvement. The BWRs of patients with high-grade malignancy (YK-4C) were significantly higher than those of patients with low-grade malignancy (YK-2 or 3). BNR and BWR on the invasion front of the tongue cancer are predictors of pathological grade of malignancy and cervical lymph node metastasis.

Measurement of liver function using hepatobiliary scintigraphy improves risk assessment in patients undergoing major liver resection.

PubMed

Cieslak, Kasia P; Bennink, Roelof J; de Graaf, Wilmar; van Lienden, Krijn P; Besselink, Marc G; Busch, Olivier R C; Gouma, Dirk J; van Gulik, Thomas M

2016-09-01

(99m)Tc-mebrofenin-hepatobiliary-scintigraphy (HBS) enables measurement of future remnant liver (FRL)-function and was implemented in our preoperative routine after calculation of the cut-off value for prediction of postoperative liver failure (LF). This study evaluates our results since the implementation of HBS. Additionally, CT-volumetric methods of FRL-assessment, standardized liver volumetry and FRL/body-weight ratio (FRL-BWR), were evaluated. 163 patients who underwent major liver resection were included. Insufficient FRL-volume and/or FRL-function <2.7%/min/m(2) were indications for portal vein embolization (PVE). Non-PVE patients were compared with a historical cohort (n = 55). Primary endpoints were postoperative LF and LF related mortality. Secondary endpoint was preoperative identification of patients at risk for LF using the CT-volumetric methods. 29/163 patients underwent PVE; 8/29 patients because of insufficient FRL-function despite sufficient FRL-volume. According to FRL-BWR and standardized liver volumetry, 16/29 and 11/29 patients, respectively, would not have undergone PVE. LF and LF related mortality were significantly reduced compared to the historical cohort. HBS appeared superior in the identification of patients with increased surgical risk compared to the CT-volumetric methods. Implementation of HBS in the preoperative work-up led to a function oriented use of PVE and was associated with a significant decrease in postoperative LF and LF related mortality. Copyright © 2016 International Hepato-Pancreato-Biliary Association Inc. Published by Elsevier Ltd. All rights reserved.

Characterization Data Package for Containerized Sludge Samples Collected from Engineered Container SCS-CON-210

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fountain, Matthew S.; Fiskum, Sandra K.; Baldwin, David L.

This data package contains the K Basin sludge characterization results obtained by Pacific Northwest National Laboratory during processing and analysis of four sludge core samples collected from Engineered Container SCS-CON-210 in 2010 as requested by CH2M Hill Plateau Remediation Company. Sample processing requirements, analytes of interest, detection limits, and quality control sample requirements are defined in the KBC-33786, Rev. 2. The core processing scope included reconstitution of a sludge core sample distributed among four to six 4-L polypropylene bottles into a single container. The reconstituted core sample was then mixed and subsampled to support a variety of characterization activities. Additionalmore » core sludge subsamples were combined to prepare a container composite. The container composite was fractionated by wet sieving through a 2,000 micron mesh and a 500-micron mesh sieve. Each sieve fraction was sampled to support a suite of analyses. The core composite analysis scope included density determination, radioisotope analysis, and metals analysis, including the Waste Isolation Pilot Plant Hazardous Waste Facility Permit metals (with the exception of mercury). The container composite analysis included most of the core composite analysis scope plus particle size distribution, particle density, rheology, and crystalline phase identification. A summary of the received samples, core sample reconstitution and subsampling activities, container composite preparation and subsampling activities, physical properties, and analytical results are presented. Supporting data and documentation are provided in the appendices. There were no cases of sample or data loss and all of the available samples and data are reported as required by the Quality Assurance Project Plan/Sampling and Analysis Plan.« less

A Full Virial Analysis of the Prestellar Cores in the Ophiuchus Molecular Cloud

NASA Astrophysics Data System (ADS)

Pattle, Kate; Ward-Thompson, Derek

We use SCUBA-2, HARP C18O J= 3 -> 2, Herschel and IRAM N2H+ J= 1 -> 0 observations of the Ophiuchus molecular cloud to identify and characterise the properties of the starless cores in the region. The SCUBA-2, HARP and Herschel data were taken as part of the JCMT and Herschel Gould Belt Surveys. We determine masses and temperatures and perform a full virial analysis on our cores, and find that our cores are all either bound or virialised, with gravitational energy and external pressure energy on average of similar importance in confining the cores. There is wide variation from region to region, with cores in the region influenced by B stars (Oph A) being substantially gravitationally bound, and cores in the most quiescent region (Oph C) being pressure-confined. We observe dissipation of turbulence in all our cores, and find that this dissipation is more effective in regions which do not contain outflow-driving protostars. Full details of this analysis are presented by Pattle et al. (2015).

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

NASA Technical Reports Server (NTRS)

Butler, C.; Albright, D.

2007-01-01

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

Development and Testing of Neutron Cross Section Covariance Data for SCALE 6.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, William BJ J; Williams, Mark L; Wiarda, Dorothea

2015-01-01

Neutron cross-section covariance data are essential for many sensitivity/uncertainty and uncertainty quantification assessments performed both within the TSUNAMI suite and more broadly throughout the SCALE code system. The release of ENDF/B-VII.1 included a more complete set of neutron cross-section covariance data: these data form the basis for a new cross-section covariance library to be released in SCALE 6.2. A range of testing is conducted to investigate the properties of these covariance data and ensure that the data are reasonable. These tests include examination of the uncertainty in critical experiment benchmark model k eff values due to nuclear data uncertainties, asmore » well as similarity assessments of irradiated pressurized water reactor (PWR) and boiling water reactor (BWR) fuel with suites of critical experiments. The contents of the new covariance library, the testing performed, and the behavior of the new covariance data are described in this paper. The neutron cross-section covariances can be combined with a sensitivity data file generated using the TSUNAMI suite of codes within SCALE to determine the uncertainty in system k eff caused by nuclear data uncertainties. The Verified, Archived Library of Inputs and Data (VALID) maintained at Oak Ridge National Laboratory (ORNL) contains over 400 critical experiment benchmark models, and sensitivity data are generated for each of these models. The nuclear data uncertainty in k eff is generated for each experiment, and the resulting uncertainties are tabulated and compared to the differences in measured and calculated results. The magnitude of the uncertainty for categories of nuclides (such as actinides, fission products, and structural materials) is calculated for irradiated PWR and BWR fuel to quantify the effect of covariance library changes between the SCALE 6.1 and 6.2 libraries. One of the primary applications of sensitivity/uncertainty methods within SCALE is the assessment of similarities between benchmark experiments and safety applications. This is described by a c k value for each experiment with each application. Several studies have analyzed typical c k values for a range of critical experiments compared with hypothetical irradiated fuel applications. The c k value is sensitive to the cross-section covariance data because the contribution of each nuclide is influenced by its uncertainty; large uncertainties indicate more likely bias sources and are thus given more weight. Changes in c k values resulting from different covariance data can be used to examine and assess underlying data changes. These comparisons are performed for PWR and BWR fuel in storage and transportation systems.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE PAGES

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; ...

2016-09-07

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Szilard, Ronaldo Henriques

A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

Voltage-spike analysis for a free-running parallel inverter

NASA Technical Reports Server (NTRS)

Lee, F. C. Y.; Wilson, T. G.

1974-01-01

Unwanted and sometimes damaging high-amplitude voltage spikes occur during each half cycle in many transistor saturable-core inverters at the moment when the core saturates and the transistors switch. The analysis shows that spikes are an intrinsic characteristic of certain types of inverters even with negligible leakage inductance and purely resistive load. The small but unavoidable after-saturation inductance of the saturable-core transformer plays an essential role in creating these undesired thigh-voltage spikes. State-plane analysis provides insight into the complex interaction between core and transistors, and shows the circuit parameters upon which the magnitude of these spikes depends.

Particle-in-cell simulation study on halo formation in anisotropic beams

NASA Astrophysics Data System (ADS)

Ikegami, Masanori

2000-11-01

In a recent paper (M. Ikegami, Nucl. Instr. and Meth. A 435 (1999) 284), we investigated halo formation processes in transversely anisotropic beams based on the particle-core model. The effect of simultaneous excitation of two normal modes of core oscillation, i.e., high- and low-frequency modes, was examined. In the present study, self-consistent particle simulations are performed to confirm the results obtained in the particle-core analysis. In these simulations, it is confirmed that the particle-core analysis can predict the halo extent accurately even in anisotropic situations. Furthermore, we find that the halo intensity is enhanced in some cases where two normal modes of core oscillation are simultaneously excited as expected in the particle-core analysis. This result is of practical importance because pure high-frequency mode oscillation has frequently been assumed in preceding halo studies. The dependence of halo intensity on the 2:1 fixed point locations is also discussed.

Damage Tolerance of Sandwich Plates With Debonded Face Sheets

NASA Technical Reports Server (NTRS)

Sankar, Bhavani V.

2001-01-01

A nonlinear finite element analysis was performed to simulate axial compression of sandwich beams with debonded face sheets. The load - end-shortening diagrams were generated for a variety of specimens used in a previous experimental study. The energy release rate at the crack tip was computed using the J-integral, and plotted as a function of the load. A detailed stress analysis was performed and the critical stresses in the face sheet and the core were computed. The core was also modeled as an isotropic elastic-perfectly plastic material and a nonlinear post buckling analysis was performed. A Graeco-Latin factorial plan was used to study the effects of debond length, face sheet and core thicknesses, and core density on the load carrying capacity of the sandwich composite. It has been found that a linear buckling analysis is inadequate in determining the maximum load a debonded sandwich beam can carry. A nonlinear post-buckling analysis combined with an elastoplastic model of the core is required to predict the compression behavior of debonded sandwich beams.

Quantifying the Impact of Nanoparticle Coatings and Non-uniformities on XPS Analysis: Gold/silver Core-shell Nanoparticles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Yung-Chen Andrew; Engelhard, Mark H.; Baer, Donald R.

2016-03-07

Abstract or short description: Spectral modeling of photoelectrons can serve as a valuable tool when combined with X-ray photoelectron spectroscopy (XPS) analysis. Herein, a new version of the NIST Simulation of Electron Spectra for Surface Analysis (SESSA 2.0) software, capable of directly simulating spherical multilayer NPs, was applied to model citrate stabilized Au/Ag-core/shell nanoparticles (NPs). The NPs were characterized using XPS and scanning transmission electron microscopy (STEM) to determine the composition and morphology of the NPs. The Au/Ag-core/shell NPs were observed to be polydispersed in size, non-circular, and contain off-centered Au-cores. Using the average NP dimensions determined from STEM analysis,more » SESSA spectral modeling indicated that washed Au/Ag-core shell NPs were stabilized with a 0.8 nm l« less

A pressure core ultrasonic test system for on-board analysis of gas hydrate-bearing sediments under in situ pressures.

PubMed

Yang, Lei; Zhou, Weihua; Xue, Kaihua; Wei, Rupeng; Ling, Zheng

2018-05-01

The enormous potential as an alternative energy resource has made natural gas hydrates a material of intense research interest. Their exploration and sample characterization require a quick and effective analysis of the hydrate-bearing cores recovered under in situ pressures. Here a novel Pressure Core Ultrasonic Test System (PCUTS) for on-board analysis of sediment cores containing gas hydrates at in situ pressures is presented. The PCUTS is designed to be compatible with an on-board pressure core transfer device and a long gravity-piston pressure-retained corer. It provides several advantages over laboratory core analysis including quick and non-destructive detection, in situ and successive acoustic property acquisition, and remission of sample storage and transportation. The design of the unique assembly units to ensure the in situ detection is demonstrated, involving the U-type protecting jackets, transducer precession device, and pressure stabilization system. The in situ P-wave velocity measurements make the detection of gas hydrate existence in the sediments possible on-board. Performance tests have verified the feasibility and sensitivity of the ultrasonic test unit, showing the dependence of P-wave velocity on gas hydrate saturation. The PCUTS has been successfully applied for analysis of natural samples containing gas hydrates recovered from the South China Sea. It is indicated that on-board P-wave measurements could provide a quick and effective understanding of the hydrate occurrence in natural samples, which can assist further resource exploration, assessment, and subsequent detailed core analysis.

A pressure core ultrasonic test system for on-board analysis of gas hydrate-bearing sediments under in situ pressures

NASA Astrophysics Data System (ADS)

Yang, Lei; Zhou, Weihua; Xue, Kaihua; Wei, Rupeng; Ling, Zheng

2018-05-01

The enormous potential as an alternative energy resource has made natural gas hydrates a material of intense research interest. Their exploration and sample characterization require a quick and effective analysis of the hydrate-bearing cores recovered under in situ pressures. Here a novel Pressure Core Ultrasonic Test System (PCUTS) for on-board analysis of sediment cores containing gas hydrates at in situ pressures is presented. The PCUTS is designed to be compatible with an on-board pressure core transfer device and a long gravity-piston pressure-retained corer. It provides several advantages over laboratory core analysis including quick and non-destructive detection, in situ and successive acoustic property acquisition, and remission of sample storage and transportation. The design of the unique assembly units to ensure the in situ detection is demonstrated, involving the U-type protecting jackets, transducer precession device, and pressure stabilization system. The in situ P-wave velocity measurements make the detection of gas hydrate existence in the sediments possible on-board. Performance tests have verified the feasibility and sensitivity of the ultrasonic test unit, showing the dependence of P-wave velocity on gas hydrate saturation. The PCUTS has been successfully applied for analysis of natural samples containing gas hydrates recovered from the South China Sea. It is indicated that on-board P-wave measurements could provide a quick and effective understanding of the hydrate occurrence in natural samples, which can assist further resource exploration, assessment, and subsequent detailed core analysis.

The low-power low-pressure flow resonance in a natural circulation cooled boiling water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hagen, T.H.J.J. van der; Stekelenburg, A.J.C.

1995-09-01

The last few years the possibility of flow resonances during the start-up phase of natural circulation cooled BWRs has been put forward by several authors. The present paper reports on actual oscillations observed at the Dodewaard reactor, the world`s only operating BWR cooled by natural circulation. In addition, results of a parameter study performed by means of a simple theoretical model are presented. The influence of relevant parameters on the resonance characteristics, being the decay ratio and the resonance frequency, is investigated and explained.

Seismic margin assessment of the Edwin I. Hatch Nuclear Plant, Unit 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barr, W.T.; Moore, D.P.; Smith, J.E.

1991-06-01

This summary presents the results and lessons learned from the seismic margin assessment (SMA) of Unit 1 of the Hatch Nuclear Plant. The primary purpose of this SMA was to assess the practicality of the EPRI SMA methodology on a BWR on a soil site such as Hatch. The major findings from the Hatch SMA are briefly described along with the lessons learned during the project implementation. The experience gained on the Hatch SMA is expected to benefit others in the performance of future SMAs. 12 refs.

Pool-site fuel inspection and examination techniques applied by the Kraftwerk Union Aktiengesellschaft Fuel Service. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knaab, H.; Knecht, K.

The need for pool-site inspection and examination of fuel assemblies was recognized by Kraftwerk Union Aktiengesellschaft with the commissioning of the first nuclear power stations. A wet sipping method has demonstrated high reliability in detection of leaking fuel assemblies. The visual inspection system is a versatile tool. It can be supplemented by attaching devices for oxide thickness measurement or surface replication. Repair of leaking pressurized water reactor fuel assemblies has improved fuel utilization. Applied methods and typical results are described.

3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langenbuch, S.; Velkov, K.; Lizorkin, M.

1997-07-01

This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

Tank 241-B-108, cores 172 and 173 analytical results for the final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nuzum, J.L., Fluoro Daniel Hanford

1997-03-04

The Data Summary Table (Table 3) included in this report compiles analytical results in compliance with all applicable DQOS. Liquid subsamples that were prepared for analysis by an acid adjustment of the direct subsample are indicated by a `D` in the A column in Table 3. Solid subsamples that were prepared for analysis by performing a fusion digest are indicated by an `F` in the A column in Table 3. Solid subsamples that were prepared for analysis by performing a water digest are indicated by a I.wl. or an `I` in the A column of Table 3. Due to poormore » precision and accuracy in original analysis of both Lower Half Segment 2 of Core 173 and the core composite of Core 173, fusion and water digests were performed for a second time. Precision and accuracy improved with the repreparation of Core 173 Composite. Analyses with the repreparation of Lower Half Segment 2 of Core 173 did not show improvement and suggest sample heterogeneity. Results from both preparations are included in Table 3.« less

Tank 241-AP-105, cores 208, 209 and 210, analytical results for the final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nuzum, J.L.

1997-10-24

This document is the final laboratory report for Tank 241-AP-105. Push mode core segments were removed from Risers 24 and 28 between July 2, 1997, and July 14, 1997. Segments were received and extruded at 222-S Laboratory. Analyses were performed in accordance with Tank 241-AP-105 Push Mode Core Sampling and Analysis Plan (TSAP) (Hu, 1997) and Tank Safety Screening Data Quality Objective (DQO) (Dukelow, et al., 1995). None of the subsamples submitted for total alpha activity (AT), differential scanning calorimetry (DSC) analysis, or total organic carbon (TOC) analysis exceeded the notification limits as stated in TSAP and DQO. The statisticalmore » results of the 95% confidence interval on the mean calculations are provided by the Tank Waste Remediation Systems Technical Basis Group, and are not considered in this report. Appearance and Sample Handling Two cores, each consisting of four segments, were expected from Tank 241-AP-105. Three cores were sampled, and complete cores were not obtained. TSAP states core samples should be transported to the laboratory within three calendar days from the time each segment is removed from the tank. This requirement was not met for all cores. Attachment 1 illustrates subsamples generated in the laboratory for analysis and identifies their sources. This reference also relates tank farm identification numbers to their corresponding 222-S Laboratory sample numbers.« less

Integration of ANFIS, NN and GA to determine core porosity and permeability from conventional well log data

NASA Astrophysics Data System (ADS)

Ja'fari, Ahmad; Hamidzadeh Moghadam, Rasoul

2012-10-01

Routine core analysis provides useful information for petrophysical study of the hydrocarbon reservoirs. Effective porosity and fluid conductivity (permeability) could be obtained from core analysis in laboratory. Coring hydrocarbon bearing intervals and analysis of obtained cores in laboratory is expensive and time consuming. In this study an improved method to make a quantitative correlation between porosity and permeability obtained from core and conventional well log data by integration of different artificial intelligent systems is proposed. The proposed method combines the results of adaptive neuro-fuzzy inference system (ANFIS) and neural network (NN) algorithms for overall estimation of core data from conventional well log data. These methods multiply the output of each algorithm with a weight factor. Simple averaging and weighted averaging were used for determining the weight factors. In the weighted averaging method the genetic algorithm (GA) is used to determine the weight factors. The overall algorithm was applied in one of SW Iran’s oil fields with two cored wells. One-third of all data were used as the test dataset and the rest of them were used for training the networks. Results show that the output of the GA averaging method provided the best mean square error and also the best correlation coefficient with real core data.

3D Magnetic Field Analysis of a Turbine Generator Stator Core-end Region

NASA Astrophysics Data System (ADS)

Wakui, Shinichi; Takahashi, Kazuhiko; Ide, Kazumasa; Takahashi, Miyoshi; Watanabe, Takashi

In this paper we calculated magnetic flux density and eddy current distributions of a 71MVA turbine generator stator core-end using three-dimensional numerical magnetic field analysis. Subsequently, the magnetic flux densities and eddy current densities in the stator core-end region on the no-load and three-phase short circuit conditions obtained by the analysis have good agreements with the measurements. Furthermore, the differences of eddy current and eddy current loss in the stator core-end region for various load conditions are shown numerically. As a result, the facing had an effect that decrease the eddy current loss of the end plate about 84%.

Analysis of core-periphery organization in protein contact networks reveals groups of structurally and functionally critical residues.

PubMed

Isaac, Arnold Emerson; Sinha, Sitabhra

2015-10-01

The representation of proteins as networks of interacting amino acids, referred to as protein contact networks (PCN), and their subsequent analyses using graph theoretic tools, can provide novel insights into the key functional roles of specific groups of residues. We have characterized the networks corresponding to the native states of 66 proteins (belonging to different families) in terms of their core-periphery organization. The resulting hierarchical classification of the amino acid constituents of a protein arranges the residues into successive layers - having higher core order - with increasing connection density, ranging from a sparsely linked periphery to a densely intra-connected core (distinct from the earlier concept of protein core defined in terms of the three-dimensional geometry of the native state, which has least solvent accessibility). Our results show that residues in the inner cores are more conserved than those at the periphery. Underlining the functional importance of the network core, we see that the receptor sites for known ligand molecules of most proteins occur in the innermost core. Furthermore, the association of residues with structural pockets and cavities in binding or active sites increases with the core order. From mutation sensitivity analysis, we show that the probability of deleterious or intolerant mutations also increases with the core order. We also show that stabilization centre residues are in the innermost cores, suggesting that the network core is critically important in maintaining the structural stability of the protein. A publicly available Web resource for performing core-periphery analysis of any protein whose native state is known has been made available by us at http://www.imsc.res.in/ ~sitabhra/proteinKcore/index.html.

Advanced Materials and Solids Analysis Research Core (AMSARC)

EPA Science Inventory

The Advanced Materials and Solids Analysis Research Core (AMSARC), centered at the U.S. Environmental Protection Agency's (EPA) Andrew W. Breidenbach Environmental Research Center in Cincinnati, Ohio, is the foundation for the Agency's solids and surfaces analysis capabilities. ...

Determination of power distribution in the VVER-440 core on the basis of data from in-core monitors by means of a metric analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kryanev, A. V.; Udumyan, D. K.; Kurchenkov, A. Yu., E-mail: s327@vver.kiae.ru

2014-12-15

Problems associated with determining the power distribution in the VVER-440 core on the basis of a neutron-physics calculation and data from in-core monitors are considered. A new mathematical scheme is proposed for this on the basis of a metric analysis. In relation to the existing mathematical schemes, the scheme in question improves the accuracy and reliability of the resulting power distribution.

Publications - GMC 385 | Alaska Division of Geological & Geophysical

Science.gov Websites

DGGS GMC 385 Publication Details Title: Porosity, permeability, and capillary pressure core analysis Shimer, G., 2011, Porosity, permeability, and capillary pressure core analysis results (2,124'-2,193 -capilar.xls (108.0 K) gmc385-cores-water.xls (19.0 K) Keywords Oil and Gas; Permeability; Porosity Top of Page

An Interpretative Phenomenological Analysis of the Common Core Standards Program in the State of South Dakota

ERIC Educational Resources Information Center

Alase, Abayomi

2017-01-01

This interpretative phenomenological analysis (IPA) study investigated and interpreted the Common Core State Standards program (the phenomenon) that has been the dominating topic of discussions amongst educators all across the country since the inauguration of the program in 2014/2015 school session. Common Core State Standards (CCSS) was a…

GALFIT-CORSAIR: Implementing the Core-Sérsic Model Into GALFIT

NASA Astrophysics Data System (ADS)

Bonfini, Paolo

2014-10-01

We introduce GALFIT-CORSAIR: a publicly available, fully retro-compatible modification of the 2D fitting software GALFIT (v.3) which adds an implementation of the core-Sersic model. We demonstrate the software by fitting the images of NGC 5557 and NGC 5813, which have been previously identified as core-Sersic galaxies by their 1D radial light profiles. These two examples are representative of different dust obscuration conditions, and of bulge/disk decomposition. To perform the analysis, we obtained deep Hubble Legacy Archive (HLA) mosaics in the F555W filter (~V-band). We successfully reproduce the results of the previous 1D analysis, modulo the intrinsic differences between the 1D and the 2D fitting procedures. The code and the analysis procedure described here have been developed for the first coherent 2D analysis of a sample of core-Sersic galaxies, which will be presented in a forth-coming paper. As the 2D analysis provides better constraining on multi-component fitting, and is fully seeing-corrected, it will yield complementary constraints on the missing mass in depleted galaxy cores.

Brazilian Portuguese version of the CORE-OM: cross-cultural adaptation of an instrument to assess the efficacy and effectiveness of psychotherapy.

PubMed

Santana, Márcia Rosane Moreira; da Silva, Marília Marques; de Moraes, Danielle Souza; Fukuda, Cláudia Cristina; Freitas, Lucia Helena; Ramos, Maria Eveline Cascardo; Fleury, Heloísa Junqueira; Evans, Chris

2015-01-01

The Clinical Outcome in Routine Evaluation - Outcome Measurement (CORE-OM) was developed in the 1990s, with the aim of assessing the efficacy and effectiveness of mental health treatments. To adapt the CORE-OM for use in the Brazilian population. The instrument was translated and adapted based on the international protocol developed by the CORE System Trust which contains seven steps: translation, semantic equivalence analysis, synthesis of the translated versions, pre-testing in the target population, data analysis and back translation. After semantic analysis, modifications were necessary in seven of the 34 original items. Changes were made to avoid repetition of words and the use of terms difficult to understand. Internal consistency analysis showed evidence of score stability in the CORE-OM adapted to Brazilian Portuguese. The instrument was successfully adapted to Brazilian Portuguese, and its semantic and conceptual properties were equivalent to those of the original instrument.

Development of a three-dimensional core dynamics analysis program for commercial boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bessho, Yasunori; Yokomizo, Osamu; Yoshimoto, Yuichiro

1997-03-01

Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates calculational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory size as attained by the latest supercomputer technology. The program consists of six major modules, namely a core neutronics module, a fuel heat conduction/transfer module, a fuel channel thermal-hydraulic module, an upper plenum/separator module, a feedwater/recirculation flow module, and amore » control system module. Its core neutronics module is based on the modified one-group neutron kinetics equation with the prompt jump approximation and with six delayed neutron precursor groups. The module is used to analyze one fuel bundle of the reactor core with one mesh (region). The fuel heat conduction/transfer module solves the one-dimensional heat conduction equation in the radial direction with ten nodes in the fuel pin. The fuel channel thermal-hydraulic module is based on separated three-equation, two-phase flow equations with the drift flux correlation, and it analyzes one fuel bundle of the reactor core with one channel to evaluate flow redistribution between channels precisely. Thermal margin is evaluated by using the GEXL correlation, for example, in the module.« less

Cyclic crack growth behavior of reactor pressure vessel steels in light water reactor environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Der Sluys, W.A.; Emanuelson, R.H.

1986-01-01

During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and SA533b-1 steels in simulated 550/sup 0/F boiling water reactor (BWR) and 550/sup 0/F pressurized water reactor (PWR) environments. Areas investigated over the coursemore » of the test program included the effects of loading frequency and r ratio (Kmin-Kmax) on crack growth rate as a function of the stress intensity factor (deltaK) range. In addition, the effect of sulfur content of the test material on the cyclic crack growth rate was studied. Cyclic crack growth rates were found to be controlled by deltaK, R ratio, and loading frequency. The sulfur impurity content of the reactor pressure vessel steels studied had a significant effect on the cyclic crack growth rates. The higher growth rates were always associated with materials of higher sulfur content. For a given level of sulfur, growth rates were in a 550/sup 0/F simulated BWR environment than in a 550/sup 0/F simulated PWR environment. In both environments cyclic crack growth rates were a strong function of the loading frequency.« less

The effect of zinc injection on the increasing of Inconel 600 TT corrosion resistances

NASA Astrophysics Data System (ADS)

Febrianto; Sriyono; Widodo, Surip; Sunaryo, Geni Rina

2018-02-01

Many failures were found in reactor pressure vessel head penetration (RPV) head material. Those failures caused by boric acid corrosion, and from visual examination were found a big hole and white deposit crystal of boric acid during shutdown maintenance at David Besse reactor. Zinc Oxide addition in BWR reactor known as Zinc Injection that has purposed to reduce radiation exposure cause of Hydrogen addition. Beside reducing the radiation exposure, Zinc injection also has an effect in reducing material corrosion. The purpose of study is to determine the effect of zinc addition, boric acid, temperature also the effects of Cobalt Nitrate and Zinc Oxide addition to Inconel 600 TT as RPV head penetration material. The result in the BWR reactor experience will be implementated at PWR reactor, weather zinc oxide addition also has an effect in reducing the corrosion of Inconel 600. The method that used in this research is to observe the corrosion rates for Inconel 600 material using Potentiostat. Examination were conducted in 30, 40, 60, 70, 80 and 80 °C using 1000, 1500, 2000, 2500 and 3000 ppm boric acid concentration. The results showed that the corrosion rate for the material were very small, but the highest corrosion rate occurred in 3000 ppm boric acid concentration at 90 °C with Cobalt Nitrate addition, around 5.210 x 10-1 mpy. In the same condition at 3000 ppm boric acid concentration for temperature at 90 °C, Inconel 600 TT corrosion rate is smaller with Zinc oxide addition, around 4.631 x 10-1 mpy.

Motion of the Mantle in the Translational Modes of the Earth and Mercury

NASA Technical Reports Server (NTRS)

Grinfeld, Pavel; Wisdom, Jack

2005-01-01

Slichter modes refer to the translational motion of the inner core with respect to the outer core and the mantle. The polar Slichter mode is the motion of the inner core along the axis of rotation. Busse presented an analysis of the polar mode which yielded an expression for its period. Busse's analysis included the assumption that the mantle was stationary. This approximation is valid for planets with small inner cores, such as the Earth whose inner core is about 1/60 of the total planet mass. On the other hand, many believe that Mercury's core may be enormous. If so, the motion of the mantle should be expected to produce a significant effect. We present a formal framework for including the motion of the mantle in the analysis of the translational motion of the inner core. We analyze the effect of the motion of the mantle on the Slichter modes for a non-rotating planet with an inner core of arbitrary size. We omit the effects of viscosity in the outer core, magnetic effects, and solid tides. Our approach is perturbative and is based on a linearization of Euler's equations for the motion of the fluid and Newton's second law for the motion of the inner core. We find an analytical expression for the period of the Slichter mode. Our result agrees with Busse's in the limiting case of small inner core. We present the unexpected result that even for Mercury the motion of the mantle does not significantly change the period of oscillation.

Preliminary Thermal Modeling of Hi-Storm 100S-218 Version B Storage Modules at Hope Creek Nuclear Power Station ISFSI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cuta, Judith M.; Adkins, Harold E.

2013-08-30

This report fulfills the M3 milestone M3FT-13PN0810022, “Report on Inspection 1”, under Work Package FT-13PN081002. Thermal analysis is being undertaken at Pacific Northwest National Laboratory (PNNL) in support of inspections of selected storage modules at various locations around the United States, as part of the Used Fuel Disposition Campaign of the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) Fuel Cycle Research and Development. This report documents pre-inspection predictions of temperatures for four modules at the Hope Creek Nuclear Generating Station ISFSI that have been identified as candidates for inspection in late summer or early fall/winter of 2013. Thesemore » are HI-STORM 100S-218 Version B modules storing BWR 8x8 fuel in MPC-68 canisters. The temperature predictions reported in this document were obtained with detailed COBRA-SFS models of these four storage systems, with the following boundary conditions and assumptions.« less

A New Capability for Nuclear Thermal Propulsion Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amiri, Benjamin W.; Nuclear and Radiological Engineering Department, University of Florida, Gainesville, FL 32611; Kapernick, Richard J.

2007-01-30

This paper describes a new capability for Nuclear Thermal Propulsion (NTP) design that has been developed, and presents the results of some analyses performed with this design tool. The purpose of the tool is to design to specified mission and material limits, while maximizing system thrust to weight. The head end of the design tool utilizes the ROCket Engine Transient Simulation (ROCETS) code to generate a system design and system design requirements as inputs to the core analysis. ROCETS is a modular system level code which has been used extensively in the liquid rocket engine industry for many years. Themore » core design tool performs high-fidelity reactor core nuclear and thermal-hydraulic design analysis. At the heart of this process are two codes TMSS-NTP and NTPgen, which together greatly automate the analysis, providing the capability to rapidly produce designs that meet all specified requirements while minimizing mass. A PERL based command script, called CORE DESIGNER controls the execution of these two codes, and checks for convergence throughout the process. TMSS-NTP is executed first, to produce a suite of core designs that meet the specified reactor core mechanical, thermal-hydraulic and structural requirements. The suite of designs consists of a set of core layouts and, for each core layout specific designs that span a range of core fuel volumes. NTPgen generates MCNPX models for each of the core designs from TMSS-NTP. Iterative analyses are performed in NTPgen until a reactor design (fuel volume) is identified for each core layout that meets cold and hot operation reactivity requirements and that is zoned to meet a radial core power distribution requirement.« less

Biomechanical Evaluation of a Tooth Restored with High Performance Polymer PEKK Post-Core System: A 3D Finite Element Analysis.

PubMed

Lee, Ki-Sun; Shin, Joo-Hee; Kim, Jong-Eun; Kim, Jee-Hwan; Lee, Won-Chang; Shin, Sang-Wan; Lee, Jeong-Yol

2017-01-01

The aim of this study was to evaluate the biomechanical behavior and long-term safety of high performance polymer PEKK as an intraradicular dental post-core material through comparative finite element analysis (FEA) with other conventional post-core materials. A 3D FEA model of a maxillary central incisor was constructed. A cyclic loading force of 50 N was applied at an angle of 45° to the longitudinal axis of the tooth at the palatal surface of the crown. For comparison with traditionally used post-core materials, three materials (gold, fiberglass, and PEKK) were simulated to determine their post-core properties. PEKK, with a lower elastic modulus than root dentin, showed comparably high failure resistance and a more favorable stress distribution than conventional post-core material. However, the PEKK post-core system showed a higher probability of debonding and crown failure under long-term cyclic loading than the metal or fiberglass post-core systems.

Biomechanical Evaluation of a Tooth Restored with High Performance Polymer PEKK Post-Core System: A 3D Finite Element Analysis

PubMed Central

Shin, Joo-Hee; Kim, Jong-Eun; Kim, Jee-Hwan; Lee, Won-Chang; Shin, Sang-Wan

2017-01-01

The aim of this study was to evaluate the biomechanical behavior and long-term safety of high performance polymer PEKK as an intraradicular dental post-core material through comparative finite element analysis (FEA) with other conventional post-core materials. A 3D FEA model of a maxillary central incisor was constructed. A cyclic loading force of 50 N was applied at an angle of 45° to the longitudinal axis of the tooth at the palatal surface of the crown. For comparison with traditionally used post-core materials, three materials (gold, fiberglass, and PEKK) were simulated to determine their post-core properties. PEKK, with a lower elastic modulus than root dentin, showed comparably high failure resistance and a more favorable stress distribution than conventional post-core material. However, the PEKK post-core system showed a higher probability of debonding and crown failure under long-term cyclic loading than the metal or fiberglass post-core systems. PMID:28386547

Comparative Analysis of Hexagonal Solid Silica and Nitro-benzene Filled Hollow Core Photonic Crystal Fiber

NASA Astrophysics Data System (ADS)

Shahiruddin; Singh, Dharmendra K.; Hassan, M. A.

2018-02-01

A comparative study of five ring solid core and nitrobenzene filled hollow core liquid filled photonic crystal fiber (PCF) are presented. Considering the same structure, one is used as solid silica and another one is filled with nitrobenzene in the core. Here the paper elaborates the confinement loss, dispersion properties and birefringence of an index-guiding PCF with asymmetric cladding designed and analyzed by the finite-element method. The proposed structure shows the low confinement loss in case of solid silica, negative dispersion in nitrobenzene filled hollow core PCF and high birefringence in both the cases. The calculated values shows flat zero confinement loss in 0.7 µm to 1.54 µm range, flat zero dispersion is achieved in solid core and -2000 ps/km-nm in nitrobenzene filled hollow core PCF and high birefringence in the range of 10-3 in nitrobenzene filled hollow core PCF. Results show the relative analysis at different air fill fraction.

Methodology of full-core Monte Carlo calculations with leakage parameter evaluations for benchmark critical experiment analysis

NASA Astrophysics Data System (ADS)

Sboev, A. G.; Ilyashenko, A. S.; Vetrova, O. A.

1997-02-01

The method of bucking evaluation, realized in the MOnte Carlo code MCS, is described. This method was applied for calculational analysis of well known light water experiments TRX-1 and TRX-2. The analysis of this comparison shows, that there is no coincidence between Monte Carlo calculations, obtained by different ways: the MCS calculations with given experimental bucklings; the MCS calculations with given bucklings evaluated on base of full core MCS direct simulations; the full core MCNP and MCS direct simulations; the MCNP and MCS calculations, where the results of cell calculations are corrected by the coefficients taking into the account the leakage from the core. Also the buckling values evaluated by full core MCS calculations have differed from experimental ones, especially in the case of TRX-1, when this difference has corresponded to 0.5 percent increase of Keff value.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

PubMed

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

PubMed Central

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

US Efforts in Support of Examinations at Fukushima Daiichi – 2016 Evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amway, P.; Andrews, N.; Bixby, Willis

Although it is clear that the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limitedmore » full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings (TEPCO) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document reports recent results from the US Forensics Effort to use information obtained by TEPCO to enhance the safety of existing and future nuclear power plant designs. This Forensics Effort, which is sponsored by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of US experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO information from Daiichi that address these needs. Examples presented in this report demonstrate that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities.« less

Compression After Impact on Honeycomb Core Sandwich Panels with Thin Facesheets, Part 2: Analysis

NASA Technical Reports Server (NTRS)

Mcquigg, Thomas D.; Kapania, Rakesh K.; Scotti, Stephen J.; Walker, Sandra P.

2012-01-01

A two part research study has been completed on the topic of compression after impact (CAI) of thin facesheet honeycomb core sandwich panels. The research has focused on both experiments and analysis in an effort to establish and validate a new understanding of the damage tolerance of these materials. Part 2, the subject of the current paper, is focused on the analysis, which corresponds to the CAI testings described in Part 1. Of interest, are sandwich panels, with aerospace applications, which consist of very thin, woven S2-fiberglass (with MTM45-1 epoxy) facesheets adhered to a Nomex honeycomb core. Two sets of materials, which were identical with the exception of the density of the honeycomb core, were tested in Part 1. The results highlighted the need for analysis methods which taken into account multiple failure modes. A finite element model (FEM) is developed here, in Part 2. A commercial implementation of the Multicontinuum Failure Theory (MCT) for progressive failure analysis (PFA) in composite laminates, Helius:MCT, is included in this model. The inclusion of PFA in the present model provided a new, unique ability to account for multiple failure modes. In addition, significant impact damage detail is included in the model. A sensitivity study, used to assess the effect of each damage parameter on overall analysis results, is included in an appendix. Analysis results are compared to the experimental results for each of the 32 CAI sandwich panel specimens tested to failure. The failure of each specimen is predicted using the high-fidelity, physicsbased analysis model developed here, and the results highlight key improvements in the understanding of honeycomb core sandwich panel CAI failure. Finally, a parametric study highlights the strength benefits compared to mass penalty for various core densities.

Tracking Student Progression through the Core Curriculum. CCRC Analytics

ERIC Educational Resources Information Center

Hodara, Michelle; Rodriguez, Olga

2013-01-01

This report demonstrates useful methods for examining student progression through the core curriculum. The authors carry out analyses at two colleges in two different states, illustrating students' overall progression through the core curriculum and the relationship of this "core" progression to their college outcomes. By means of this analysis,…

Waveguides for performing enzymatic reactions

DOEpatents

Levene; Michael J. , Korlach; Jonas , Turner; Stephen W. , Craighead; Harold G. , Webb; Watt W.

2007-11-06

The present invention is directed to a method and an apparatus for analysis of an analyte. The method involves providing a zero-mode waveguide which includes a cladding surrounding a core where the cladding is configured to preclude propagation of electromagnetic energy of a frequency less than a cutoff frequency longitudinally through the core of the zero-mode waveguide. The analyte is positioned in the core of the zero-mode waveguide and is then subjected, in the core of the zero-mode wave guide, to activating electromagnetic radiation of a frequency less than the cut-off frequency under conditions effective to permit analysis of the analyte in an effective observation volume which is more compact than if the analysis were carried out in the absence of the zero-mode waveguide.

Zero-mode waveguides

DOEpatents

Levene, Michael J.; Korlach, Jonas; Turner, Stephen W.; Craighead, Harold G.; Webb, Watt W.

2007-02-20

The present invention is directed to a method and an apparatus for analysis of an analyte. The method involves providing a zero-mode waveguide which includes a cladding surrounding a core where the cladding is configured to preclude propagation of electromagnetic energy of a frequency less than a cutoff frequency longitudinally through the core of the zero-mode waveguide. The analyte is positioned in the core of the zero-mode waveguide and is then subjected, in the core of the zero-mode waveguide, to activating electromagnetic radiation of a frequency less than the cut-off frequency under conditions effective to permit analysis of the analyte in an effective observation volume which is more compact than if the analysis were carried out in the absence of the zero-mode waveguide.

ANSYS-based birefringence property analysis of side-hole fiber induced by pressure and temperature

NASA Astrophysics Data System (ADS)

Zhou, Xinbang; Gong, Zhenfeng

2018-03-01

In this paper, we theoretically investigate the influences of pressure and temperature on the birefringence property of side-hole fibers with different shapes of holes using the finite element analysis method. A physical mechanism of the birefringence of the side-hole fiber is discussed with the presence of different external pressures and temperatures. The strain field distribution and birefringence values of circular-core, rectangular-core, and triangular-core side-hole fibers are presented. Our analysis shows the triangular-core side-hole fiber has low temperature sensitivity which weakens the cross sensitivity of temperature and strain. Additionally, an optimized structure design of the side-hole fiber is presented which can be used for the sensing application.

Publications - GMC 217 | Alaska Division of Geological & Geophysical

Science.gov Websites

DGGS GMC 217 Publication Details Title: Results of core analysis (ambient porosity and grain density Wilson and Associates, and Geocore, 1993, Results of core analysis (ambient porosity and grain density

Publications - GMC 335 | Alaska Division of Geological & Geophysical

Science.gov Websites

DGGS GMC 335 Publication Details Title: Geochemical analysis of core (3340'-3625') from the BP Reference ExxonMobil, 2006, Geochemical analysis of core (3340'-3625') from the BP Exploration (Alaska) Inc

Divers muscle Fitzpatrick`s mussels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hobbs, B.; Kahabka, J.

1996-01-01

This article describes how an effective manual cleaning technique has helped rid submerged intake structures of this passive-aggressive pest. The New York Power Authority`s (NYPA) James A. Fitzpatrick (JAF) Nuclear Power Plant is located in Lycoming, NY, on the southeast shore of Lake Ontario. An 850-MWe, GE-design boiling water reactor (BWR), the unit has been in service since 1975. Water drawn from the Lake supplies cooling to plant loads via the circulating water system and three service water systems. These share a common intake system consisting of an offshore cap (crib), horizontal intake tunnel, two vertical risers and forebay/screenwell area.

Spent fuel data base: commercial light water reactors. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

Barium and calcium analyses in sediment cores using µ-XRF core scanners

NASA Astrophysics Data System (ADS)

Acar, Dursun; Çaǧatay, Namık; Genç, S. Can; Eriş, K. Kadir; Sarı, Erol; Uçarkus, Gülsen

2017-04-01

Barium and Ca are used as proxies for organic productivity in paleooceanographic studies. With its heavy atomic weight (137.33 u), barium is easily detectable in small concentrations (several ppm levels) in marine sediments using XRF methods, including the analysis by µ-XRF core scanners. Calcium has an intermediate atomic weight (40.078 u) but is a major element in the earth's crust and in sediments and sedimentary rocks, and hence it is easily detectable by µ-XRF techniques. Normally, µ-XRF elemental analysis of cores are carried out using split half cores or 1-2 cm thich u-channels with an original moisture. Sediment cores show variation in different water content (and porosity) along their length. This in turn results in variation in the XRF counts of the elements and causes error in the elemental concentrations. We tried µ-XRF elemental analysis of split half cores, subsampled as 1 cm thick u-channels with original moisture and 0.3 mm-thin film slices of the core with original wet sample and after air drying with humidity protector mylar film. We found considerable increase in counts of most elements, and in particular for Ba and Ca, when we used 0.3 mm thin film, dried slice. In the case of Ba, the counts increased about three times that of the analysis made with wet and 1 cm thick u-channels. The higher Ba and Ca counts are mainly due to the possible precipitation of Ba as barite and Ca as gypsum from oxidation of Fe-sulphides and the evaporation of pore waters. The secondary barite and gypsum precipitation would be especially serious in unoxic sediment units, such as sapropels, with considerable Fe-sulphides and bio-barite.It is therefore suggested that reseachers should be cautious of such secondary precipitation on core surfaces when analyzing cores that have long been exposed to the atmospheric conditions.

Determination of lead isotopes in a new Greenland deep ice core at the sub-picogram per gram level by thermal ionization mass spectrometry using an improved decontamination method.

PubMed

Han, Changhee; Burn-Nunes, Laurie J; Lee, Khanghyun; Chang, Chaewon; Kang, Jung-Ho; Han, Yeongcheol; Hur, Soon Do; Hong, Sungmin

2015-08-01

An improved decontamination method and ultraclean analytical procedures have been developed to minimize Pb contamination of processed glacial ice cores and to achieve reliable determination of Pb isotopes in North Greenland Eemian Ice Drilling (NEEM) deep ice core sections with concentrations at the sub-picogram per gram level. A PL-7 (Fuso Chemical) silica-gel activator has replaced the previously used colloidal silica activator produced by Merck and has been shown to provide sufficiently enhanced ion beam intensity for Pb isotope analysis for a few tens of picograms of Pb. Considering the quantities of Pb contained in the NEEM Greenland ice core and a sample weight of 10 g used for the analysis, the blank contribution from the sample treatment was observed to be negligible. The decontamination and analysis of the artificial ice cores and selected NEEM Greenland ice core sections confirmed the cleanliness and effectiveness of the overall analytical process. Copyright © 2015 Elsevier B.V. All rights reserved.

Magnetic resonance imaging in laboratory petrophysical core analysis

NASA Astrophysics Data System (ADS)

Mitchell, J.; Chandrasekera, T. C.; Holland, D. J.; Gladden, L. F.; Fordham, E. J.

2013-05-01

Magnetic resonance imaging (MRI) is a well-known technique in medical diagnosis and materials science. In the more specialized arena of laboratory-scale petrophysical rock core analysis, the role of MRI has undergone a substantial change in focus over the last three decades. Initially, alongside the continual drive to exploit higher magnetic field strengths in MRI applications for medicine and chemistry, the same trend was followed in core analysis. However, the spatial resolution achievable in heterogeneous porous media is inherently limited due to the magnetic susceptibility contrast between solid and fluid. As a result, imaging resolution at the length-scale of typical pore diameters is not practical and so MRI of core-plugs has often been viewed as an inappropriate use of expensive magnetic resonance facilities. Recently, there has been a paradigm shift in the use of MRI in laboratory-scale core analysis. The focus is now on acquiring data in the laboratory that are directly comparable to data obtained from magnetic resonance well-logging tools (i.e., a common physics of measurement). To maintain consistency with well-logging instrumentation, it is desirable to measure distributions of transverse (T2) relaxation time-the industry-standard metric in well-logging-at the laboratory-scale. These T2 distributions can be spatially resolved over the length of a core-plug. The use of low-field magnets in the laboratory environment is optimal for core analysis not only because the magnetic field strength is closer to that of well-logging tools, but also because the magnetic susceptibility contrast is minimized, allowing the acquisition of quantitative image voxel (or pixel) intensities that are directly scalable to liquid volume. Beyond simple determination of macroscopic rock heterogeneity, it is possible to utilize the spatial resolution for monitoring forced displacement of oil by water or chemical agents, determining capillary pressure curves, and estimating wettability. The history of MRI in petrophysics is reviewed and future directions considered, including advanced data processing techniques such as compressed sensing reconstruction and Bayesian inference analysis of under-sampled data. Although this review focuses on rock core analysis, the techniques described are applicable in a wider context to porous media in general, such as cements, soils, ceramics, and catalytic materials.

Real-Time and Delayed Analysis of Tree and Shrub Cores as Indicators of Subsurface Volatile Organic Compound Contamination, Durham Meadows Superfund Site, Durham, Connecticut, August 29, 2006

USGS Publications Warehouse

Vroblesky, Don A.; Willey, Richard E.; Clifford, Scott; Murphy, James J.

2008-01-01

This study examined volatile organic compound concentrations in cores from trees and shrubs for use as indicators of vadose-zone contamination or potential vapor intrusion by volatile organic compounds into buildings at the Durham Meadows Superfund Site, Durham, Connecticut. The study used both (1) real-time tree- and shrub-core analysis, which involved field heating the core samples for 5 to 10 minutes prior to field analysis, and (2) delayed analysis, which involved allowing the gases in the cores to equilibrate with the headspace gas in the sample vials unheated for 1 to 2 days prior to analysis. General correspondence was found between the two approaches, indicating that preheating and field analysis of vegetation cores is a viable approach to real-time monitoring of subsurface volatile organic compounds. In most cases, volatile organic compounds in cores from trees and shrubs at the Merriam Manufacturing Company property showed a general correspondence to the distribution of volatile organic compounds detected in a soil-gas survey, despite the fact that most of the soil-gas survey data in close proximity to the relevant trees were collected about 3 years prior to the tree-core collection. Most of the trees cored at the Durham Meadows Superfund Site, outside of the Merriam Manufacturing Company property, contained no volatile organic compounds and were in areas where indoor air sampling and soil-gas sampling showed little or no volatile organic compound concentrations. An exception was tree DM11, which contained barely detectable concentrations of trichloroethene near a house where previous investigations found low concentrations of trichloroethene (0.13 to 1.2 parts per billion by volume) in indoor air and 7.7 micrograms per liter of trichloroethene in the ground water. The barely detectable concentration of trichloroethene in tree DM11 and the lack of volatile organic compound detection in nearby tree DM10 (adjacent to the well having 7.7 micrograms of trichloroethene) may be attributable to the relatively large depth to water (17.6 feet), the relatively low soil-vapor trichloroethene concentration, and the large amount of rainfall during and preceding the tree-coring event. The data indicate that real-time and delayed analyses of tree cores are viable approaches to examining subsurface volatile organic compound soil-gas or vadose-zone contamination at the Durham Meadows Superfund Site and other similar sites. Thus, the methods may have application for determining the potential for vapor intrusion into buildings.

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope

2011-10-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh

2010-09-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrews, Nathan; Faucett, Christopher; Haskin, Troy Christopher

Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recoverymore » of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.« less

Logging-while-coring method and apparatus

DOEpatents

Goldberg, David S.; Myers, Gregory J.

2007-11-13

A method and apparatus for downhole coring while receiving logging-while-drilling tool data. The apparatus includes core collar and a retrievable core barrel. The retrievable core barrel receives core from a borehole which is sent to the surface for analysis via wireline and latching tool The core collar includes logging-while-drilling tools for the simultaneous measurement of formation properties during the core excavation process. Examples of logging-while-drilling tools include nuclear sensors, resistivity sensors, gamma ray sensors, and bit resistivity sensors. The disclosed method allows for precise core-log depth calibration and core orientation within a single borehole, and without at pipe trip, providing both time saving and unique scientific advantages.

Logging-while-coring method and apparatus

DOEpatents

Goldberg, David S.; Myers, Gregory J.

2007-01-30

A method and apparatus for downhole coring while receiving logging-while-drilling tool data. The apparatus includes core collar and a retrievable core barrel. The retrievable core barrel receives core from a borehole which is sent to the surface for analysis via wireline and latching tool The core collar includes logging-while-drilling tools for the simultaneous measurement of formation properties during the core excavation process. Examples of logging-while-drilling tools include nuclear sensors, resistivity sensors, gamma ray sensors, and bit resistivity sensors. The disclosed method allows for precise core-log depth calibration and core orientation within a single borehole, and without at pipe trip, providing both time saving and unique scientific advantages.

Synthesis of parallel and antiparallel core-shell triangular nanoparticles

NASA Astrophysics Data System (ADS)

Bhattacharjee, Gourab; Satpati, Biswarup

2018-04-01

Core-shell triangular nanoparticles were synthesized by seed mediated growth. Using triangular gold (Au) nanoparticle as template, we have grown silver (Ag) shellto get core-shell nanoparticle. Here by changing the chemistry we have grown two types of core-shell structures where core and shell is having same symmetry and also having opposite symmetry. Both core and core-shell nanoparticles were characterized using transmission electron microscopy (TEM) and energy dispersive X-ray spectroscopy (EDX) to know the crystal structure and composition of these synthesized core-shell nanoparticles. From diffraction pattern analysis and energy filtered TEM (EFTEM) we have confirmed the crystal facet in core is responsible for such two dimensional growth of core-shell nanostructures.

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

NASA Astrophysics Data System (ADS)

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

2017-02-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2008-07-15

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

Gap analysis: a method to assess core competency development in the curriculum.

PubMed

Fater, Kerry H

2013-01-01

To determine the extent to which safety and quality improvement core competency development occurs in an undergraduate nursing program. Rapid change and increased complexity of health care environments demands that health care professionals are adequately prepared to provide high quality, safe care. A gap analysis compared the present state of competency development to a desirable (ideal) state. The core competencies, Nurse of the Future Nursing Core Competencies, reflect the ideal state and represent minimal expectations for entry into practice from pre-licensure programs. Findings from the gap analysis suggest significant strengths in numerous competency domains, deficiencies in two competency domains, and areas of redundancy in the curriculum. Gap analysis provides valuable data to direct curriculum revision. Opportunities for competency development were identified, and strategies were created jointly with the practice partner, thereby enhancing relevant knowledge, attitudes, and skills nurses need for clinical practice currently and in the future.

Brightness analysis of an electron beam with a complex profile

NASA Astrophysics Data System (ADS)

Maesaka, Hirokazu; Hara, Toru; Togawa, Kazuaki; Inagaki, Takahiro; Tanaka, Hitoshi

2018-05-01

We propose a novel analysis method to obtain the core bright part of an electron beam with a complex phase-space profile. This method is beneficial to evaluate the performance of simulation data of a linear accelerator (linac), such as an x-ray free electron laser (XFEL) machine, since the phase-space distribution of a linac electron beam is not simple, compared to a Gaussian beam in a synchrotron. In this analysis, the brightness of undulator radiation is calculated and the core of an electron beam is determined by maximizing the brightness. We successfully extracted core electrons from a complex beam profile of XFEL simulation data, which was not expressed by a set of slice parameters. FEL simulations showed that the FEL intensity was well remained even after extracting the core part. Consequently, the FEL performance can be estimated by this analysis without time-consuming FEL simulations.

Publications - GMC 313 | Alaska Division of Geological & Geophysical

Science.gov Websites

'-13,075.0') from the Pan American Redoubt Shoal State (29690) #1 well Authors: Saltmarsh, Art, and Core . Bibliographic Reference Saltmarsh, Art, and Core Laboratories, 2004, Porosity and permeability core analysis

Combustion and Engine-Core Noise

NASA Astrophysics Data System (ADS)

Ihme, Matthias

2017-01-01

The implementation of advanced low-emission aircraft engine technologies and the reduction of noise from airframe, fan, and jet exhaust have made noise contributions from an engine core increasingly important. Therefore, meeting future ambitious noise-reduction goals requires the consideration of engine-core noise. This article reviews progress on the fundamental understanding, experimental analysis, and modeling of engine-core noise; addresses limitations of current techniques; and identifies opportunities for future research. After identifying core-noise contributions from the combustor, turbomachinery, nozzles, and jet exhaust, they are examined in detail. Contributions from direct combustion noise, originating from unsteady combustion, and indirect combustion noise, resulting from the interaction of flow-field perturbations with mean-flow variations in turbine stages and nozzles, are analyzed. A new indirect noise-source contribution arising from mixture inhomogeneities is identified by extending the theory. Although typically omitted in core-noise analysis, the impact of mean-flow variations and nozzle-upstream perturbations on the jet-noise modulation is examined, providing potential avenues for future core-noise mitigation.

Subannual layer variability in Greenland firn cores

NASA Astrophysics Data System (ADS)

Kjær, Helle Astrid; Vallelonga, Paul; Vinther, Bo; Winstrup, Mai; Simonsen, Marius; Maffezzoli, Niccoló; Jensen, Camilla Marie

2017-04-01

Ice cores are used to infer information about the past and modern techniques allow for high resolution (< cm) continuous flow analysis (CFA) of the ice. Such analysis is often used to inform on annual layers to constrain dating of ice cores, but can also be extended to provide information on sub-annual deposition patterns. In this study we use available high resolution data from multiple shallow cores around Greenland to investigate the seasonality and trends in the most often continuously measured components sodium, insoluble dust, calcium, ammonium and conductivity (or acidity) from 1800 AD to today. We evaluate the similarities and differences between the records and discuss the causes from different sources and transport to deposition and post-deposition effects over differences in measurement set up. Further we add to the array of cores already published with measurements from the newly drilled ReCAP ice core from a coastal ice cap in eastern Greenland and from a shallow core drilled at the high accumulation site at the Greenland South Dome.

Structural Interface Forms and Their Involvement in Stabilization of Multidomain Proteins or Protein Complexes.

PubMed

Dygut, Jacek; Kalinowska, Barbara; Banach, Mateusz; Piwowar, Monika; Konieczny, Leszek; Roterman, Irena

2016-10-18

The presented analysis concerns the inter-domain and inter-protein interface in protein complexes. We propose extending the traditional understanding of the protein domain as a function of local compactness with an additional criterion which refers to the presence of a well-defined hydrophobic core. Interface areas in selected homodimers vary with respect to their contribution to share as well as individual (domain-specific) hydrophobic cores. The basic definition of a protein domain, i.e., a structural unit characterized by tighter packing than its immediate environment, is extended in order to acknowledge the role of a structured hydrophobic core, which includes the interface area. The hydrophobic properties of interfaces vary depending on the status of interacting domains-In this context we can distinguish: (1) Shared hydrophobic cores (spanning the whole dimer); (2) Individual hydrophobic cores present in each monomer irrespective of whether the dimer contains a shared core. Analysis of interfaces in dystrophin and utrophin indicates the presence of an additional quasi-domain with a prominent hydrophobic core, consisting of fragments contributed by both monomers. In addition, we have also attempted to determine the relationship between the type of interface (as categorized above) and the biological function of each complex. This analysis is entirely based on the fuzzy oil drop model.

Research on Shock Responses of Three Types of Honeycomb Cores

NASA Astrophysics Data System (ADS)

Peng, Fei; Yang, Zhiguang; Jiang, Liangliang; Ren, Yanting

2018-03-01

The shock responses of three kinds of honeycomb cores have been investigated and analyzed based on explicit dynamics analysis. According to the real geometric configuration and the current main manufacturing methods of aluminum alloy honeycomb cores, the finite element models of honeycomb cores with three different cellular configurations (conventional hexagon honeycomb core, rectangle honeycomb core and auxetic honeycomb core with negative Poisson’s ratio) have been established through FEM parametric modeling method based on Python and Abaqus. In order to highlight the impact response characteristics of the above three honeycomb cores, a 5 mm thick panel with the same mass and material was taken as contrast. The analysis results showed that the peak values of longitudinal acceleration history curves of the three honeycomb cores were lower than those of the aluminum alloy panel in all three reference points under the loading of a longitudinal pulse pressure load with the peak value of 1 MPa and the pulse width of 1 μs. It could be concluded that due to the complex reflection and diffraction of stress wave induced by shock in honeycomb structures, the impact energy was redistributed which led to a decrease in the peak values of the longitudinal acceleration at the measuring points of honeycomb cores relative to the panel.

Zero-mode clad waveguides for performing spectroscopy with confined effective observation volumes

DOEpatents

Levene, Michael J.; Korlach, Jonas; Turner, Stephen W.; Craighead, Harold G.; Webb, Watt W.

2005-07-12

The present invention is directed to a method and an apparatus for analysis of an analyte. The method involves providing a zero-mode waveguide which includes a cladding surrounding a core where the cladding is configured to preclude propagation of electromagnetic energy of a frequency less than a cutoff frequency longitudinally through the core of the zero-mode waveguide. The analyte is positioned in the core of the zero-mode waveguide and is then subjected, in the core of the zero-mode waveguide, to activating electromagnetic radiation of a frequency less than the cut-off frequency under conditions effective to permit analysis of the analyte in an effective observation volume which is more compact than if the analysis were carried out in the absence of the zero-mode waveguide.

Waveguides for performing spectroscopy with confined effective observation volumes

DOEpatents

Levene, Michael J.; Korlach, Jonas; Turner, Stephen W.; Craighead, Harold G.; Webb, Watt W.

2006-03-14

The present invention is directed to a method and an apparatus for analysis of an analyte. The method involves providing a zero-mode waveguide which includes a cladding surrounding a core where the cladding is configured to preclude propagation of electromagnetic energy of a frequency less than a cutoff frequency longitudinally through the core of the zero-mode waveguide. The analyte is positioned in the core of the zero-mode waveguide and is then subjected, in the core of the zero-mode waveguide, to activating electromagnetic radiation of a frequency less than the cut-off frequency under conditions effective to permit analysis of the analyte in an effective observation volume which is more compact than if the analysis were carried out in the absence of the zero-mode waveguide.

Thermo-Physics Technical Note No. 60: thermal analysis of SNAP 10A reactor core during atmospheric reentry and resulting core disintegration and fuel element separation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mouradian, E.M.

1966-02-16

A thermal analysis is carried out to determine the temperature distribution throughout a SNAP 10A reactor core, particularly in the vicinity of the grid plates, during atmospheric reentry. The transient temperatue distribution of the grid plate indicates when sufficient melting occurs so that fuel elements are free to be released and continue their descent individually.

Application of the Monte Carlo method to estimate doses due to neutron activation of different materials in a nuclear reactor

NASA Astrophysics Data System (ADS)

Ródenas, José

2017-11-01

All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.

Characterizing core-periphery structure of complex network by h-core and fingerprint curve

NASA Astrophysics Data System (ADS)

Li, Simon S.; Ye, Adam Y.; Qi, Eric P.; Stanley, H. Eugene; Ye, Fred Y.

2018-02-01

It is proposed that the core-periphery structure of complex networks can be simulated by h-cores and fingerprint curves. While the features of core structure are characterized by h-core, the features of periphery structure are visualized by rose or spiral curve as the fingerprint curve linking to entire-network parameters. It is suggested that a complex network can be approached by h-core and rose curves as the first-order Fourier-approach, where the core-periphery structure is characterized by five parameters: network h-index, network radius, degree power, network density and average clustering coefficient. The simulation looks Fourier-like analysis.

Dynamic interaction between fetal adversity and a genetic score reflecting dopamine function on developmental outcomes at 36 months.

PubMed

Bischoff, Adrianne R; Pokhvisneva, Irina; Léger, Étienne; Gaudreau, Hélène; Steiner, Meir; Kennedy, James L; O'Donnell, Kieran J; Diorio, Josie; Meaney, Michael J; Silveira, Patrícia P

2017-01-01

Fetal adversity, evidenced by poor fetal growth for instance, is associated with increased risk for several diseases later in life. Classical cut-offs to characterize small (SGA) and large for gestational age (LGA) newborns are used to define long term vulnerability. We aimed at exploring the possible dynamism of different birth weight cut-offs in defining vulnerability in developmental outcomes (through the Bayley Scales of Infant and Toddler Development), using the example of a gene vs. fetal adversity interaction considering gene choices based on functional relevance to the studied outcome. 36-month-old children from an established prospective birth cohort (Maternal Adversity, Vulnerability, and Neurodevelopment) were classified according to birth weight ratio (BWR) (SGA ≤0.85, LGA >1.15, exploring a wide range of other cut-offs) and genotyped for polymorphisms associated with dopamine signaling (TaqIA-A1 allele, DRD2-141C Ins/Ins, DRD4 7-repeat, DAT1-10- repeat, Met/Met-COMT), composing a score based on the described function, in which hypofunctional variants received lower scores. There were 251 children (123 girls and 128 boys). Using the classic cut-offs (0.85 and 1.15), there were no statistically significant interactions between the neonatal groups and the dopamine genetic score. However, when changing the cut-offs, it is possible to see ranges of BWR that could be associated with vulnerability to poorer development according to the variation in the dopamine function. The classic birth weight cut-offs to define SGA and LGA newborns should be seen with caution, as depending on the outcome in question, the protocols for long-term follow up could be either too inclusive-therefore most costly, or unable to screen true vulnerabilities-and therefore ineffective to establish early interventions and primary prevention.

Effects of materials and design on the criticality and shielding assessment of canister concepts for the disposal of spent nuclear fuel.

PubMed

Gutiérrez, Miguel Morales; Caruso, Stefano; Diomidis, Nikitas

2018-05-19

According to the Swiss disposal concept, the safety of a deep geological repository for spent nuclear fuel (SNF) is based on a multi-barrier system. The disposal canister is an important component of the engineered barrier system, aiming to provide containment of the SNF for thousands of years. This study evaluates the criticality safety and shielding of candidate disposal canister concepts, focusing on the fulfilment of the sub-criticality criterion and on limiting radiolysis processes at the outer surface of the canister which can enhance corrosion mechanisms. The effective neutron multiplication factor (k-eff) and the surface dose rates are calculated for three different canister designs and material combinations for boiling water reactor (BWR) canisters, containing 12 spent fuel assemblies (SFA), and pressurized water reactor (PWR) canisters, with 4 SFAs. For each configuration, individual criticality and shielding calculations were carried out. The results show that k-eff falls below the defined upper safety limit (USL) of 0.95 for all BWR configurations, while staying above USL for the PWR ones. Therefore, the application of a burnup credit methodology for the PWR case is required, being currently under development. Relevant is also the influence of canister material and internal geometry on criticality, enabling the identification of safer fuel arrangements. For a final burnup of 55MWd/kgHM and 30y cooling time, the combined photon-neutron surface dose rate is well below the threshold of 1 Gy/h defined to limit radiation-induced corrosion of the canister in all cases. Copyright © 2018 Elsevier Ltd. All rights reserved.

Analysis of No-load Iron Losses of Turbine Generators by 3D Magnetic Field Analysis

NASA Astrophysics Data System (ADS)

Nakahara, Akihito; Mogi, Hisashi; Takahashi, Kazuhiko; Ide, Kazumasa; Kaneda, Junya; Hattori, Ken'Ichi; Watanabe, Takashi; Kaido, Chikara; Minematsu, Eisuke; Hanzawa, Kazufumi

This paper focuses on no-load iron losses of turbine generators. To calculate iron losses of turbine generators a program was developed. In the program, core loss curves of materials used for stator core were reproduced precisely by using tables of loss coefficients. Accuracy of calculation by this method was confirmed by comparing calculated values with measured in a model stator core. The iron loss of a turbine generator estimated with considering three-dimensional distribution of magnetic fluxes. And additional losses included in measured iron loss was evaluated with three-dimensional magnetic field analysis.

Co-axial Electrospun Polyacrylonitrile-Poly(methylmethacrylate) Nanofibers: Atomic Force Microscopy and Compositional Characterization

PubMed Central

Zander, N.E.; Strawhecker, K.E.; Orlicki, J.A.; Rawlett, A.M.; Beebe, T.P.

2011-01-01

Poly(methylmethacrylate) (PMMA)- Polyacrylonitrile (PAN) fibers were prepared using a conventional single-nozzle electrospinning technique. The as-spun fibers exhibited core-shell morphology as verified by transmission electron microscopy (TEM) and atomic force microscopy (AFM). AFM-phase and modulus mapping images of the fiber cross-section and x-ray photoelectron spectroscopy (XPS) analysis indicated PAN formed the shell and PMMA the core material. XPS, thermal gravimetric analysis (TGA), and elemental analysis were used to determine fiber compositional information. Soaking the fibers in solvent demonstrated removal of the core material, generating hollow PAN fibers. PMID:21928836

The study on the core personality trait words of Chinese medical university students based on social network analysis

PubMed Central

Wu, Ying; Xue, Yunzhen; Xue, Zhanling

2017-01-01

Abstract The medical university students in China whose school work is relatively heavy and educational system is long are a special professional group. Many students have psychological problems more or less. So, to understand their personality characteristics will provide a scientific basis for the intervention of psychological health. We selected top 30 personality trait words according to the order of frequency. Additionally, some methods such as social network analysis (SNA) and visualization technology of mapping knowledge domain were used in this study. Among these core personality trait words Family conscious had the 3 highest centralities and possessed the largest core status and influence. From the analysis of core-peripheral structure, we can see polarized core-perpheral structure was quite obvious. From the analysis of K-plex, there were in total 588 “K-2”K-plexs. From the analysis of Principal Components, we selected the 11 principal components. This study of personality not only can prevent disease, but also provide a scientific basis for students’ psychological healthy education. In addition, we have adopted SNA to pay more attention to the relationship between personality trait words and the connection among personality dimensions. This study may provide the new ideas and methods for the research of personality structure. PMID:28906409

The study on the core personality trait words of Chinese medical university students based on social network analysis.

PubMed

Wu, Ying; Xue, Yunzhen; Xue, Zhanling

2017-09-01

The medical university students in China whose school work is relatively heavy and educational system is long are a special professional group. Many students have psychological problems more or less. So, to understand their personality characteristics will provide a scientific basis for the intervention of psychological health.We selected top 30 personality trait words according to the order of frequency. Additionally, some methods such as social network analysis (SNA) and visualization technology of mapping knowledge domain were used in this study.Among these core personality trait words Family conscious had the 3 highest centralities and possessed the largest core status and influence. From the analysis of core-peripheral structure, we can see polarized core-perpheral structure was quite obvious. From the analysis of K-plex, there were in total 588 "K-2"K-plexs. From the analysis of Principal Components, we selected the 11 principal components.This study of personality not only can prevent disease, but also provide a scientific basis for students' psychological healthy education. In addition, we have adopted SNA to pay more attention to the relationship between personality trait words and the connection among personality dimensions. This study may provide the new ideas and methods for the research of personality structure.

A Text Corpus Approach to an Analysis of the Shared Use of Core Terminology.

ERIC Educational Resources Information Center

Patrick, Timothy B.; Sievert, MaryEllen; Reid, John C.; Rice, Frances Ellis; Gigantelli, James W.; Schiffman, Jade S.; Shelton, Mark E.

2003-01-01

Investigates the shared use of core Ophthalmology terms in the domains of Ophthalmology, Family Practice and Radiology. Core terms were searched for in a text corpus of 38,695 MEDLINE abstracts covering 1970-1999 from journals representing the three domains. Findings indicated core Ophthalmology terms were used significantly more by Ophthalmology…

Cores Of Recurrent Events (CORE) | Informatics Technology for Cancer Research (ITCR)

Cancer.gov

CORE is a statistically supported computational method for finding recurrently targeted regions in massive collections of genomic intervals, such as those arising from DNA copy number analysis of single tumor cells or bulk tumor tissues.

Unravelling the Paleoenvironmental and Diagenetic History of Fluviolacustrine Sediments from a Northern Kenya Rift Basin Through Analysis of HSPDP West Turkana-Kaitio Core Material

NASA Astrophysics Data System (ADS)

Rabideaux, N. M.; Chaudhary, M. S.; Deocampo, D.; Feibel, C. S.; Cohen, A. S.

2016-12-01

The Hominin Sites and Paleolakes Drilling Project (HSPDP) collected sediment cores from six rift basins in Ethiopia and Kenya. The goal of HSPDP is to construct high-resolution records of environmental change, and to understand how those changes relate to early human evolution and cultural adaptations. The West Turkana-Kaitio (WTK) site was targeted due to the abundant archeological and paleontological artifacts and fossils discovered around the basin. We conducted XRD and XRF analyses on HSPDP-WTK core material to construct a high-resolution record of paleoenvironmental conditions in the Kenya Rift during the Early Pleistocene ( 1.9-1.35 Ma). Mineralogical and geochemical trends were also used to identify the diagenetic history of fluviolacustrine sediments in the basin. The bulk mineralogy is comprised of mostly detrital feldspars, muscovite, α-quartz, and carbonates. Zeolites are present in intervals throughout the core, possibly suggesting pulses of increased salinity. Oxides and S-bearing minerals are abundant from 100-170 mbs, which may be indicative of redox and or hydrothermal processes in that interval. The lowermost portion of the core contains α- and β-quartz, pyrite and zeolites, suggesting either low-oxygen saline conditions or hydrothermal activity. Oriented clay analysis indicated multiple intervals of diagenesis, with the illitization of smectite related to hydrothermal and or microbial activity. Clay analysis provided evidence for a low degree of illitization in the upper portion of the core, whereas mixed-layered illite-smectite (I/S) contained 30-50% illite proximal to fault breccia and up to 70% illite below the faulted section, indicative of significant alteration in the lowermost portion of the core. Coupled mineralogical and geochemical analysis revealed a complex alteration history in the basin indicated by: 1) the presence of mixed-layer I/S throughout the 216 m core; 2) pronounced alteration proximal to faulting; and 3) authigenic silicates and pyrite in the basal section of the core.

Core Needle Lung Biopsy Specimens: Adequacy for EGFR and KRAS Mutational Analysis

PubMed Central

Zakowski, Maureen F.; Pao, William; Thornton, Raymond H.; Ladanyi, Marc; Kris, Mark G.; Rusch, Valerie W.; Rizvi, Naiyer A.

2013-01-01

OBJECTIVE The purpose of this study was to prospectively compare the adequacy of core needle biopsy specimens with the adequacy of specimens from resected tissue, the histologic reference standard, for mutational analysis of malignant tumors of the lung. SUBJECTS AND METHODS The first 18 patients enrolled in a phase 2 study of gefitinib for lung cancer in July 2004 through August 2005 underwent CT- or fluoroscopy-guided lung biopsy before the start of gefitinib therapy. Three weeks after gefitinib therapy, the patients underwent lung tumor resection. The results of EGFR and KRAS mutational analysis of the core needle biopsy specimens were compared with those of EGFR and KRAS mutational analysis of the surgical specimens. RESULTS Two specimens were unsatisfactory for mutational analysis. The results of mutational assay results of the other 16 specimens were the same as those of analysis of the surgical specimens obtained an average of 31 days after biopsy. CONCLUSION Biopsy with small (18- to 20-gauge) core needles can yield sufficient and reliable samples for mutational analysis. This technique is likely to become an important tool with the increasing use of pharmacotherapy based on the genetics of specific tumors in individual patients. PMID:20028932

Development of Safety Analysis Code System of Beam Transport and Core for Accelerator Driven System

NASA Astrophysics Data System (ADS)

Aizawa, Naoto; Iwasaki, Tomohiko

2014-06-01

Safety analysis code system of beam transport and core for accelerator driven system (ADS) is developed for the analyses of beam transients such as the change of the shape and position of incident beam. The code system consists of the beam transport analysis part and the core analysis part. TRACE 3-D is employed in the beam transport analysis part, and the shape and incident position of beam at the target are calculated. In the core analysis part, the neutronics, thermo-hydraulics and cladding failure analyses are performed by the use of ADS dynamic calculation code ADSE on the basis of the external source database calculated by PHITS and the cross section database calculated by SRAC, and the programs of the cladding failure analysis for thermoelastic and creep. By the use of the code system, beam transient analyses are performed for the ADS proposed by Japan Atomic Energy Agency. As a result, the rapid increase of the cladding temperature happens and the plastic deformation is caused in several seconds. In addition, the cladding is evaluated to be failed by creep within a hundred seconds. These results have shown that the beam transients have caused a cladding failure.

Multivariate analysis of heavy metal contamination using river sediment cores of Nankan River, northern Taiwan

NASA Astrophysics Data System (ADS)

Lee, An-Sheng; Lu, Wei-Li; Huang, Jyh-Jaan; Chang, Queenie; Wei, Kuo-Yen; Lin, Chin-Jung; Liou, Sofia Ya Hsuan

2016-04-01

Through the geology and climate characteristic in Taiwan, generally rivers carry a lot of suspended particles. After these particles settled, they become sediments which are good sorbent for heavy metals in river system. Consequently, sediments can be found recording contamination footprint at low flow energy region, such as estuary. Seven sediment cores were collected along Nankan River, northern Taiwan, which is seriously contaminated by factory, household and agriculture input. Physico-chemical properties of these cores were derived from Itrax-XRF Core Scanner and grain size analysis. In order to interpret these complex data matrices, the multivariate statistical techniques (cluster analysis, factor analysis and discriminant analysis) were introduced to this study. Through the statistical determination, the result indicates four types of sediment. One of them represents contamination event which shows high concentration of Cu, Zn, Pb, Ni and Fe, and low concentration of Si and Zr. Furthermore, three possible contamination sources of this type of sediment were revealed by Factor Analysis. The combination of sediment analysis and multivariate statistical techniques used provides new insights into the contamination depositional history of Nankan River and could be similarly applied to other river systems to determine the scale of anthropogenic contamination.

2016 American College of Rheumatology/European League Against Rheumatism Criteria for Minimal, Moderate, and Major Clinical Response in Adult Dermatomyositis and Polymyositis: An International Myositis Assessment and Clinical Studies Group/Paediatric Rheumatology International Trials Organisation Collaborative Initiative.

PubMed

Aggarwal, Rohit; Rider, Lisa G; Ruperto, Nicolino; Bayat, Nastaran; Erman, Brian; Feldman, Brian M; Oddis, Chester V; Amato, Anthony A; Chinoy, Hector; Cooper, Robert G; Dastmalchi, Maryam; Fiorentino, David; Isenberg, David; Katz, James D; Mammen, Andrew; de Visser, Marianne; Ytterberg, Steven R; Lundberg, Ingrid E; Chung, Lorinda; Danko, Katalin; García-De la Torre, Ignacio; Song, Yeong Wook; Villa, Luca; Rinaldi, Mariangela; Rockette, Howard; Lachenbruch, Peter A; Miller, Frederick W; Vencovsky, Jiri

2017-05-01

To develop response criteria for adult dermatomyositis (DM) and polymyositis (PM). Expert surveys, logistic regression, and conjoint analysis were used to develop 287 definitions using core set measures. Myositis experts rated greater improvement among multiple pairwise scenarios in conjoint analysis surveys, where different levels of improvement in 2 core set measures were presented. The PAPRIKA (Potentially All Pairwise Rankings of All Possible Alternatives) method determined the relative weights of core set measures and conjoint analysis definitions. The performance characteristics of the definitions were evaluated on patient profiles using expert consensus (gold standard) and were validated using data from a clinical trial. The nominal group technique was used to reach consensus. Consensus was reached for a conjoint analysis-based continuous model using absolute percent change in core set measures (physician, patient, and extramuscular global activity, muscle strength, Health Assessment Questionnaire, and muscle enzyme levels). A total improvement score (range 0-100), determined by summing scores for each core set measure, was based on improvement in and relative weight of each core set measure. Thresholds for minimal, moderate, and major improvement were ≥20, ≥40, and ≥60 points in the total improvement score. The same criteria were chosen for juvenile DM, with different improvement thresholds. Sensitivity and specificity in DM/PM patient cohorts were 85% and 92%, 90% and 96%, and 92% and 98% for minimal, moderate, and major improvement, respectively. Definitions were validated in the clinical trial analysis for differentiating the physician rating of improvement (P < 0.001). The response criteria for adult DM/PM consisted of the conjoint analysis model based on absolute percent change in 6 core set measures, with thresholds for minimal, moderate, and major improvement. © 2017, American College of Rheumatology.

Automated podosome identification and characterization in fluorescence microscopy images.

PubMed

Meddens, Marjolein B M; Rieger, Bernd; Figdor, Carl G; Cambi, Alessandra; van den Dries, Koen

2013-02-01

Podosomes are cellular adhesion structures involved in matrix degradation and invasion that comprise an actin core and a ring of cytoskeletal adaptor proteins. They are most often identified by staining with phalloidin, which binds F-actin and therefore visualizes the core. However, not only podosomes, but also many other cytoskeletal structures contain actin, which makes podosome segmentation by automated image processing difficult. Here, we have developed a quantitative image analysis algorithm that is optimized to identify podosome cores within a typical sample stained with phalloidin. By sequential local and global thresholding, our analysis identifies up to 76% of podosome cores excluding other F-actin-based structures. Based on the overlap in podosome identifications and quantification of podosome numbers, our algorithm performs equally well compared to three experts. Using our algorithm we show effects of actin polymerization and myosin II inhibition on the actin intensity in both podosome core and associated actin network. Furthermore, by expanding the core segmentations, we reveal a previously unappreciated differential distribution of cytoskeletal adaptor proteins within the podosome ring. These applications illustrate that our algorithm is a valuable tool for rapid and accurate large-scale analysis of podosomes to increase our understanding of these characteristic adhesion structures.

A motional Stark effect diagnostic analysis routine for improved resolution of iota in the core of the large helical device.

PubMed

Dobbins, T J; Ida, K; Suzuki, C; Yoshinuma, M; Kobayashi, T; Suzuki, Y; Yoshida, M

2017-09-01

A new Motional Stark Effect (MSE) analysis routine has been developed for improved spatial resolution in the core of the Large Helical Device (LHD). The routine was developed to reduce the dependency of the analysis on the Pfirsch-Schlüter (PS) current in the core. The technique used the change in the polarization angle as a function of flux in order to find the value of diota/dflux at each measurement location. By integrating inwards from the edge, the iota profile can be recovered from this method. This reduces the results' dependency on the PS current because the effect of the PS current on the MSE measurement is almost constant as a function of flux in the core; therefore, the uncertainty in the PS current has a minimal effect on the calculation of the iota profile. In addition, the VMEC database was remapped from flux into r/a space by interpolating in mode space in order to improve the database core resolution. These changes resulted in a much smoother iota profile, conforming more to the physics expectations of standard discharge scenarios in the core of the LHD.

The WAIS Melt Monitor: An automated ice core melting system for meltwater sample handling and the collection of high resolution microparticle size distribution data

NASA Astrophysics Data System (ADS)

Breton, D. J.; Koffman, B. G.; Kreutz, K. J.; Hamilton, G. S.

2010-12-01

Paleoclimate data are often extracted from ice cores by careful geochemical analysis of meltwater samples. The analysis of the microparticles found in ice cores can also yield unique clues about atmospheric dust loading and transport, dust provenance and past environmental conditions. Determination of microparticle concentration, size distribution and chemical makeup as a function of depth is especially difficult because the particle size measurement either consumes or contaminates the meltwater, preventing further geochemical analysis. Here we describe a microcontroller-based ice core melting system which allows the collection of separate microparticle and chemistry samples from the same depth intervals in the ice core, while logging and accurately depth-tagging real-time electrical conductivity and particle size distribution data. This system was designed specifically to support microparticle analysis of the WAIS Divide WDC06A deep ice core, but many of the subsystems are applicable to more general ice core melting operations. Major system components include: a rotary encoder to measure ice core melt displacement with 0.1 millimeter accuracy, a meltwater tracking system to assign core depths to conductivity, particle and sample vial data, an optical debubbler level control system to protect the Abakus laser particle counter from damage due to air bubbles, a Rabbit 3700 microcontroller which communicates with a host PC, collects encoder and optical sensor data and autonomously operates Gilson peristaltic pumps and fraction collectors to provide automatic sample handling, melt monitor control software operating on a standard PC allowing the user to control and view the status of the system, data logging software operating on the same PC to collect data from the melting, electrical conductivity and microparticle measurement systems. Because microparticle samples can easily be contaminated, we use optical air bubble sensors and high resolution ice core density profiles to guide the melting process. The combination of these data allow us to analyze melt head performance, minimize outer-to-inner fraction contamination and avoid melt head flooding. The WAIS Melt Monitor system allows the collection of real-time, sub-annual microparticle and electrical conductivity data while producing and storing enough sample for traditional Coulter-Counter particle measurements as well long term acid leaching of bioactive metals (e.g., Fe, Co, Cd, Cu, Zn) prior to chemical analysis.

A Two-Step Approach to Uncertainty Quantification of Core Simulators

DOE PAGES

Yankov, Artem; Collins, Benjamin; Klein, Markus; ...

2012-01-01

For the multiple sources of error introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily to quantify the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the “two-step” method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the two-step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor andmore » in the core power distribution was examined in the framework of phase I-3 of the OECD Uncertainty Analysis in Modeling benchmark. With the Three Mile Island Unit 1 core as a base model for analysis, the XSUSA and two-step methods were applied with certain limitations, and the results were compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions were drawn as to the method that is currently more viable for computing uncertainties in burnup and transient calculations.« less

Publications - GMC 41 | Alaska Division of Geological & Geophysical Surveys

Science.gov Websites

DGGS GMC 41 Publication Details Title: X-ray diffraction clay mineralogy analysis of core samples from Unknown, [n.d.], X-ray diffraction clay mineralogy analysis of core samples from Mobil West Staines State

DOE Office of Scientific and Technical Information (OSTI.GOV)

Muljadi, Eduard; Hasan, Iftekhar; Husain, Tausif

This research examines the vibration and thermal characteristics of double-sided flux concentrating Transverse Flux Machines (TFM), designed for direct drive application. Two TFM prototypes with different stator cores, one with Quasi U-Core and the other with E-Core, has been used for the study. 3D Finite Element Analysis (FEA) has been carried out to determine the no-load and with load performance of the TFMs along with their fluctuating axial electromagnetic force densities acting on the stator teeth. The deformation response of the stator cores was observed in the static structural analysis. Thermal analysis for the TFM was performed through FEA basedmore » on copper and iron losses in the machine to examine the temperature rise in different parts of the machine structure. Acceleration and noise measurements were experimentally obtained to characterize the vibrational performance of the prototypes.« less

Response of Honeycomb Core Sandwich Panel with Minimum Gage GFRP Face-Sheets to Compression Loading After Impact

NASA Technical Reports Server (NTRS)

McQuigg, Thomas D.; Kapania, Rakesh K.; Scotti, Stephen J.; Walker, Sandra P.

2011-01-01

A compression after impact study has been conducted to determine the residual strength of three sandwich panel constructions with two types of thin glass fiber reinforced polymer face-sheets and two hexagonal honeycomb Nomex core densities. Impact testing is conducted to first determine the characteristics of damage resulting from various impact energy levels. Two modes of failure are found during compression after impact tests with the density of the core precipitating the failure mode present for a given specimen. A finite element analysis is presented for prediction of the residual compressive strength of the impacted specimens. The analysis includes progressive damage modeling in the face-sheets. Preliminary analysis results were similar to the experimental results; however, a higher fidelity core material model is expected to improve the correlation.

Preliminary posttest analysis of LOFT loss-of-coolant experiment L2-2

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, J.R.; Grush, W.H.; Keeler, C.D.

A preliminary posttest analysis of Loss-of-Coolant Experiment (LOCE) L2-2, which was conducted in the Loss-of-Fluid Test (LOFT) facility, was performed to gain an understanding of the cause of the disparity between predicted and measured fuel rod cladding temperature responses in the LOFT core. LOCE L2-2 is the first experiment in the LOFT Power Ascension Series L2 (first series of LOFT nuclear experiments), which was designed to investigate the response of the LOFT nuclear core to the blowdown, refill, and reflood transients during LOCEs conducted at gradually increasing power levels. LOCE L2-2 was conducted at 50% power (25 MW, 26.38 kW/m).more » Results show that a core-wide rewet occurred early in the transient (during blowdown starting at about 7 s after rupture) which was not calculated in the pretest prediction analysis. This early core-wide rewet resulted in the peak fuel rod cladding temperatures being lower (by a mean value of 166/sup 0/K for 24 thermocouples) than had been calculated. This preliminary posttest analysis was concerned solely with determining why the early core-wide rewet was not predicted by the RELAP4/MOD6 pretest analysis and be no means is it a complete posttest analysis of LOCE L2-2 results. However, during this analysis, several errors made in the prettest analysis were found, and their impact on the predicted results is assessed. Three factors were postulated to have caused the disparity between predicted and measured fuel rod cladding temperatures for LOCE L2-2: (a) the initial fuel rod stored energy, (b) the heat transfer surface, and (c) the hydraulics calculation. These factors were examined and are discussed in this report. It was determined that core hydraulics, as influenced by the calculation of broken loop cold leg break flow, was the major factor causing the disparity.« less

Selenium semiconductor core optical fibers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tang, G. W.; Qian, Q., E-mail: qianqi@scut.edu.cn; Peng, K. L.

2015-02-15

Phosphate glass-clad optical fibers containing selenium (Se) semiconductor core were fabricated using a molten core method. The cores were found to be amorphous as evidenced by X-ray diffraction and corroborated by Micro-Raman spectrum. Elemental analysis across the core/clad interface suggests that there is some diffusion of about 3 wt % oxygen in the core region. Phosphate glass-clad crystalline selenium core optical fibers were obtained by a postdrawing annealing process. A two-cm-long crystalline selenium semiconductor core optical fibers, electrically contacted to external circuitry through the fiber end facets, exhibit a three times change in conductivity between dark and illuminated states. Suchmore » crystalline selenium semiconductor core optical fibers have promising utility in optical switch and photoconductivity of optical fiber array.« less

Impact of the dynamical core on the direct simulation of tropical cyclones in a high-resolution global model

DOE PAGES

Reed, K. A.; Bacmeister, J. T.; Rosenbloom, N. A.; ...

2015-05-13

Our paper examines the impact of the dynamical core on the simulation of tropical cyclone (TC) frequency, distribution, and intensity. The dynamical core, the central fluid flow component of any general circulation model (GCM), is often overlooked in the analysis of a model's ability to simulate TCs compared to the impact of more commonly documented components (e.g., physical parameterizations). The Community Atmosphere Model version 5 is configured with multiple dynamics packages. This analysis demonstrates that the dynamical core has a significant impact on storm intensity and frequency, even in the presence of similar large-scale environments. In particular, the spectral elementmore » core produces stronger TCs and more hurricanes than the finite-volume core using very similar parameterization packages despite the latter having a slightly more favorable TC environment. Furthermore, these results suggest that more detailed investigations into the impact of the GCM dynamical core on TC climatology are needed to fully understand these uncertainties. Key Points The impact of the GCM dynamical core is often overlooked in TC assessments The CAM5 dynamical core has a significant impact on TC frequency and intensity A larger effort is needed to better understand this uncertainty« less

Construct and Predictive Validity of the Core Phonics Survey: A Diagnostic Assessment for Students with Specific Learning Disabilities

ERIC Educational Resources Information Center

Park, Yujeong; Benedict, Amber E.; Brownell, Mary T.

2014-01-01

The factor structure of the CORE Phonics Survey was analyzed using a sample of 165 students in upper elementary school with specific learning disabilities. Confirmatory factor analysis was used to identify the hypothesized constructs of the CORE Phonics Survey and predictive validity of the CORE Phonics Survey to predict students' success in word…

Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

NASA Technical Reports Server (NTRS)

Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

1972-01-01

A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

Optimal design for crosstalk analysis in 12-core 5-LP mode homogeneous multicore fiber for different lattice structure

NASA Astrophysics Data System (ADS)

Kumar, Dablu; Ranjan, Rakesh

2018-03-01

12-Core 5-LP mode homogeneous multicore fibers have been proposed for analysis of inter-core crosstalk and dispersion, with four different lattice structures (circular, 2-ring, square lattice, and triangular lattice) having cladding diameter of 200 μm and a fixed cladding thickness of 35 μm. The core-to-core crosstalk impact has been studied numerically with respect to bending radius, core pitch, transmission distance, wavelength, and core diameter for all 5-LP modes. In anticipation of further reduction in crosstalk levels, the trench-assisted cores have been incorporated for all respective designs. Ultra-low crosstalk (-138 dB/100 km) has been achieved through the triangular lattice arrangement, with trench depth Δ2 = -1.40% for fundamental (LP01) mode. It has been noted that the impact of mode polarization on crosstalk behavior is minor, with difference in crosstalk levels between two polarized spatial modes as ≤0.2 dB. Moreover, the optimized cladding diameter has been obtained for all 5-LP modes for a target value of crosstalk of -50 dB/100 km, with all the core arrangements. The dispersion characteristic has also been analyzed with respect to wavelength, which is nearly 2.5 ps/nm km at operating wavelength 1550 nm. The relative core multiplicity factor (RCMF) for the proposed design is obtained as 64.

Task related doses in Spanish pressurized water reactors over the period 1988-1992

DOE Office of Scientific and Technical Information (OSTI.GOV)

O`Donnell, P.; Labarta, T.; Amor, I.

1995-03-01

In order to evaluate in depth the collective dose trend and its correlation with the effectiveness of the practical application of the ALARA principle in Spanish nuclear facilities, and base the different policy lines to promote this criteria, the CSN has fullfilled an analysis of the task related doses data over the period 1988-1992. Previously, the CSN had required to the utilities the compilation of their refuelling outage collective dose from 1988 according with a predeterminate number of tasks, in order to have available a representative and retrospective set of data in an homogeneous way and coherent with the internationalmore » data banks on occupational exposure in NPP, as the CEC and the NEA ones. The scope of this analysis was the following: first, the collective dose summaries for outage tasks and departments for PWR and for BWR, including the minimum, maximum and average dose (and statistics data) for 18 different refuelling outage tasks and 12 personal departments for each generation of each type of rector, the task and department related collective dose trends in each plant and in each generation, and second, the dose reduction techniques having been used during that period in each plant and the relative level of adoption. In this presentation the main results and conclusions of the first part of the study are reviewed for PWR.« less

Exploring cosmic origins with CORE: Mitigation of systematic effects

NASA Astrophysics Data System (ADS)

Natoli, P.; Ashdown, M.; Banerji, R.; Borrill, J.; Buzzelli, A.; de Gasperis, G.; Delabrouille, J.; Hivon, E.; Molinari, D.; Patanchon, G.; Polastri, L.; Tomasi, M.; Bouchet, F. R.; Henrot-Versillé, S.; Hoang, D. T.; Keskitalo, R.; Kiiveri, K.; Kisner, T.; Lindholm, V.; McCarthy, D.; Piacentini, F.; Perdereau, O.; Polenta, G.; Tristram, M.; Achucarro, A.; Ade, P.; Allison, R.; Baccigalupi, C.; Ballardini, M.; Banday, A. J.; Bartlett, J.; Bartolo, N.; Basak, S.; Baumann, D.; Bersanelli, M.; Bonaldi, A.; Bonato, M.; Boulanger, F.; Brinckmann, T.; Bucher, M.; Burigana, C.; Cai, Z.-Y.; Calvo, M.; Carvalho, C.-S.; Castellano, M. G.; Challinor, A.; Chluba, J.; Clesse, S.; Colantoni, I.; Coppolecchia, A.; Crook, M.; D'Alessandro, G.; de Bernardis, P.; De Zotti, G.; Di Valentino, E.; Diego, J.-M.; Errard, J.; Feeney, S.; Fernandez-Cobos, R.; Finelli, F.; Forastieri, F.; Galli, S.; Genova-Santos, R.; Gerbino, M.; González-Nuevo, J.; Grandis, S.; Greenslade, J.; Gruppuso, A.; Hagstotz, S.; Hanany, S.; Handley, W.; Hernandez-Monteagudo, C.; Hervías-Caimapo, C.; Hills, M.; Keihänen, E.; Kitching, T.; Kunz, M.; Kurki-Suonio, H.; Lamagna, L.; Lasenby, A.; Lattanzi, M.; Lesgourgues, J.; Lewis, A.; Liguori, M.; López-Caniego, M.; Luzzi, G.; Maffei, B.; Mandolesi, N.; Martinez-González, E.; Martins, C. J. A. P.; Masi, S.; Matarrese, S.; Melchiorri, A.; Melin, J.-B.; Migliaccio, M.; Monfardini, A.; Negrello, M.; Notari, A.; Pagano, L.; Paiella, A.; Paoletti, D.; Piat, M.; Pisano, G.; Pollo, A.; Poulin, V.; Quartin, M.; Remazeilles, M.; Roman, M.; Rossi, G.; Rubino-Martin, J.-A.; Salvati, L.; Signorelli, G.; Tartari, A.; Tramonte, D.; Trappe, N.; Trombetti, T.; Tucker, C.; Valiviita, J.; Van de Weijgaert, R.; van Tent, B.; Vennin, V.; Vielva, P.; Vittorio, N.; Wallis, C.; Young, K.; Zannoni, M.

2018-04-01

We present an analysis of the main systematic effects that could impact the measurement of CMB polarization with the proposed CORE space mission. We employ timeline-to-map simulations to verify that the CORE instrumental set-up and scanning strategy allow us to measure sky polarization to a level of accuracy adequate to the mission science goals. We also show how the CORE observations can be processed to mitigate the level of contamination by potentially worrying systematics, including intensity-to-polarization leakage due to bandpass mismatch, asymmetric main beams, pointing errors and correlated noise. We use analysis techniques that are well validated on data from current missions such as Planck to demonstrate how the residual contamination of the measurements by these effects can be brought to a level low enough not to hamper the scientific capability of the mission, nor significantly increase the overall error budget. We also present a prototype of the CORE photometric calibration pipeline, based on that used for Planck, and discuss its robustness to systematics, showing how CORE can achieve its calibration requirements. While a fine-grained assessment of the impact of systematics requires a level of knowledge of the system that can only be achieved in a future study phase, the analysis presented here strongly suggests that the main areas of concern for the CORE mission can be addressed using existing knowledge, techniques and algorithms.

Variation of heavy metals in recent sediments from Piratininga Lagoon (Brazil): interpretation of geochemical data with the aid of multivariate analysis

NASA Astrophysics Data System (ADS)

Huang, W.; Campredon, R.; Abrao, J. J.; Bernat, M.; Latouche, C.

1994-06-01

In the last decade, the Atlantic coast of south-eastern Brazil has been affected by increasing deforestation and anthropogenic effluents. Sediments in the coastal lagoons have recorded the process of such environmental change. Thirty-seven sediment samples from three cores in Piratininga Lagoon, Rio de Janeiro, were analyzed for their major components and minor element concentrations in order to examine geochemical characteristics and the depositional environment and to investigate the variation of heavy metals of environmental concern. Two multivariate analysis methods, principal component analysis and cluster analysis, were performed on the analytical data set to help visualize the sample clusters and the element associations. On the whole, the sediment samples from each core are similar and the sample clusters corresponding to the three cores are clearly separated, as a result of the different conditions of sedimentation. Some changes in the depositional environment are recognized using the results of multivariate analysis. The enrichment of Pb, Cu, and Zn in the upper parts of cores is in agreement with increasing anthropogenic influx (pollution).

ENANTIOMERIC RATIOS OF CHIRAL PCB ATROPISOMERS IN RADIODATED SEDIMENT CORES

EPA Science Inventory

Enantiomeric ratios (ERs)) of chiral polychlorinated biphenyl (PCB) atropisomers were quantified in radiodated sediment cores of Lake Hartwell SC, a reservoir heavily impacted by PCBS, to study spatial and temporal changes in chirality. A chiral analysis of cores showed accumulat...

Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Hongbin; Szilard, Ronaldo; Epiney, Aaron

Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance duringmore » LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.« less

Petrographic Analysis of Cores from Plant 42

DTIC Science & Technology

2016-10-01

ER D C TR -X X- D R AF T Petrographic Analysis of Cores from Plant 42 E n gi n ee r R es ea rc h a n d D ev el op m en t C en te r...E. Rae Reed-Gore, Kyle Klaus, and Robert D. Moser October 2016 Approved for public release; distribution is unlimited. The U.S. Army...ORIGINATOR. ii Figures and Tables Figures Figure 1. Test location map of AF Plant 42 with core number locations. ............................. 2

Lightweight Low Force Rotary Percussive Coring Tool for Planetary Applications

NASA Technical Reports Server (NTRS)

Hironaka, Ross; Stanley, Scott

2010-01-01

A prototype low-force rotary-percussive rock coring tool for use in acquiring samples for geological surveys in future planetary missions was developed. The coring tool could eventually enable a lightweight robotic system to operate from a relatively small (less than 200 kg) mobile or fixed platform to acquire and cache Mars or other planetary rock samples for eventual return to Earth for analysis. To gain insight needed to design an integrated coring tool, the coring ability of commercially available coring bits was evaluated for effectiveness of varying key parameters: weight-on-bit, rotation speed, percussive rate and force. Trade studies were performed for different methods of breaking a core at its base and for retaining the core in a sleeve to facilitate sample transfer. This led to a custom coring tool design which incorporated coring, core breakage, core retention, and core extraction functions. The coring tool was tested on several types of rock and demonstrated the overall feasibility of this approach for robotic rock sample acquisition.

Opportunities and Challenges of Linking Scientific Core Samples to the Geoscience Data Ecosystem

NASA Astrophysics Data System (ADS)

Noren, A. J.

2016-12-01

Core samples generated in scientific drilling and coring are critical for the advancement of the Earth Sciences. The scientific themes enabled by analysis of these samples are diverse, and include plate tectonics, ocean circulation, Earth-life system interactions (paleoclimate, paleobiology, paleoanthropology), Critical Zone processes, geothermal systems, deep biosphere, and many others, and substantial resources are invested in their collection and analysis. Linking core samples to researchers, datasets, publications, and funding agencies through registration of globally unique identifiers such as International Geo Sample Numbers (IGSNs) offers great potential for advancing several frontiers. These include maximizing sample discoverability, access, reuse, and return on investment; a means for credit to researchers; and documentation of project outputs to funding agencies. Thousands of kilometers of core samples and billions of derivative subsamples have been generated through thousands of investigators' projects, yet the vast majority of these samples are curated at only a small number of facilities. These numbers, combined with the substantial similarity in sample types, make core samples a compelling target for IGSN implementation. However, differences between core sample communities and other geoscience disciplines continue to create barriers to implementation. Core samples involve parent-child relationships spanning 8 or more generations, an exponential increase in sample numbers between levels in the hierarchy, concepts related to depth/position in the sample, requirements for associating data derived from core scanning and lithologic description with data derived from subsample analysis, and publications based on tens of thousands of co-registered scan data points and thousands of analyses of subsamples. These characteristics require specialized resources for accurate and consistent assignment of IGSNs, and a community of practice to establish norms, workflows, and infrastructure to support implementation.

A Methodology for Loading the Advanced Test Reactor Driver Core for Experiment Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cowherd, Wilson M.; Nielsen, Joseph W.; Choe, Dong O.

In support of experiments in the ATR, a new methodology was devised for loading the ATR Driver Core. This methodology will replace the existing methodology used by the INL Neutronic Analysis group to analyze experiments. Studied in this paper was the as-run analysis for ATR Cycle 152B, specifically comparing measured lobe powers and eigenvalue calculations.

Development of ECT/UT inspection system for bottom mounted instrumentation nozzle of PWR reactor vessels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tanaka, H.; Fukui, S.; Iwahashi, Y.

1994-12-31

The development of inspection technique and tool for Bottom Mounted Instrument (BMI) nozzle of PWR plant was performed for countermeasure of leakage accident at incore instrument nozzle of Hamaoka-1 (BWR). MHI achieved the following development, of which object was PWR Plant R/V: (1) development of ECT/UT Multi-sensored Probe; (2) development of Inspection System (3) development of Data Processing System. The Inspection System had been functionally tested using full scale mock-up. As the result of the functional test, this system was confirmed to be very effective, and assumed to be hopeful for the actual application on site.

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR). Data will be collected under simulated transportation environments using the cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). These data will be used to support ongoing SNF modeling activities and to address regulatory issues associated with SNF transport.

External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure – Scoping Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R.

This report documents a scoping assessment of a potential accident mitigation action applicable to the US fleet of boiling water reactors with Mark I and II containments. The mitigation action is to externally flood the primary containment vessel drywell head using portable pumps or other means. A scoping assessment of the potential benefits of this mitigation action was conducted focusing on the ability to (1) passively remove heat from containment, (2) prevent or delay leakage through the drywell head seal (due to high temperatures and/or pressure), and (3) scrub radionuclide releases if the drywell head seal leaks.

Current and anticipated uses of thermal-hydraulic codes in Germany

DOE Office of Scientific and Technical Information (OSTI.GOV)

Teschendorff, V.; Sommer, F.; Depisch, F.

1997-07-01

In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.

Us army corps of engineers - Engineering research and development center - Petrographic analysis of section 3 personnel tunnel concrete

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carter, J. M.

The Concrete and Materials Branch (CMB) of the Geotechnical and Structures Laboratory was requested to perform an analysis on concrete cores collected from the north and south walls of the H-Canyon Section 3 Personnel Tunel, Savannah River Site, Aiken, South Carolina to determine the cause of the lower than expected compressive strength. This study examined five cores provided to the ERDC by the Department of Energy. The cores were logged in as CMB No. 170051-1 to 170051-5 and subjected to petrographic examination, air void analysis, chemical sprays, scanning electron microscopy, and x-ray diffraction.

Identification of both copy number variation-type and constant-type core elements in a large segmental duplication region of the mouse genome

PubMed Central

2013-01-01

Background Copy number variation (CNV), an important source of diversity in genomic structure, is frequently found in clusters called CNV regions (CNVRs). CNVRs are strongly associated with segmental duplications (SDs), but the composition of these complex repetitive structures remains unclear. Results We conducted self-comparative-plot analysis of all mouse chromosomes using the high-speed and large-scale-homology search algorithm SHEAP. For eight chromosomes, we identified various types of large SD as tartan-checked patterns within the self-comparative plots. A complex arrangement of diagonal split lines in the self-comparative-plots indicated the presence of large homologous repetitive sequences. We focused on one SD on chromosome 13 (SD13M), and developed SHEPHERD, a stepwise ab initio method, to extract longer repetitive elements and to characterize repetitive structures in this region. Analysis using SHEPHERD showed the existence of 60 core elements, which were expected to be the basic units that form SDs within the repetitive structure of SD13M. The demonstration that sequences homologous to the core elements (>70% homology) covered approximately 90% of the SD13M region indicated that our method can characterize the repetitive structure of SD13M effectively. Core elements were composed largely of fragmented repeats of a previously identified type, such as long interspersed nuclear elements (LINEs), together with partial genic regions. Comparative genome hybridization array analysis showed that whereas 42 core elements were components of CNVR that varied among mouse strains, 8 did not vary among strains (constant type), and the status of the others could not be determined. The CNV-type core elements contained significantly larger proportions of long terminal repeat (LTR) types of retrotransposon than the constant-type core elements, which had no CNV. The higher divergence rates observed in the CNV-type core elements than in the constant type indicate that the CNV-type core elements have a longer evolutionary history than constant-type core elements in SD13M. Conclusions Our methodology for the identification of repetitive core sequences simplifies characterization of the structures of large SDs and detailed analysis of CNV. The results of detailed structural and quantitative analyses in this study might help to elucidate the biological role of one of the SDs on chromosome 13. PMID:23834397

Sub-core permeability and relative permeability characterization with Positron Emission Tomography

NASA Astrophysics Data System (ADS)

Zahasky, C.; Benson, S. M.

2017-12-01

This study utilizes preclinical micro-Positron Emission Tomography (PET) to image and quantify the transport behavior of pulses of a conservative aqueous radiotracer injected during single and multiphase flow experiments in a Berea sandstone core with axial parallel bedding heterogeneity. The core is discretized into streamtubes, and using the micro-PET data, expressions are derived from spatial moment analysis for calculating sub-core scale tracer flux and pore water velocity. Using the flux and velocity data, it is then possible to calculate porosity and saturation from volumetric flux balance, and calculate permeability and water relative permeability from Darcy's law. Full 3D simulations are then constructed based on this core characterization. Simulation results are compared with experimental results in order to test the assumptions of the simple streamtube model. Errors and limitations of this analysis will be discussed. These new methods of imaging and sub-core permeability and relative permeability measurements enable experimental quantification of transport behavior across scales.

Core ADHD Symptom Improvement with Atomoxetine versus Methylphenidate: A Direct Comparison Meta-Analysis

ERIC Educational Resources Information Center

Hazell, Philip L.; Kohn, Michael R.; Dickson, Ruth; Walton, Richard J.; Granger, Renee E.; van Wyk, Gregory W.

2011-01-01

Objective: Previous studies comparing atomoxetine and methylphenidate to treat ADHD symptoms have been equivocal. This noninferiority meta-analysis compared core ADHD symptom response between atomoxetine and methylphenidate in children and adolescents. Method: Selection criteria included randomized, controlled design; duration 6 weeks; and…

Textual Demands of Passages in Three English/Language Arts Common Core Assessments and One Core Literacy Program for Intermediate Elementary Grades

ERIC Educational Resources Information Center

Morrison, Timothy G.; Wilcox, Brad; Murdoch, Erica; Bird, Lauren

2018-01-01

The Common Core has emphasized reading for comprehension, including making inferences. However, little is known about the textual demands found within assessment and instructional passages that are promoted as being in line with Common Core expectations. The purpose of this content analysis was to identify the readability levels, passage length,…

The Information Systems Core: A Study from the Perspective of IS Core Curricula in the U.S.

ERIC Educational Resources Information Center

Hwang, Drew; Ma, Zhongming; Wang, Ming

2015-01-01

To keep up with technology changes and industry trends, it is essential for Information Systems (IS) programs to maintain up to date curricula. In doing so, IS educators need to determine what the IS core is and implement it in their curriculum. This study performed a descriptive analysis of 2,229 core courses offered by 394 undergraduate IS…

The Translational Genomics Core at Partners Personalized Medicine: Facilitating the Transition of Research towards Personalized Medicine

PubMed Central

Blau, Ashley; Brown, Alison; Mahanta, Lisa; Amr, Sami S.

2016-01-01

The Translational Genomics Core (TGC) at Partners Personalized Medicine (PPM) serves as a fee-for-service core laboratory for Partners Healthcare researchers, providing access to technology platforms and analysis pipelines for genomic, transcriptomic, and epigenomic research projects. The interaction of the TGC with various components of PPM provides it with a unique infrastructure that allows for greater IT and bioinformatics opportunities, such as sample tracking and data analysis. The following article describes some of the unique opportunities available to an academic research core operating within PPM, such the ability to develop analysis pipelines with a dedicated bioinformatics team and maintain a flexible Laboratory Information Management System (LIMS) with the support of an internal IT team, as well as the operational challenges encountered to respond to emerging technologies, diverse investigator needs, and high staff turnover. In addition, the implementation and operational role of the TGC in the Partners Biobank genotyping project of over 25,000 samples is presented as an example of core activities working with other components of PPM. PMID:26927185

The Translational Genomics Core at Partners Personalized Medicine: Facilitating the Transition of Research towards Personalized Medicine.

PubMed

Blau, Ashley; Brown, Alison; Mahanta, Lisa; Amr, Sami S

2016-02-26

The Translational Genomics Core (TGC) at Partners Personalized Medicine (PPM) serves as a fee-for-service core laboratory for Partners Healthcare researchers, providing access to technology platforms and analysis pipelines for genomic, transcriptomic, and epigenomic research projects. The interaction of the TGC with various components of PPM provides it with a unique infrastructure that allows for greater IT and bioinformatics opportunities, such as sample tracking and data analysis. The following article describes some of the unique opportunities available to an academic research core operating within PPM, such the ability to develop analysis pipelines with a dedicated bioinformatics team and maintain a flexible Laboratory Information Management System (LIMS) with the support of an internal IT team, as well as the operational challenges encountered to respond to emerging technologies, diverse investigator needs, and high staff turnover. In addition, the implementation and operational role of the TGC in the Partners Biobank genotyping project of over 25,000 samples is presented as an example of core activities working with other components of PPM.

Investigation of mechanical properties of hydrate-bearing pressure core sediments recovered from the Eastern Nankai Trough using transparent acrylic cell triaxial testing system (TACTT-system)

NASA Astrophysics Data System (ADS)

Yoneda, J.; Masui, A.; Konno, Y.; Jin, Y.; Kida, M.; Suzuki, K.; Nakatsuka, Y.; Tenma, N.; Nagao, J.

2014-12-01

Natural gas hydrate-bearing pressure core sediments have been sheared in compression using a newly developed Transparent Acrylic Cell Triaxial Testing (TACTT) system to investigate the geophysical and geomechanical behavior of sediments recovered from the deep seabed in the Eastern Nankai Trough, the first Japanese offshore production test region. The sediments were recovered by hybrid pressure core system (hybrid PCS) and pressure cores were cut by pressure core analysis tools (PCATs) on board. These pressure cores were transferred to the AIST Hokkaido centre and trimmed by pressure core non-destructive analysis tools (PNATs) for TACTT system which maintained the pressure and temperature conditions within the hydrate stability boundary, through the entire process of core handling from drilling to the end of laboratory testing. An image processing technique was used to capture the motion of sediment in a transparent acrylic cell, and digital photographs were obtained at every 0.1% of vertical strain during the test. Analysis of the optical images showed that sediments with 63% hydrate saturation exhibited brittle failure, although nonhydrate-bearing sediments exhibited ductile failure. In addition, the increase in shear strength with hydrate saturation increase of natural gas hydrate is in agreement with previous data from synthetic gas hydrate. This research was financially supported by the Research Consortium for Methane Hydrate Resources in Japan (MH21 Research Consortium) that carries out Japan's Methane Hydrate R&D Program by the Ministry of Economy, Trade and Industry (METI).

The subsurface geology of Río Tinto: material examined during a simulated Mars drilling mission for the Mars Astrobiology Research and Technology Experiment (MARTE).

PubMed

Prieto-Ballesteros, Olga; Martínez-Frías, Jesús; Schutt, John; Sutter, Brad; Heldmann, Jennifer L; Bell, Mary Sue; Battler, Melissa; Cannon, Howard; Gómez-Elvira, Javier; Stoker, Carol R

2008-10-01

The 2005 Mars Astrobiology Research and Technology Experiment (MARTE) project conducted a simulated 1-month Mars drilling mission in the Río Tinto district, Spain. Dry robotic drilling, core sampling, and biological and geological analytical technologies were collectively tested for the first time for potential use on Mars. Drilling and subsurface sampling and analytical technologies are being explored for Mars because the subsurface is the most likely place to find life on Mars. The objectives of this work are to describe drilling, sampling, and analytical procedures; present the geological analysis of core and borehole material; and examine lessons learned from the drilling simulation. Drilling occurred at an undisclosed location, causing the science team to rely only on mission data for geological and biological interpretations. Core and borehole imaging was used for micromorphological analysis of rock, targeting rock for biological analysis, and making decisions regarding the next day's drilling operations. Drilling reached 606 cm depth into poorly consolidated gossan that allowed only 35% of core recovery and contributed to borehole wall failure during drilling. Core material containing any indication of biology was sampled and analyzed in more detail for its confirmation. Despite the poorly consolidated nature of the subsurface gossan, dry drilling was able to retrieve useful core material for geological and biological analysis. Lessons learned from this drilling simulation can guide the development of dry drilling and subsurface geological and biological analytical technologies for future Mars drilling missions.

The Subsurface Geology of Río Tinto: Material Examined During a Simulated Mars Drilling Mission for the Mars Astrobiology Research and Technology Experiment (MARTE)

NASA Astrophysics Data System (ADS)

Prieto-Ballesteros, Olga; Martínez-Frías, Jesús; Schutt, John; Sutter, Brad; Heldmann, Jennifer L.; Bell Johnson, Mary Sue; Battler, Melissa; Cannon, Howard; Gómez-Elvira, Javier; Stoker, Carol R.

2008-10-01

The 2005 Mars Astrobiology Research and Technology Experiment (MARTE) project conducted a simulated 1-month Mars drilling mission in the Río Tinto district, Spain. Dry robotic drilling, core sampling, and biological and geological analytical technologies were collectively tested for the first time for potential use on Mars. Drilling and subsurface sampling and analytical technologies are being explored for Mars because the subsurface is the most likely place to find life on Mars. The objectives of this work are to describe drilling, sampling, and analytical procedures; present the geological analysis of core and borehole material; and examine lessons learned from the drilling simulation. Drilling occurred at an undis closed location, causing the science team to rely only on mission data for geological and biological interpretations. Core and borehole imaging was used for micromorphological analysis of rock, targeting rock for biological analysis, and making decisions regarding the next day's drilling operations. Drilling reached 606 cm depth into poorly consolidated gossan that allowed only 35% of core recovery and contributed to borehole wall failure during drilling. Core material containing any indication of biology was sampled and analyzed in more detail for its confirmation. Despite the poorly consolidated nature of the subsurface gossan, dry drilling was able to retrieve useful core material for geological and biological analysis. Lessons learned from this drilling simulation can guide the development of dry drilling and subsurface geological and biological analytical technologies for future Mars drilling missions.

Influence of a Fluid Lunar Core on the Moons Orientation

NASA Technical Reports Server (NTRS)

Williams, J. G.; Boggs, D. H.; Ratcliff, J. T.; Yoder, C. F.; Dickey, J. O.

2001-01-01

Oblateness of and dissipation at the lunar liquid-core/solid-mantle boundary affects the precession of core and mantle. Analysis of Lunar Laser ranges gives a weak detection of oblateness and a strong determination of dissipation. Additional information is contained in the original extended abstract.

Natural Resources. Ohio's Competency Analysis Profile. Forest Industry Worker. Resource Conservation.

ERIC Educational Resources Information Center

Ohio State Univ., Columbus. Vocational Instructional Materials Lab.

This competency analysis profile lists 155 competencies that have been identified by employers as core competencies for inclusion in programs to train forest industry and resource conservation workers. The core competencies are organized into 10 units dealing the following: general safety precautions, natural resource industry operations, soil…

Three-dimensional characterization of microporosity and permeability in fault zones hosted in heterolithic succession

NASA Astrophysics Data System (ADS)

Riegel, H. B.; Zambrano, M.; Jablonska, D.; Emanuele, T.; Agosta, F.; Mattioni, L.; Rustichelli, A.

2017-12-01

The hydraulic properties of fault zones depend upon the individual contributions of the damage zone and the fault core. In the case of the damage zone, it is generally characterized by means of fracture analysis and modelling implementing multiple approaches, for instance the discrete fracture network model, the continuum model, and the channel network model. Conversely, the fault core is more difficult to characterize because it is normally composed of fine grain material generated by friction and wear. If the dimensions of the fault core allows it, the porosity and permeability are normally studied by means of laboratory analysis or in the other case by two dimensional microporosity analysis and in situ measurements of permeability (e.g. micro-permeameter). In this study, a combined approach consisting of fracture modeling, three-dimensional microporosity analysis, and computational fluid dynamics was applied to characterize the hydraulic properties of fault zones. The studied fault zones crosscut a well-cemented heterolithic succession (sandstone and mudstones) and may vary in terms of fault core thickness and composition, fracture properties, kinematics (normal or strike-slip), and displacement. These characteristics produce various splay and fault core behavior. The alternation of sandstone and mudstone layers is responsible for the concurrent occurrence of brittle (fractures) and ductile (clay smearing) deformation. When these alternating layers are faulted, they produce corresponding fault cores which act as conduits or barriers for fluid migration. When analyzing damage zones, accurate field and data acquisition and stochastic modeling was used to determine the hydraulic properties of the rock volume, in relation to the surrounding, undamaged host rock. In the fault cores, the three-dimensional pore network quantitative analysis based on X-ray microtomography images includes porosity, pore connectivity, and specific surface area. In addition, images were used to perform computational fluid simulation (Lattice-Boltzmann multi relaxation time method) and estimate the permeability. These results will be useful for understanding the deformation process and hydraulic properties across meter-scale damage zones.

Field project to obtain pressure core, wireline log, and production test data for evaluation of CO/sub 2/ flooding potential, Conoco MCA unit well No. 358, Maljamar Field, Lea County, New Mexico

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swift, T.E.; Marlow, R.E.; Wilhelm, M.H.

1981-11-01

This report describes part of the work done to fulfill a contract awarded to Gruy Federal, Inc., by the Department of Energy (DOE) on Feburary 12, 1979. The work includes pressure-coring and associated logging and testing programs to provide data on in-situ oil saturation, porosity and permeability distribution, and other data needed for resource characterization of fields and reservoirs in which CO/sub 2/ injection might have a high probability of success. This report details the second such project. Core porosities agreed well with computed log porosities. Core water saturation and computed log porosities agree fairly well from 3692 to 3712more » feet, poorly from 3712 to 3820 feet and in a general way from 4035 to 4107 feet. Computer log analysis techniques incorporating the a, m, and n values obtained from Core Laboratories analysis did not improve the agreement of log versus core derived water saturations. However, both core and log analysis indicated the ninth zone had the highest residual hydrocarbon saturations and production data confirmed the validity of oil saturation determinations. Residual oil saturation, for the perforated and tested intervals were 259 STB/acre-ft for the interval from 4035 to 4055 feet, and 150 STB/acre-ft for the interval from 3692 to 3718 feet. Nine BOPD was produced from the interval 4035 to 4055 feet and no oil was produced from interval 3692 to 3718 feet, qualitatively confirming the relative oil saturations as calculated. The low oil production in the zone from 4022 to 4055 and the lack of production from 3692 to 3718 feet indicated the zone to be at or near residual waterflood conditions as determined by log analysis. This project demonstrates the usefulness of integrating pressure core, log, and production data to realistically evaluate a reservoir for carbon dioxide flood.« less

The Analytic Onion: Examining Training Issues from Different Levels of Analysis. Interim Technical Paper for Period July 1989-June 1991.

ERIC Educational Resources Information Center

Lamb, Theodore A.; Chin, Keric B. O.

This paper proposes a conceptual framework based on different levels of analysis using the metaphor of the layers of an onion to help organize and structure thinking on research issues concerning training. It discusses the core of the "analytic onion," the biological level, and seven levels of analysis that surround that core: the individual, the…

A full virial analysis of the prestellar cores in the Ophiuchus molecular cloud

NASA Astrophysics Data System (ADS)

Pattle, Kate; Ward-Thompson, Derek

2015-08-01

We present the first observations of the Ophiuchus molecular cloud performed as part of the James Clerk Maxwell (JCMT) Gould Belt Survey with the SCUBA-2 instrument. We demonstrate methods for combining these data with HARP CO, Herschel and IRAM N2H+ observations in order to accurately quantify the properties of the SCUBA-2 sources in Ophiuchus.We perform a full virial analysis on the starless cores in Ophiuchus, including external pressure. We find that the majority of our cores are either bound or virialised, and that gravity and external pressure are typically of similar importance in confining cores. We find that the critical Bonnor-Ebert stability criterion is not a good indicator of the boundedness of our cores. We determine that N2H+ is a good tracer of the bound material of prestellar cores, and find that non-thermal linewidths decrease substantially between the intermediate-density gas traced by C18O and the high-density gas traced by N2H+, indicating the dissipation of turbulence within cores.We find variation from region to region in the virial balance of cores and the relative contributions of pressure and gravity to core support, as well as variation in the degree to which turbulence is dissipated within cores and in the relative numbers of protostellar and starless sources. We find further support for our previous hypothesis of a global evolutionary gradient from southwest to northeast across Ophiuchus, indicating sequential star formation across the region.

Modelling of magnetostriction of transformer magnetic core for vibration analysis

NASA Astrophysics Data System (ADS)

Marks, Janis; Vitolina, Sandra

2017-12-01

Magnetostriction is a phenomenon occurring in transformer core in normal operation mode. Yet in time, it can cause the delamination of magnetic core resulting in higher level of vibrations that are measured on the surface of transformer tank during diagnostic tests. The aim of this paper is to create a model for evaluating elastic deformations in magnetic core that can be used for power transformers with intensive vibrations in order to eliminate magnetostriction as a their cause. Description of the developed model in Matlab and COMSOL software is provided including restrictions concerning geometry and properties of materials, and the results of performed research on magnetic core anisotropy are provided. As a case study modelling of magnetostriction for 5-legged 200 MVA power transformer with the rated voltage of 13.8/137kV is conducted, based on which comparative analysis of vibration levels and elastic deformations is performed.

Clinical features and ryanodine receptor type 1 gene mutation analysis in a Chinese family with central core disease.

PubMed

Chang, Xingzhi; Jin, Yiwen; Zhao, Haijuan; Huang, Qionghui; Wang, Jingmin; Yuan, Yun; Han, Ying; Qin, Jiong

2013-03-01

Central core disease is a rare inherited neuromuscular disorder caused by mutations in ryanodine receptor type 1 gene. The clinical phenotype of the disease is highly variable. We report a Chinese pedigree with central core disease confirmed by the gene sequencing. All 3 patients in the family presented with mild proximal limb weakness. The serum level of creatine kinase was normal, and electromyography suggested myogenic changes. The histologic analysis of muscle biopsy showed identical central core lesions in almost all of the muscle fibers in the index case. Exon 90-106 in the C-terminal domain of the ryanodine receptor type 1 gene was amplified using polymerase chain reaction. One heterozygous missense mutation G14678A (Arg4893Gln) in exon 102 was identified in all 3 patients. This is the first report of a familial case of central core disease confirmed by molecular study in mainland China.

Size-exclusion chromatography using core-shell particles.

PubMed

Pirok, Bob W J; Breuer, Pascal; Hoppe, Serafine J M; Chitty, Mike; Welch, Emmet; Farkas, Tivadar; van der Wal, Sjoerd; Peters, Ron; Schoenmakers, Peter J

2017-02-24

Size-exclusion chromatography (SEC) is an indispensable technique for the separation of high-molecular-weight analytes and for determining molar-mass distributions. The potential application of SEC as second-dimension separation in comprehensive two-dimensional liquid chromatography demands very short analysis times. Liquid chromatography benefits from the advent of highly efficient core-shell packing materials, but because of the reduced total pore volume these materials have so far not been explored in SEC. The feasibility of using core-shell particles in SEC has been investigated and contemporary core-shell materials were compared with conventional packing materials for SEC. Columns packed with very small core-shell particles showed excellent resolution in specific molar-mass ranges, depending on the pore size. The analysis times were about an order of magnitude shorter than what could be achieved using conventional SEC columns. Copyright © 2016 Elsevier B.V. All rights reserved.

The ADNI PET Core: 2015

PubMed Central

Jagust, William J.; Landau, Susan M.; Koeppe, Robert A.; Reiman, Eric M.; Chen, Kewei; Mathis, Chester A.; Price, Julie C.; Foster, Norman L.; Wang, Angela Y.

2015-01-01

INTRODUCTION This paper reviews the work done in the ADNI PET core over the past 5 years, largely concerning techniques, methods, and results related to amyloid imaging in ADNI. METHODS The PET Core has utilized [18F]florbetapir routinely on ADNI participants, with over 1600 scans available for download. Four different laboratories are involved in data analysis, and have examined factors such as longitudinal florbetapir analysis, use of FDG-PET in clinical trials, and relationships between different biomarkers and cognition. RESULTS Converging evidence from the PET Core has indicated that cross-sectional and longitudinal florbetapir analyses require different reference regions. Studies have also examined the relationship between florbetapir data obtained immediately after injection, which reflects perfusion, and FDG-PET results. Finally, standardization has included the translation of florbetapir PET data to a centiloid scale. CONCLUSION The PET Core has demonstrated a variety of methods for standardization of biomarkers such as florbetapir PET in a multicenter setting. PMID:26194311

Ultra low-loss hybrid core porous fiber for broadband applications.

PubMed

Islam, Md Saiful; Sultana, Jakeya; Atai, Javid; Abbott, Derek; Rana, Sohel; Islam, Mohammad Rakibul

2017-02-01

In this paper, we present the design and analysis of a novel hybrid porous core octagonal lattice photonic crystal fiber for terahertz (THz) wave guidance. The numerical analysis is performed using a full-vector finite element method (FEM) that shows that 80% of bulk absorption material loss of cyclic olefin copolymer (COC), commercially known as TOPAS can be reduced at a core diameter of 350 μm. The obtained effective material loss (EML) is as low as 0.04  cm^-1 at an operating frequency of 1 THz with a core porosity of 81%. Moreover, the proposed photonic crystal fiber also exhibits comparatively higher core power fraction, lower confinement loss, higher effective mode area, and an ultra-flattened dispersion profile with single mode propagation. This fiber can be readily fabricated using capillary stacking and sol-gel techniques, and it can be used for broadband terahertz applications.

The statistical analysis of circadian phase and amplitude in constant-routine core-temperature data

NASA Technical Reports Server (NTRS)

Brown, E. N.; Czeisler, C. A.

1992-01-01

Accurate estimation of the phases and amplitude of the endogenous circadian pacemaker from constant-routine core-temperature series is crucial for making inferences about the properties of the human biological clock from data collected under this protocol. This paper presents a set of statistical methods based on a harmonic-regression-plus-correlated-noise model for estimating the phases and the amplitude of the endogenous circadian pacemaker from constant-routine core-temperature data. The methods include a Bayesian Monte Carlo procedure for computing the uncertainty in these circadian functions. We illustrate the techniques with a detailed study of a single subject's core-temperature series and describe their relationship to other statistical methods for circadian data analysis. In our laboratory, these methods have been successfully used to analyze more than 300 constant routines and provide a highly reliable means of extracting phase and amplitude information from core-temperature data.

Multivariate analysis and geochemical approach for assessment of metal pollution state in sediment cores.

PubMed

Jamshidi-Zanjani, Ahmad; Saeedi, Mohsen

2017-07-01

Vertical distribution of metals (Cu, Zn, Cr, Fe, Mn, Pb, Ni, Cd, and Li) in four sediment core samples (C 1 , C 2 , C 3 , and C 4 ) from Anzali international wetland located southwest of the Caspian Sea was examined. Background concentration of each metal was calculated according to different statistical approaches. The results of multivariate statistical analysis showed that Fe and Mn might have significant role in the fate of Ni and Zn in sediment core samples. Different sediment quality indexes were utilized to assess metal pollution in sediment cores. Moreover, a new sediment quality index named aggregative toxicity index (ATI) based on sediment quality guidelines (SQGs) was developed to assess the degree of metal toxicity in an aggregative manner. The increasing pattern of metal pollution and their toxicity degree in upper layers of core samples indicated increasing effects of anthropogenic sources in the study area.

Internal structure of InP/ZnS nanocrystals unraveled by high-resolution soft X-ray photoelectron spectroscopy.

PubMed

Huang, Kai; Demadrille, Renaud; Silly, Mathieu G; Sirotti, Fausto; Reiss, Peter; Renault, Olivier

2010-08-24

High-energy resolution photoelectron spectroscopy (DeltaE < 200 meV) is used to investigate the internal structure of semiconductor quantum dots containing low Z-contrast elements. In InP/ZnS core/shell nanocrystals synthesized using a single-step procedure (core and shell precursors added at the same time), a homogeneously alloyed InPZnS core structure is evidenced by quantitative analysis of their In3d(5/2) spectra recorded at variable excitation energy. When using a two-step method (core InP nanocrystal synthesis followed by subsequent ZnS shell growth), XPS analysis reveals a graded core/shell interface. We demonstrate the existence of In-S and S(x)-In-P(1-x) bonding states in both types of InP/ZnS nanocrystals, which allows a refined view on the underlying reaction mechanisms.

[Analysis of core virion polypeptides from the pathogen causing chicken egg-drop syndrome].

PubMed

Iurov, G K; Dadykov, V A; Neugodova, G L; Naroditskiĭ, B S

1998-01-01

The cores of egg-drop syndrome virus (EDS-76) were isolated by the pyridine technique. EDS-76 proved to be much more resistant to pyridine disruption than other adenoviruses and treatment with 10% pyridine did not lead to complete dissociation of capsid and cores; only increase of pyridine concentration to 20% produced satisfactory results. At least three polypeptides (24, 10.5, and 6.5 kDa) were found in the core by SDS-PAGE, whereas the 40 kDa reacting with the core is most probably not a core component. Much more intensive reactions of the core with EDS-76 virion capsid suggest that its virion structure differs from that of other adenoviruses.

Core psychopathology in anorexia nervosa and bulimia nervosa: A network analysis.

PubMed

Forrest, Lauren N; Jones, Payton J; Ortiz, Shelby N; Smith, April R

2018-04-25

The cognitive-behavioral theory of eating disorders (EDs) proposes that shape and weight overvaluation are the core ED psychopathology. Core symptoms can be statistically identified using network analysis. Existing ED network studies support that shape and weight overvaluation are the core ED psychopathology, yet no studies have estimated AN core psychopathology and concerns exist about the replicability of network analysis findings. The current study estimated ED symptom networks among people with anorexia nervosa (AN) and bulimia nervosa (BN) and among a combined group of people with AN and BN. Participants were girls and women with AN (n = 604) and BN (n = 477) seeking residential ED treatment. ED symptoms were assessed with the Eating Disorder Examination-Questionnaire (EDE-Q); 27 of the EDE-Q items were included as nodes in symptom networks. Core symptoms were determined by expected influence and strength values. In all networks, desiring weight loss, restraint, shape and weight preoccupation, and shape overvaluation emerged as the most important symptoms. In addition, in the AN and combined networks, fearing weight gain emerged as an important symptom. In the BN network, weight overvaluation emerged as another important symptom. Findings support the cognitive-behavioral premise that shape and weight overvaluation are at the core of AN psychopathology. Our BN and combined network findings provide a high degree of replication of previous findings. Clinically, findings highlight the importance of considering shape and weight overvaluation as a severity specifier and primary treatment target for people with EDs. © 2018 Wiley Periodicals, Inc.

Drilling and Caching Architecture for the Mars2020 Mission

NASA Astrophysics Data System (ADS)

Zacny, K.

2013-12-01

We present a Sample Acquisition and Caching (SAC) architecture for the Mars2020 mission and detail how the architecture meets the sampling requirements described in the Mars2020 Science Definition Team (SDT) report. The architecture uses 'One Bit per Core' approach. Having dedicated bit for each rock core allows a reduction in the number of core transfer steps and actuators and this reduces overall mission risk. It also alleviates the bit life problem, eliminates cross contamination, and aids in hermetic sealing. An added advantage is faster drilling time, lower power, lower energy, and lower Weight on Bit (which reduces Arm preload requirements). To enable replacing of core samples, the drill bits are based on the BigTooth bit design. The BigTooth bit cuts a core diameter slightly smaller than the imaginary hole inscribed by the inner surfaces of the bits. Hence the rock core could be much easier ejected along the gravity vector. The architecture also has three additional types of bits that allow analysis of rocks. Rock Abrasion and Brushing Bit (RABBit) allows brushing and grinding of rocks in the same was as Rock Abrasion Tool does on MER. PreView bit allows viewing and analysis of rock core surfaces. Powder and Regolith Acquisition Bit (PRABit) captures regolith and rock powder either for in situ analysis or sample return. PRABit also allows sieving capabilities. The architecture can be viewed here: http://www.youtube.com/watch?v=_-hOO4-zDtE

Quantity of dates trumps quality of dates for dense Bayesian radiocarbon sediment chronologies - Gas ion source 14C dating instructed by simultaneous Bayesian accumulation rate modeling

NASA Astrophysics Data System (ADS)

Rosenheim, B. E.; Firesinger, D.; Roberts, M. L.; Burton, J. R.; Khan, N.; Moyer, R. P.

2016-12-01

Radiocarbon (14C) sediment core chronologies benefit from a high density of dates, even when precision of individual dates is sacrificed. This is demonstrated by a combined approach of rapid 14C analysis of CO2 gas generated from carbonates and organic material coupled with Bayesian statistical modeling. Analysis of 14C is facilitated by the gas ion source on the Continuous Flow Accelerator Mass Spectrometry (CFAMS) system at the Woods Hole Oceanographic Institution's National Ocean Sciences Accelerator Mass Spectrometry facility. This instrument is capable of producing a 14C determination of +/- 100 14C y precision every 4-5 minutes, with limited sample handling (dissolution of carbonates and/or combustion of organic carbon in evacuated containers). Rapid analysis allows over-preparation of samples to include replicates at each depth and/or comparison of different sample types at particular depths in a sediment or peat core. Analysis priority is given to depths that have the least chronologic precision as determined by Bayesian modeling of the chronology of calibrated ages. Use of such a statistical approach to determine the order in which samples are run ensures that the chronology constantly improves so long as material is available for the analysis of chronologic weak points. Ultimately, accuracy of the chronology is determined by the material that is actually being dated, and our combined approach allows testing of different constituents of the organic carbon pool and the carbonate minerals within a core. We will present preliminary results from a deep-sea sediment core abundant in deep-sea foraminifera as well as coastal wetland peat cores to demonstrate statistical improvements in sediment- and peat-core chronologies obtained by increasing the quantity and decreasing the quality of individual dates.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bi, G.; Liu, C.; Si, S.

This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis ofmore » reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)« less

2016 American College of Rheumatology/European League Against Rheumatism criteria for minimal, moderate, and major clinical response in adult dermatomyositis and polymyositis: An International Myositis Assessment and Clinical Studies Group/Paediatric Rheumatology International Trials Organisation Collaborative Initiative.

PubMed

Aggarwal, Rohit; Rider, Lisa G; Ruperto, Nicolino; Bayat, Nastaran; Erman, Brian; Feldman, Brian M; Oddis, Chester V; Amato, Anthony A; Chinoy, Hector; Cooper, Robert G; Dastmalchi, Maryam; Fiorentino, David; Isenberg, David; Katz, James D; Mammen, Andrew; de Visser, Marianne; Ytterberg, Steven R; Lundberg, Ingrid E; Chung, Lorinda; Danko, Katalin; García-De la Torre, Ignacio; Song, Yeong Wook; Villa, Luca; Rinaldi, Mariangela; Rockette, Howard; Lachenbruch, Peter A; Miller, Frederick W; Vencovsky, Jiri

2017-05-01

To develop response criteria for adult dermatomyositis (DM) and polymyositis (PM). Expert surveys, logistic regression, and conjoint analysis were used to develop 287 definitions using core set measures. Myositis experts rated greater improvement among multiple pairwise scenarios in conjoint analysis surveys, where different levels of improvement in 2 core set measures were presented. The PAPRIKA (Potentially All Pairwise Rankings of All Possible Alternatives) method determined the relative weights of core set measures and conjoint analysis definitions. The performance characteristics of the definitions were evaluated on patient profiles using expert consensus (gold standard) and were validated using data from a clinical trial. The nominal group technique was used to reach consensus. Consensus was reached for a conjoint analysis-based continuous model using absolute per cent change in core set measures (physician, patient, and extramuscular global activity, muscle strength, Health Assessment Questionnaire, and muscle enzyme levels). A total improvement score (range 0-100), determined by summing scores for each core set measure, was based on improvement in and relative weight of each core set measure. Thresholds for minimal, moderate, and major improvement were ≥20, ≥40, and ≥60 points in the total improvement score. The same criteria were chosen for juvenile DM, with different improvement thresholds. Sensitivity and specificity in DM/PM patient cohorts were 85% and 92%, 90% and 96%, and 92% and 98% for minimal, moderate, and major improvement, respectively. Definitions were validated in the clinical trial analysis for differentiating the physician rating of improvement (p<0.001). The response criteria for adult DM/PM consisted of the conjoint analysis model based on absolute per cent change in 6 core set measures, with thresholds for minimal, moderate, and major improvement. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/.

Post impact behavior of mobile reactor core containment systems

NASA Technical Reports Server (NTRS)

Puthoff, R. L.; Parker, W. G.; Vanbibber, L. E.

1972-01-01

The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.

The JCMT Gould Belt Survey: Dense Core Clusters in Orion B

NASA Astrophysics Data System (ADS)

Kirk, H.; Johnstone, D.; Di Francesco, J.; Lane, J.; Buckle, J.; Berry, D. S.; Broekhoven-Fiene, H.; Currie, M. J.; Fich, M.; Hatchell, J.; Jenness, T.; Mottram, J. C.; Nutter, D.; Pattle, K.; Pineda, J. E.; Quinn, C.; Salji, C.; Tisi, S.; Hogerheijde, M. R.; Ward-Thompson, D.; The JCMT Gould Belt Survey Team

2016-04-01

The James Clerk Maxwell Telescope Gould Belt Legacy Survey obtained SCUBA-2 observations of dense cores within three sub-regions of Orion B: LDN 1622, NGC 2023/2024, and NGC 2068/2071, all of which contain clusters of cores. We present an analysis of the clustering properties of these cores, including the two-point correlation function and Cartwright’s Q parameter. We identify individual clusters of dense cores across all three regions using a minimal spanning tree technique, and find that in each cluster, the most massive cores tend to be centrally located. We also apply the independent M-Σ technique and find a strong correlation between core mass and the local surface density of cores. These two lines of evidence jointly suggest that some amount of mass segregation in clusters has happened already at the dense core stage.

Two emissive-magnetic composite platforms for Hg(II) sensing and removal: The combination of magnetic core, silica molecular sieve and rhodamine chemosensors

NASA Astrophysics Data System (ADS)

Mao, Hanping; Liu, Zhongshou

2018-01-01

In this paper, a composite sensing platform for Hg(II) optical sensing and removal was designed and reported. A core-shell structure was adopted, using magnetic Fe3O4 nanoparticles as the core, silica molecular sieve MCM-41 as the shell, respectively. Two rhodamine derivatives were synthesized as chemosensor and covalently immobilized into MCM-41 tunnels. Corresponding composite samples were characterized with SEM/TEM images, XRD analysis, IR spectra, thermogravimetry and N2 adsorption/desorption analysis, which confirmed their core-shell structure. Their emission was increased by Hg(II), showing emission turn on effect. High selectivity, linear working curves and recyclability were obtained from these composite samples.

TRAC analyses for CCTF and SCTF tests and UPTF design/operation. [Cylindrical Core Test Facility; Slab Core Test Facility; Upper Plenum Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spore, J.W.; Cappiello, M.W.; Dotson, P.J.

The analytical support in 1985 for Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF), and Upper Plenum Test Facility (UPTF) tests involves the posttest analysis of 16 tests that have already been run in the CCTF and the SCTF and the pretest analysis of 3 tests to be performed in the UPTF. Posttest analysis is used to provide insight into the detailed thermal-hydraulic phenomena occurring during the refill and reflood tests performed in CCTF and SCTF. Pretest analysis is used to ensure that the test facility is operated in a manner consistent with the expected behavior of anmore » operating full-scale plant during an accident. To obtain expected behavior of a plant during an accident, two plant loss-of-coolant-accident (LOCA) calculations were performed: a 200% cold-leg-break LOCA calculation for a 2772 MW(t) Babcock and Wilcox plant and a 200% cold-leg-break LOCA calculation for a 3315 MW(t) Westinghouse plant. Detailed results are presented for several CCTF UPI tests and the Westinghouse plant analysis.« less

MEGA-CC: computing core of molecular evolutionary genetics analysis program for automated and iterative data analysis.

PubMed

Kumar, Sudhir; Stecher, Glen; Peterson, Daniel; Tamura, Koichiro

2012-10-15

There is a growing need in the research community to apply the molecular evolutionary genetics analysis (MEGA) software tool for batch processing a large number of datasets and to integrate it into analysis workflows. Therefore, we now make available the computing core of the MEGA software as a stand-alone executable (MEGA-CC), along with an analysis prototyper (MEGA-Proto). MEGA-CC provides users with access to all the computational analyses available through MEGA's graphical user interface version. This includes methods for multiple sequence alignment, substitution model selection, evolutionary distance estimation, phylogeny inference, substitution rate and pattern estimation, tests of natural selection and ancestral sequence inference. Additionally, we have upgraded the source code for phylogenetic analysis using the maximum likelihood methods for parallel execution on multiple processors and cores. Here, we describe MEGA-CC and outline the steps for using MEGA-CC in tandem with MEGA-Proto for iterative and automated data analysis. http://www.megasoftware.net/.

Computed Tomography Scanning and Geophysical Measurements of Core from the Coldstream 1MH Well

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crandall, Dustin M.; Brown, Sarah; Moore, Johnathan E.

The computed tomography (CT) facilities and the Multi-Sensor Core Logger (MSCL) at the National Energy Technology Laboratory (NETL) Morgantown, West Virginia site were used to characterize core of the Marcellus Shale from a vertical well, the Coldstream 1MH Well in Clearfield County, PA. The core is comprised primarily of the Marcellus Shale from a depth of 7,002 to 7,176 ft. The primary impetus of this work is a collaboration between West Virginia University (WVU) and NETL to characterize core from multiple wells to better understand the structure and variation of the Marcellus and Utica shale formations. As part of thismore » effort, bulk scans of core were obtained from the Coldstream 1MH well, provided by the Energy Corporation of America (now Greylock Energy). This report, and the associated scans, provide detailed datasets not typically available from unconventional shales for analysis. The resultant datasets are presented in this report, and can be accessed from NETL's Energy Data eXchange (EDX) online system using the following link: https://edx.netl.doe.gov/dataset/coldstream-1mh-well. All equipment and techniques used were non-destructive, enabling future examinations to be performed on these cores. None of the equipment used was suitable for direct visualization of the shale pore space, although fractures and discontinuities were detectable with the methods tested. Low resolution CT imagery with the NETL medical CT scanner was performed on the entire core. Qualitative analysis of the medical CT images, coupled with x-ray fluorescence (XRF), P-wave, and magnetic susceptibility measurements from the MSCL were useful in identifying zones of interest for more detailed analysis as well as fractured zones. En echelon fractures were observed at 7,100 ft and were CT scanned using NETL’s industrial CT scanner at higher resolution. The ability to quickly identify key areas for more detailed study with higher resolution will save time and resources in future studies. The combination of methods used provided a multi-scale analysis of this core and provides both a macro and micro description of the core that is relevant for many subsurface energy-related examinations that have traditionally been performed at NETL.« less

Extremely high energy hadron and gamma-ray families(3). Core structure of the halo of superfamily

NASA Technical Reports Server (NTRS)

Yamashita, S.; Ohsawa, A.; Chinellato, J. A.; Shibuya, E. H.

1985-01-01

The study of the core structure seen in the halo of Mini-Andromeda 3(M.A.3), which was observed in the Chacaltaya emulsion chamber, is presented. On the assumption that lateral distribution of darkness of the core is exponential type, i.e., D=D0exp(-R/r0), subtraction of D from halo darkness is performed until the cores are gone. The same quantity on cores obtained by this way are summarized. The analysis is preliminary and is going to be developed.

Who Is Opposed to Common Core and Why?

ERIC Educational Resources Information Center

Polikoff, Morgan S.; Hardaway, Tenice; Marsh, Julie A.; Plank, David N.

2016-01-01

Rising opposition to the Common Core Standards (CCS) has undermined implementation throughout the country. Yet there has been no scholarly analysis of the predictors of CCS opposition in the populace. This analysis uses poll data from a statewide poll of California voters to explore the demographic and policy predictors of CCS opposition. We find…

Energy efficient engine component development and integration program

NASA Technical Reports Server (NTRS)

1981-01-01

Accomplishments in the Energy Efficient Engine Component Development and Integration program during the period of April 1, 1981 through September 30, 1981 are discussed. The major topics considered are: (1) propulsion system analysis, design, and integration; (2) engine component analysis, design, and development; (3) core engine tests; and (4) integrated core/low spool testing.

Enhancements to the Image Analysis Tool for Core Punch Experiments and Simulations (vs. 2014)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hogden, John Edward; Unal, Cetin

A previous paper (Hogden & Unal, 2012, Image Analysis Tool for Core Punch Experiments and Simulations) described an image processing computer program developed at Los Alamos National Laboratory. This program has proven useful so developement has been continued. In this paper we describe enhacements to the program as of 2014.

Energy efficient engine. Volume 1: Component development and integration program

NASA Technical Reports Server (NTRS)

1981-01-01

Technology for achieving lower installed fuel consumption and lower operating costs in future commercial turbofan engines are developed, evaluated, and demonstrated. The four program objectives are: (1) propulsion system analysis; (2) component analysis, design, and development; (3) core design, fabrication, and test; and (4) integrated core/low spoon design, fabrication, and test.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

GVIPS Models and Software

NASA Technical Reports Server (NTRS)

Arnold, Steven M.; Gendy, Atef; Saleeb, Atef F.; Mark, John; Wilt, Thomas E.

2007-01-01

Two reports discuss, respectively, (1) the generalized viscoplasticity with potential structure (GVIPS) class of mathematical models and (2) the Constitutive Material Parameter Estimator (COMPARE) computer program. GVIPS models are constructed within a thermodynamics- and potential-based theoretical framework, wherein one uses internal state variables and derives constitutive equations for both the reversible (elastic) and the irreversible (viscoplastic) behaviors of materials. Because of the underlying potential structure, GVIPS models not only capture a variety of material behaviors but also are very computationally efficient. COMPARE comprises (1) an analysis core and (2) a C++-language subprogram that implements a Windows-based graphical user interface (GUI) for controlling the core. The GUI relieves the user of the sometimes tedious task of preparing data for the analysis core, freeing the user to concentrate on the task of fitting experimental data and ultimately obtaining a set of material parameters. The analysis core consists of three modules: one for GVIPS material models, an analysis module containing a specialized finite-element solution algorithm, and an optimization module. COMPARE solves the problem of finding GVIPS material parameters in the manner of a design-optimization problem in which the parameters are the design variables.

The influence of clarification and threats on life situation: patients’ experiences 1 year after TSCI

PubMed Central

Bjørnshave Noe, Bodil; Bjerrum, Merete; Angel, Sanne

2017-01-01

Introduction: The study was conducted at the Spinal Cord Injury Centre of Western Denmark (VCR). The aim of the study was to explore patients’ experiences following traumatic spinal cord injury and to identify characteristics of positive versus negative life situation 1 year post discharge from hospital rehabilitation. This was a qualitative study conducted using inductive content analysis. Case Presentation: In this qualitative study seven patients were interviewed one year after discharge from initial rehabilitation at the VCR. The interviews were analysed using inductive content analysis. Discussion: We found that two categories condensed the patients’ experiences of their life situation 1 year post discharge: ‘clarification in relation to overall life situation’ and ‘threat to core competences’. The transversal analysis across the derived categories identified different combinations of clarification and threats to core competences explaining the patients’ experiences: high degree of clarification combined with low degree of threat to core competences was indicative of positive life situation. Also, positive life situation was seen when a high degree of clarification compensated for high degree of threats on core competencies. In contrast, an overall stressful and negative life situation was influenced by poor clarification combined with a high degree of threat to core competences. However, when core competences can be transformed into new skills, threats were manageable. This study revealed that clarification related to overall life situation in combination with threat to core competences may explain traumatic spinal cord injury patients’ overall life situation 1 year post discharge. An appropriate balance characterises a positive life situation. There might be a need to pay attention to patients who are challenged by low degree of clarification and high degree of threats on core competencies 1 year post discharge as this may influence the life situation negatively. PMID:28435741

An advancement in removing extraneous color from wood for low-magnification reflected-light image analysis of conifer tree rings

Treesearch

Paul R. Sheppard; Alex Wiedenhoeft

2007-01-01

This paper describes the removal of extraneous color from increment cores of conifers prior to reflected-light image analysis of tree rings. Ponderosa pine in central New Mexico was chosen for study. Peroxide bleaching was used as a pretreatment to remove extraneous color and still yield usable wood for image analysis. The cores were bleached in 3% peroxide raised to...

Composite Sandwich Structures for Shock Mitigation and Energy Absorption

DTIC Science & Technology

2016-06-28

analysis of the blast performance of foam -core, composite sandwich panels was that on a per unit areal weight density basis, lighter and more crushable... foam cores offered greater blast resistance and energy absorption than the heavier and stronger foam cores. This was found to be the case even on an...absolute weight basis for cuNed sandwich panels and panels subjected to underwater blast. 15. SUBJECT TERMS composite; foam -core sandwich; blast

Reliability Analysis of Surface Mount Technology (SMT)

DTIC Science & Technology

1993-03-01

INSULATING PROTECTIVE COAT RESISTIVE FILM CCYLINDRICALCCERAMIC CORE 15057-3 Figure 1.3.3.1-2. Metal Electrode Face Bonding (MELF) Resistor 1.3.3.2...control. The dielectric materials are typically G-10 or polyimide. Exotic applications such as porcelain on a core have been found. The core material must...invar cores that provide both a heatsink conduction path and mechanical restraint that forces the composite P&IS CTE to more closely match the SMD CTE

Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ishii, Mamoru

The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results andmore » models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup power density on the flow instability. The experimental startup transient results showed the existence of three different flow instability mechanisms, i.e., flashing instability, condensation induced flow instability, and density wave oscillations. In addition, the void-reactivity feedback did not have significant effects on the flow instability during the startup transients for NMR-50. ii Several initial startup procedures with different power ramp rates were experimentally investigated to eliminate the flow instabilities observed from the startup transients. Particularly, the very slow startup transient and pressurized startup transient tests were performed and compared. It was found that the very slow startup transients by applying very small power density can eliminate the flashing oscillations in the single-phase natural circulation and stabilize the flow oscillations in the phase of net vapor generation. The initially pressurized startup procedure was tested to eliminate the flashing instability during the startup transients as well. The pressurized startup procedure included the initial pressurization, heat-up, and venting process. The startup transient tests showed that the pressurized startup procedure could eliminate the flow instability during the transition from single-phase flow to two-phase flow at low pressure conditions. The experimental results indicated that both startup procedures were applicable to the initial startup of NMR. However, the pressurized startup procedures might be preferred due to short operating hours required. In order to have a deeper understanding of natural circulation flow instability, the quasi-steady tests were performed using the test facility installed with preheater and subcooler. The effect of system pressure, core inlet subcooling, core power density, inlet flow resistance coefficient, and void reactivity feedback were investigated in the quasi-steady state tests. The experimental stability boundaries were determined between unstable and stable flow conditions in the dimensionless stability plane of inlet subcooling number and Zuber number. To predict the stability boundary theoretically, linear stability analysis in the frequency domain was performed at four sections of the natural circulation test loop. The flashing phenomena in the chimney section was considered as an axially uniform heat source. And the dimensionless characteristic equation of the pressure drop perturbation was obtained by considering the void fraction effect and outlet flow resistance in the core section. The theoretical flashing boundary showed some discrepancies with previous experimental data from the quasi-steady state tests. In the future, thermal non-equilibrium was recommended to improve the accuracy of flashing instability boundary. As another part of the funded research, flow instabilities of a PWR-type SMR under low pressure and low power conditions were investigated experimentally as well. The NuScale reactor design was selected as the prototype for the PWR-type SMR. In order to experimentally study the natural circulation behavior of NuScale iii reactor during accidental scenarios, detailed scaling analyses are necessary to ensure that the scaled phenomena could be obtained in a laboratory test facility. The three-level scaling method is used as well to obtain the scaling ratios derived from various non-dimensional numbers. The design of the ideally scaled facility (ISF) was initially accomplished based on these scaling ratios. Then the engineering scaled facility (ESF) was designed and constructed based on the ISF by considering engineering limitations including laboratory space, pipe size, and pipe connections etc. PWR-type SMR experiments were performed in this well-scaled test facility to investigate the potential thermal hydraulic flow instability during the blowdown events, which might occur during the loss of coolant accident (LOCA) and loss of heat sink accident (LOHS) of the prototype PWR-type SMR. Two kinds of experiments, normal blowdown event and cold blowdown event, were experimentally investigated and compared with code predictions. The normal blowdown event was experimentally simulated since an initial condition where the pressure was lower than the designed pressure of the experiment facility, while the code prediction of blowdown started from the normal operation condition. Important thermal hydraulic parameters including reactor pressure vessel (RPV) pressure, containment pressure, local void fraction and temperature, pressure drop and natural circulation flow rate were measured and analyzed during the blowdown event. The pressure and water level transients are similar to the experimental results published by NuScale [51], which proves the capability of current loop in simulating the thermal hydraulic transient of real PWR-type SMR. During the 20000s blowdown experiment, water level in the core was always above the active fuel assemble during the experiment and proved the safety of natural circulation cooling and water recycling design of PWR-type SMR. Besides, pressure, temperature, and water level transient can be accurately predicted by RELAP5 code. However, the oscillations of natural circulation flow rate, water level and pressure drops were observed during the blowdown transients. This kind of flow oscillations are related to the water level and the location upper plenum, which is a path for coolant flow from chimney to steam generator and down comer. In order to investigate the transients start from the opening of ADS valve in both experimental and numerical way, the cold blow-down experiment is conducted. For the cold blowdown event, different from setting both reactor iv pressure vessel (RPV) and containment at high temperature and pressure, only RPV was heated close to the highest designed pressure and then open the ADS valve, same process was predicted using RELAP5 code. By doing cold blowdown experiment, the entire transients from the opening of ADS can be investigated by code and benchmarked with experimental data. Similar flow instability observed in the cold blowdown experiment. The comparison between code prediction and experiment data showed that the RELAP5 code can successfully predict the pressure void fraction and temperature transient during the cold blowdown event with limited error, but numerical instability exists in predicting natural circulation flow rate. Besides, the code is lack of capability in predicting the water level related flow instability observed in experiments.« less

Simulation of cracking cores when molding piston components

NASA Astrophysics Data System (ADS)

Petrenko, Alena; Soukup, Josef

2014-08-01

The article deals with pistons casting made from aluminum alloy. Pistons are casting at steel mold with steel core. The casting is provided by gravity casting machine. The each machine is equipped by two metal molds, which are preheated above temperature 160 °C before use. The steel core is also preheated by flame. The metal molds and cores are heated up within the casting process. The temperature of the metal mold raise up to 200 °C and temperature of core is higher. The surface of the core is treated by nitration. The mold and core are cooled down by water during casting process. The core is overheated and its top part is finally cracked despite its intensive water-cooling. The life time cycle of the core is decreased to approximately 5 to 15 thousands casting, which is only 15 % of life time cycle of core for production of other pistons. The article presents the temperature analysis of the core.

TRAC-PD2 posttest analysis of the CCTF Evaluation-Model Test C1-19 (Run 38). [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Motley, F.

The results of a Transient Reactor Analysis Code posttest analysis of the Cylindral Core Test Facility Evaluation-Model Test agree very well with the results of the experiment. The good agreement obtained verifies the multidimensional analysis capability of the TRAC code. Because of the steep radial power profile, the importance of using fine noding in the core region was demonstrated (as compared with poorer results obtained from an earlier pretest prediction that used a coarsely noded model).

MetaCoMET: a web platform for discovery and visualization of the core microbiome

USDA-ARS?s Scientific Manuscript database

A key component of the analysis of microbiome datasets is the identification of OTUs shared between multiple experimental conditions, commonly referred to as the core microbiome. Results: We present a web platform named MetaCoMET that enables the discovery and visualization of the core microbiome an...

Explicit Instruction in Core Reading Programs

ERIC Educational Resources Information Center

Reutzel, D. Ray; Child, Angela; Jones, Cindy D.; Clark, Sarah K.

2014-01-01

The purpose of this study was to conduct a content analysis of the types and occurrences of explicit instructional moves recommended for teaching five essentials of effective reading instruction in grades 1, 3, and 5 core reading program teachers' editions in five widely marketed core reading programs. Guided practice was the most frequently…

Cone Calorimeter Analysis of FRT Intumescent and Untreated Foam Core Particleboards

Treesearch

Mark A. Dietenberger; Ali Shalbafan; Johannes Welling; Charles Boardman

2012-01-01

The effectiveness of treatments of the surface layer of novel foam core particleboards were evaluated by means of Cone calorimeter tests. Foam core particleboards with variations of surface layer treatment, adhesives and surface layer thicknesses under similar processing conditions were used to produce the test specimen for the Cone calorimeter tests. Ignitability,...

GENIE: a software package for gene-gene interaction analysis in genetic association studies using multiple GPU or CPU cores.

PubMed

Chikkagoudar, Satish; Wang, Kai; Li, Mingyao

2011-05-26

Gene-gene interaction in genetic association studies is computationally intensive when a large number of SNPs are involved. Most of the latest Central Processing Units (CPUs) have multiple cores, whereas Graphics Processing Units (GPUs) also have hundreds of cores and have been recently used to implement faster scientific software. However, currently there are no genetic analysis software packages that allow users to fully utilize the computing power of these multi-core devices for genetic interaction analysis for binary traits. Here we present a novel software package GENIE, which utilizes the power of multiple GPU or CPU processor cores to parallelize the interaction analysis. GENIE reads an entire genetic association study dataset into memory and partitions the dataset into fragments with non-overlapping sets of SNPs. For each fragment, GENIE analyzes: 1) the interaction of SNPs within it in parallel, and 2) the interaction between the SNPs of the current fragment and other fragments in parallel. We tested GENIE on a large-scale candidate gene study on high-density lipoprotein cholesterol. Using an NVIDIA Tesla C1060 graphics card, the GPU mode of GENIE achieves a speedup of 27 times over its single-core CPU mode run. GENIE is open-source, economical, user-friendly, and scalable. Since the computing power and memory capacity of graphics cards are increasing rapidly while their cost is going down, we anticipate that GENIE will achieve greater speedups with faster GPU cards. Documentation, source code, and precompiled binaries can be downloaded from http://www.cceb.upenn.edu/~mli/software/GENIE/.

Coherent network analysis of gravitational waves from three-dimensional core-collapse supernova models

NASA Astrophysics Data System (ADS)

Hayama, Kazuhiro; Kuroda, Takami; Kotake, Kei; Takiwaki, Tomoya

2015-12-01

Using predictions from three-dimensional (3D) hydrodynamics simulations of core-collapse supernovae (CCSNe), we present a coherent network analysis for the detection, reconstruction, and source localization of the gravitational-wave (GW) signals. We use the RIDGE pipeline for the analysis, in which the network of LIGO Hanford, LIGO Livingston, VIRGO, and KAGRA is considered. By combining with a GW spectrogram analysis, we show that several important hydrodynamics features in the original waveforms persist in the waveforms of the reconstructed signals. The characteristic excess in the spectrograms originates not only from the rotating core collapse, bounce, and subsequent ringdown of the proto-neutron star (PNS) as previously identified, but also from the formation of magnetohydrodynamics jets and nonaxisymmetric instabilities in the vicinity of the PNS. Regarding the GW signals emitted near the rotating core bounce, the horizon distance extends up to ˜18 kpc for the most rapidly rotating 3D model in this work. Following the rotating core bounce, the dominant source of the GW emission shifts to the nonaxisymmetric instabilities. The horizon distances extend maximally up to ˜40 kpc seen from the spin axis. With an increasing number of 3D models trending towards explosion recently, our results suggest that in addition to the best-studied GW signals due to rotating core collapse and bounce, the time is ripe to consider how we can do science from GWs of CCSNe much more seriously than before. In particular, the quasiperiodic signals due to the nonaxisymmetric instabilities and the detectability deserves further investigation to elucidate the inner workings of the rapidly rotating CCSNe.

GENIE: a software package for gene-gene interaction analysis in genetic association studies using multiple GPU or CPU cores

PubMed Central

2011-01-01

Background Gene-gene interaction in genetic association studies is computationally intensive when a large number of SNPs are involved. Most of the latest Central Processing Units (CPUs) have multiple cores, whereas Graphics Processing Units (GPUs) also have hundreds of cores and have been recently used to implement faster scientific software. However, currently there are no genetic analysis software packages that allow users to fully utilize the computing power of these multi-core devices for genetic interaction analysis for binary traits. Findings Here we present a novel software package GENIE, which utilizes the power of multiple GPU or CPU processor cores to parallelize the interaction analysis. GENIE reads an entire genetic association study dataset into memory and partitions the dataset into fragments with non-overlapping sets of SNPs. For each fragment, GENIE analyzes: 1) the interaction of SNPs within it in parallel, and 2) the interaction between the SNPs of the current fragment and other fragments in parallel. We tested GENIE on a large-scale candidate gene study on high-density lipoprotein cholesterol. Using an NVIDIA Tesla C1060 graphics card, the GPU mode of GENIE achieves a speedup of 27 times over its single-core CPU mode run. Conclusions GENIE is open-source, economical, user-friendly, and scalable. Since the computing power and memory capacity of graphics cards are increasing rapidly while their cost is going down, we anticipate that GENIE will achieve greater speedups with faster GPU cards. Documentation, source code, and precompiled binaries can be downloaded from http://www.cceb.upenn.edu/~mli/software/GENIE/. PMID:21615923

Dating a tropical ice core by time-frequency analysis of ion concentration depth profiles

NASA Astrophysics Data System (ADS)

Gay, M.; De Angelis, M.; Lacoume, J.-L.

2014-09-01

Ice core dating is a key parameter for the interpretation of the ice archives. However, the relationship between ice depth and ice age generally cannot be easily established and requires the combination of numerous investigations and/or modelling efforts. This paper presents a new approach to ice core dating based on time-frequency analysis of chemical profiles at a site where seasonal patterns may be significantly distorted by sporadic events of regional importance, specifically at the summit area of Nevado Illimani (6350 m a.s.l.), located in the eastern Bolivian Andes (16°37' S, 67°46' W). We used ion concentration depth profiles collected along a 100 m deep ice core. The results of Fourier time-frequency and wavelet transforms were first compared. Both methods were applied to a nitrate concentration depth profile. The resulting chronologies were checked by comparison with the multi-proxy year-by-year dating published by de Angelis et al. (2003) and with volcanic tie points. With this first experiment, we demonstrated the efficiency of Fourier time-frequency analysis when tracking the nitrate natural variability. In addition, we were able to show spectrum aliasing due to under-sampling below 70 m. In this article, we propose a method of de-aliasing which significantly improves the core dating in comparison with annual layer manual counting. Fourier time-frequency analysis was applied to concentration depth profiles of seven other ions, providing information on the suitability of each of them for the dating of tropical Andean ice cores.

Design of a Modular E-Core Flux Concentrating Axial Flux Machine: Preprint

DOE Office of Scientific and Technical Information (OSTI.GOV)

Husain, Tausif; Sozer, Yilmaz; Husain, Iqbal

2015-08-24

In this paper a novel E-Core axial flux machine is proposed. The machine has a double-stator, single-rotor configuration with flux-concentrating ferrite magnets and pole windings across each leg of an E-Core stator. E-Core stators with the proposed flux-concentrating rotor arrangement result in better magnet utilization and higher torque density. The machine also has a modular structure facilitating simpler construction. This paper presents a single-phase and a three-phase version of the E-Core machine. Case studies for a 1.1-kW, 400-rpm machine for both the single-phase and three-phase axial flux machines are presented. The results are verified through 3D finite element analysis. facilitatingmore » simpler construction. This paper presents a single-phase and a three-phase version of the E-Core machine. Case studies for a 1.1-kW, 400-rpm machine for both the single-phase and three-phase axial flux machines are presented. The results are verified through 3D finite element analysis.« less

Influence of peak oral temperatures on veneer–core interface stress state

PubMed Central

Marrelli, Massimo; Pujia, Antonella; Apicella, Davide; Sansalone, Salvatore; Tatullo, Marco

2015-01-01

Abstract Objective: There is a growing interest for the use of Y-TZP zirconia as core material in veneered all-ceramic prostheses. The objective of this study was to evaluate the influence of CET on the stress distribution of a porcelain layered to zirconia core single crowns by finite elements analysis. Material and methods: CET of eight different porcelains was considered during the analysis. Results: Results of this study indicated that the mismatch in CET between the veneering porcelain and the Y-TZP zirconia core has to be minimum (0.5–1 μm/mK) so as to decrease the growing of residual stresses which could bring chipping. Conclusions: The stress state due to temperature variation should be carefully taken into consideration while studying the effect of mechanical load on zirconia core crown by FEA. The interfacial stress state can be increased by temperature variation up to 20% with respect to the relative failure parameter (interface strength in this case). This means that stress due to mechanical load combined to temperature variation-induced stress can lead porcelain veneer–zirconia core interfaces to failure. PMID:28642897

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki; Miura, Ryosuke

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design.more » The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.« less

Core-valence stockholder AIM analysis and its connection to nonadiabatic effects in small molecules.

PubMed

Amaral, Paulo H R; Mohallem, José R

2017-05-21

A previous theory of separation of motions of core and valence fractions of electrons in a molecule [J. R. Mohallem et al., Chem. Phys. Lett. 501, 575 (2011)] is invoked as basis for the useful concept of Atoms-in-Molecules (AIM) in the stockholder scheme. The output is a new tool for the analysis of the chemical bond that identifies core and valence electron density fractions (core-valence stockholder AIM (CVSAIM)). One-electron effective potentials for each atom are developed, which allow the identification of the parts of the AIM which move along with the nuclei (cores). This procedure results in a general method for obtaining effective masses that yields accurate non-adiabatic corrections to vibrational energies, necessary to attain cm -1 accuracy in molecular spectroscopy. The clear-cut determination of the core masses is exemplified for either homonuclear (H 2 + , H 2 ) or heteronuclear (HeH + , LiH) molecules. The connection of CVSAIM with independent physically meaningful quantities can resume the question of whether they are observable or not.

Core-valence stockholder AIM analysis and its connection to nonadiabatic effects in small molecules

PubMed Central

Amaral, Paulo H. R.; Mohallem, José R.

2017-01-01

A previous theory of separation of motions of core and valence fractions of electrons in a molecule [J. R. Mohallem et al., Chem. Phys. Lett. 501, 575 (2011)] is invoked as basis for the useful concept of Atoms-in-Molecules (AIM) in the stockholder scheme. The output is a new tool for the analysis of the chemical bond that identifies core and valence electron density fractions (core-valence stockholder AIM (CVSAIM)). One-electron effective potentials for each atom are developed, which allow the identification of the parts of the AIM which move along with the nuclei (cores). This procedure results in a general method for obtaining effective masses that yields accurate non-adiabatic corrections to vibrational energies, necessary to attain cm−1 accuracy in molecular spectroscopy. The clear-cut determination of the core masses is exemplified for either homonuclear (H2+, H2) or heteronuclear (HeH+, LiH) molecules. The connection of CVSAIM with independent physically meaningful quantities can resume the question of whether they are observable or not. PMID:28527456

Microbial Analysis of Australian Dry Lake Cores; Analogs For Biogeochemical Processes

NASA Astrophysics Data System (ADS)

Nguyen, A. V.; Baldridge, A. M.; Thomson, B. J.

2014-12-01

Lake Gilmore in Western Australia is an acidic ephemeral lake that is analogous to Martian geochemical processes represented by interbedded phyllosilicates and sulfates. These areas demonstrate remnants of a global-scale change on Mars during the late Noachian era from a neutral to alkaline pH to relatively lower pH in the Hesperian era that continues to persist today. The geochemistry of these areas could possibly be caused by small-scale changes such as microbial metabolism. Two approaches were used to determine the presence of microbes in the Australian dry lake cores: DNA analysis and lipid analysis. Detecting DNA or lipids in the cores will provide evidence of living or deceased organisms since they provide distinct markers for life. Basic DNA analysis consists of extraction, amplification through PCR, plasmid cloning, and DNA sequencing. Once the sequence of unknown DNA is known, an online program, BLAST, will be used to identify the microbes for further analysis. The lipid analysis approach consists of phospholipid fatty acid analysis that is done by Microbial ID, which will provide direct identification any microbes from the presence of lipids. Identified microbes are then compared to mineralogy results from the x-ray diffraction of the core samples to determine if the types of metabolic reactions are consistent with the variation in composition in these analog deposits. If so, it provides intriguing implications for the presence of life in similar Martian deposits.

Fractographic study of the behavior of different ceramic veneers on full coverage crowns in relation to supporting core materials

PubMed Central

Agustín-Panadero, Rubén; Román-Rodriguez, Juan L.; Solá-Ruíz, María F.; Granell-Ruíz, María; Fons-Font, Antonio

2013-01-01

Objectives: To observe porcelain veneer behavior of zirconia and metal-ceramic full coverage crowns when subjected to compression testing, comparing zirconia cores to metal cores. Study Design: The porcelain fracture surfaces of 120 full coverage crowns (60 with a metal core and 60 with a zirconia core) subjected to static load (compression) testing were analyzed. Image analysis was performed using macroscopic processing with 8x and 12x enlargement. Five samples from each group were prepared and underwent scanning electron microscope (SEM) analysis in order to make a fractographic study of fracture propagation in the contact area and composition analysis in the most significant areas of the specimen. Results: Statistically significant differences in fracture type (cohesive or adhesive) were found between the metal-ceramic and zirconia groups: the incidence of adhesive fracture was seen to be greater in metal-ceramic groups (92%) and cohesive fracture was more frequent in zirconium oxide groups (72%). The fracture propagation pattern was on the periphery of the contact area in the full coverage crown restorations selected for fractographic study. Conclusions: The greater frequency of cohesive fracture in restorations with zirconia cores indicates that their behavior is inadequate compared to metal-ceramic restorations and that further research is needed to improve their clinical performance. Key words:Zirconia, zirconium oxide, fractography, composition, porcelain veneers, fracture, cohesive, adhesive. PMID:24455092

Cadmium-free aqueous synthesis of ZnSe and ZnSe@ZnS core-shell quantum dots and their differential bioanalyte sensing potential

NASA Astrophysics Data System (ADS)

Mir, Irshad Ahmad; Rawat, Kamla; Bohidar, H. B.

2016-10-01

Herein we report a facile and cadmium-free approach to prepare water-soluble fluorescent ZnSe@ZnS core-shell quantum dots (QDs), using thioglycolic acid (TGA) ligand as a stabilizer and thiourea as a sulfur source. The optical properties and morphology of the obtained core-shell QDs were characterized by UV-vis and fluorescence spectroscopy, transmission electron microscopy (TEM), energy-dispersive x-ray analysis (EDX), x-ray diffraction (XRD), electrophoresis and dynamic light scattering (DLS) techniques. TEM analysis, and electrophoresis data showed that ZnSe core had an average size of 3.60 ± 0.12 nm and zeta potential of -38 mV; and for ZnSe@ZnS QDs, the mean size was 4.80 ± 0.20 nm and zeta potential was -45 mV. Compared to the core ZnSe QDs, the quantum yield of these core-shell structures was higher (13% versus 32%). These were interacted with five common bioanalytes such as, ascorbic acid, citric acid, oxalic acid, glucose and cholesterol which revealed fluorescence quenching due to concentration dependent binding of analytes to the core only, and core-shell QDs. The binding pattern followed the sequence: cholesterol < glucose < ascorbic acid < oxalic acid < citric acid for ZnSe, and cholesterol < glucose < oxalic acid < ascorbic acid < citric acid for core-shell QDs. Thus, enhanced binding was noticed for the analyte citric acid which may facilitate development of a fluorescence-based sensor based on the ZnSe core-only quantum dot platform. Further, the hydrophilic core-shell structure may find use in cell imaging applications.

The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

NASA Astrophysics Data System (ADS)

Susilo, J.; Suparlina, L.; Deswandri; Sunaryo, G. R.

2018-02-01

The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

Intratumoral heterogeneity as a source of discordance in breast cancer biomarker classification.

PubMed

Allott, Emma H; Geradts, Joseph; Sun, Xuezheng; Cohen, Stephanie M; Zirpoli, Gary R; Khoury, Thaer; Bshara, Wiam; Chen, Mengjie; Sherman, Mark E; Palmer, Julie R; Ambrosone, Christine B; Olshan, Andrew F; Troester, Melissa A

2016-06-28

Spatial heterogeneity in biomarker expression may impact breast cancer classification. The aims of this study were to estimate the frequency of spatial heterogeneity in biomarker expression within tumors, to identify technical and biological factors contributing to spatial heterogeneity, and to examine the impact of discordant biomarker status within tumors on clinical record agreement. Tissue microarrays (TMAs) were constructed using two to four cores (1.0 mm) for each of 1085 invasive breast cancers from the Carolina Breast Cancer Study, which is part of the AMBER Consortium. Immunohistochemical staining for estrogen receptor (ER), progesterone receptor (PR), and human epidermal growth factor receptor 2 (HER2) was quantified using automated digital imaging analysis. The biomarker status for each core and for each case was assigned using clinical thresholds. Cases with core-to-core biomarker discordance were manually reviewed to distinguish intratumoral biomarker heterogeneity from misclassification of biomarker status by the automated algorithm. The impact of core-to-core biomarker discordance on case-level agreement between TMAs and the clinical record was evaluated. On the basis of automated analysis, discordant biomarker status between TMA cores occurred in 9 %, 16 %, and 18 % of cases for ER, PR, and HER2, respectively. Misclassification of benign epithelium and/or ductal carcinoma in situ as invasive carcinoma by the automated algorithm was implicated in discordance among cores. However, manual review of discordant cases confirmed spatial heterogeneity as a source of discordant biomarker status between cores in 2 %, 7 %, and 8 % of cases for ER, PR, and HER2, respectively. Overall, agreement between TMA and clinical record was high for ER (94 %), PR (89 %), and HER2 (88 %), but it was reduced in cases with core-to-core discordance (agreement 70 % for ER, 61 % for PR, and 57 % for HER2). Intratumoral biomarker heterogeneity may impact breast cancer classification accuracy, with implications for clinical management. Both manually confirmed biomarker heterogeneity and misclassification of biomarker status by automated image analysis contribute to discordant biomarker status between TMA cores. Given that manually confirmed heterogeneity is uncommon (<10 % of cases), large studies are needed to study the impact of heterogeneous biomarker expression on breast cancer classification and outcomes.

Posttest analysis of the FFTF inherent safety tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Padilla, A. Jr.; Claybrook, S.W.

Inherent safety tests were performed during 1986 in the 400-MW (thermal) Fast Flux Test Facility (FFTF) reactor to demonstrate the effectiveness of an inherent shutdown device called the gas expansion module (GEM). The GEM device provided a strong negative reactivity feedback during loss-of-flow conditions by increasing the neutron leakage as a result of an expanding gas bubble. The best-estimate pretest calculations for these tests were performed using the IANUS plant analysis code (Westinghouse Electric Corporation proprietary code) and the MELT/SIEX3 core analysis code. These two codes were also used to perform the required operational safety analyses for the FFTF reactormore » and plant. Although it was intended to also use the SASSYS systems (core and plant) analysis code, the calibration of the SASSYS code for FFTF core and plant analysis was not completed in time to perform pretest analyses. The purpose of this paper is to present the results of the posttest analysis of the 1986 FFTF inherent safety tests using the SASSYS code.« less

Analysis of Stainless Steel Sandwich Panels with a Metal Foam Core for Lightweight Fan Blade Design

NASA Technical Reports Server (NTRS)

Min, James B.; Ghosn, Louis J.; Lerch, Bradley A.; Raj, Sai V.; Holland, Frederic A., Jr.; Hebsur, Mohan G.

2004-01-01

The quest for cheap, low density and high performance materials in the design of aircraft and rotorcraft engine fan and propeller blades poses immense challenges to the materials and structural design engineers. The present study investigates the use of a sandwich foam fan blade mae up of solid face sheets and a metal foam core. The face sheets and the metal foam core material were an aerospace grade precipitation hardened 17-4 PH stainless steel with high strength and high toughness. The resulting structures possesses a high stiffness while being lighter than a similar solid construction. The material properties of 17-4 PH metal foam are reviewed briefly to describe the characteristics of sandwich structure for a fan blade application. A vibration analysis for natural frequencies and a detailed stress analysis on the 17-4 PH sandwich foam blade design for different combinations of kin thickness and core volume are presented with a comparison to a solid titanium blade.

Application Performance Analysis and Efficient Execution on Systems with multi-core CPUs, GPUs and MICs: A Case Study with Microscopy Image Analysis

PubMed Central

Teodoro, George; Kurc, Tahsin; Andrade, Guilherme; Kong, Jun; Ferreira, Renato; Saltz, Joel

2015-01-01

We carry out a comparative performance study of multi-core CPUs, GPUs and Intel Xeon Phi (Many Integrated Core-MIC) with a microscopy image analysis application. We experimentally evaluate the performance of computing devices on core operations of the application. We correlate the observed performance with the characteristics of computing devices and data access patterns, computation complexities, and parallelization forms of the operations. The results show a significant variability in the performance of operations with respect to the device used. The performances of operations with regular data access are comparable or sometimes better on a MIC than that on a GPU. GPUs are more efficient than MICs for operations that access data irregularly, because of the lower bandwidth of the MIC for random data accesses. We propose new performance-aware scheduling strategies that consider variabilities in operation speedups. Our scheduling strategies significantly improve application performance compared to classic strategies in hybrid configurations. PMID:28239253

Synthesis, characterization and nitrite ion sensing performance of reclaimable composite samples through a core-shell structure

NASA Astrophysics Data System (ADS)

Cui, Xiao; Yuqing, Zhao; Cui, Jiantao; Zheng, Qian; Bo, Wang

2018-02-01

The following paper reported and discussed a nitrite ion optical sensing platform based on a core-shell structure, using superamagnetic nanoparticles as the core, a silica molecular sieve MCM-41 as the shell and two rhodamine derivatives as probe, respectively. This superamagnetic core made this sensing platform reclaimable after finishing nitrite ion sensing procedure. This sensing platform was carefully characterized by means of electron microscopy images, porous structure analysis, magnetic response, IR spectra and thermal stability analysis. Detailed analysis suggested that the emission of these composite samples was quenchable by nitrite ion, showing emission turn off effect. A static sensing mechanism based on an additive reaction between chemosensors and nitrite ion was proposed. These composite samples followed Demas quenching equation against different nitrite ion concentrations. Limit of detection value was obtained as low as 0.4 μM. It was found that, after being quenched by nitrite ion, these composite samples could be reclaimed and recovered by sulphamic acid, confirming their recyclability.

Field analysis of the Cerenkov doubling of infrared coherent radiation utilizing an organic crystal core bounded by a glass capillary

NASA Astrophysics Data System (ADS)

Hayata, K.; Yanagawa, K.; Koshiba, M.

1990-12-01

A mode field analysis is presented of the second-harmonic electromagnetic wave that radiates from a nonlinear core bounded by a dielectric cladding. With this analysis the ultimate performance of the organic crystal-cored single-mode optical fiber waveguide as a guided-wave frequency doubler is evaluated through the solution of nonlinear parametric equations derived from Maxwell's equations under some assumptions. As a phase-matching scheme, a Cerenkov approach is considered because of advantages in actual device applications, in which the phase matching is achievable between the fundamental guided LP01 mode and the second-harmonic radiation (leaky) mode. Calculated results for organic cores made of benzil, 4-(N,N-dimethyl-amino)-3-acetamidonitrobenzen, 2-methyl-4-nitroaniline, and 4'-nitrobenzilidene-3-acetoamino-4-metxianiline provide useful data for designing an efficient fiber-optic wavelength converter utilizing nonlinear parametric processes. A detailed comparison is made between results for infinite and finite cladding thicknesses.

Exploring LIS Students' Beliefs in Importance and Self-Efficacy of Core Information Literacy Competencies

ERIC Educational Resources Information Center

Pinto, Maria; Pascual, Rosaura Fernandez

2016-01-01

Understanding perceptions of Library and Information Science (LIS) students on two dimensions--belief in the importance (BIM) of a set of core information competencies, and Self-Efficacy (SE)--is pursued. Factor analysis implementation raises a clear distinction between BIM and SE results. This analysis points to two sets of competencies:…

Publications - GMC 232 | Alaska Division of Geological & Geophysical

Science.gov Websites

Surveys Skip to content State of Alaska myAlaska My Government Resident Business in Alaska DGGS GMC 232 Publication Details Title: Petrographic analysis and formation damage potential of core Reference Hickey, J.J., 1994, Petrographic analysis and formation damage potential of core plugs (12,017

The Relationship of Core Self-Evaluations and Life Satisfaction in College Students with Disabilities: Evaluation of a Mediator Model

ERIC Educational Resources Information Center

Smedema, Susan Miller; Chan, Fong; Yaghmaian, Rana A.; Cardoso, Elizabeth DaSilva; Muller, Veronica; Keegan, John; Dutta, Alo; Ebener, Deborah J.

2015-01-01

This study examined the factorial structure of the construct core self-evaluations (CSE) and tested a mediational model of the relationship between CSE and life satisfaction in college students with disabilities. We conducted a quantitative descriptive design using exploratory and confirmatory factor analysis and multiple regression analysis.…

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.

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

Two emissive-magnetic composite platforms for Hg(II) sensing and removal: The combination of magnetic core, silica molecular sieve and rhodamine chemosensors.

PubMed

Mao, Hanping; Liu, Zhongshou

2018-01-15

In this paper, a composite sensing platform for Hg(II) optical sensing and removal was designed and reported. A core-shell structure was adopted, using magnetic Fe 3 O 4 nanoparticles as the core, silica molecular sieve MCM-41 as the shell, respectively. Two rhodamine derivatives were synthesized as chemosensor and covalently immobilized into MCM-41 tunnels. Corresponding composite samples were characterized with SEM/TEM images, XRD analysis, IR spectra, thermogravimetry and N 2 adsorption/desorption analysis, which confirmed their core-shell structure. Their emission was increased by Hg(II), showing emission turn on effect. High selectivity, linear working curves and recyclability were obtained from these composite samples. Copyright © 2017 Elsevier B.V. All rights reserved.

Fabrication of Ni@Ti core-shell nanoparticles by modified gas aggregation source

NASA Astrophysics Data System (ADS)

Hanuš, J.; Vaidulych, M.; Kylián, O.; Choukourov, A.; Kousal, J.; Khalakhan, I.; Cieslar, M.; Solař, P.; Biederman, H.

2017-11-01

Ni@Ti core-shell nanoparticles were prepared by a vacuum based method using the gas aggregation source (GAS) of nanoparticles. Ni nanoparticles fabricated in the GAS were afterwards coated by a Ti shell. The Ti shell was deposited by means of magnetron sputtering. The Ni nanoparticles were decelerated in the vicinity of the magnetron to the Ar drift velocity in the second deposition chamber. X-ray photoelectron spectroscopy and energy dispersive x-ray spectroscopy analysis of the nanoparticles showed the core-shell structure. It was shown that the thickness of the shell can be easily tuned by the process parameters with a maximum achieved thickness of the Ti shell ~2.5 nm. The core-shell structure was confirmed by the STEM analysis of the particles.

A core handling device for the Mars Sample Return Mission

NASA Technical Reports Server (NTRS)

Gwynne, Owen

1989-01-01

A core handling device for use on Mars is being designed. To provide a context for the design study, it was assumed that a Mars Rover/Sample Return (MRSR) Mission would have the following characteristics: a year or more in length; visits by the rover to 50 or more sites; 100 or more meter-long cores being drilled by the rover; and the capability of returning about 5 kg of Mars regolith to Earth. These characteristics lead to the belief that in order to bring back a variegated set of samples that can address the range of scientific objetives for a MRSR mission to Mars there needs to be considerable analysis done on board the rover. Furthermore, the discrepancy between the amount of sample gathered and the amount to be returned suggests that there needs to be some method of choosing the optimal set of samples. This type of analysis will require pristine material-unaltered by the drilling process. Since the core drill thermally and mechanically alters the outer diameter (about 10 pct) of the core sample, this outer area cannot be used. The primary function of the core handling device is to extract subsamples from the core and to position these subsamples, and the core itself if needed, with respect to the various analytical instruments that can be used to perform these analyses.

Preliminary engineering design of sodium-cooled CANDLE core

NASA Astrophysics Data System (ADS)

Takaki, Naoyuki; Namekawa, Azuma; Yoda, Tomoyuki; Mizutani, Akihiko; Sekimoto, Hiroshi

2012-06-01

The CANDLE burning process is characterized by the autonomous shifting of burning region with constant reactivity and constant spacial power distribution. Evaluations of such critical burning process by using widely used neutron diffusion and burning codes under some realistic engineering constraints are valuable to confirm the technical feasibility of the CANDLE concept and to put the idea into concrete core design. In the first part of this paper, it is discussed that whether the sustainable and stable CANDLE burning process can be reproduced even by using conventional core analysis tools such as SLAROM and CITATION-FBR. As a result, it is certainly possible to demonstrate it if the proper core configuration and initial fuel composition required as CANDLE core are applied to the analysis. In the latter part, an example of a concrete image of sodium cooled, metal fuel, 2000MWt rating CANDLE core has been presented by assuming an emerging inevitable technology of recladding. The core satisfies engineering design criteria including cladding temperature, pressure drop, linear heat rate, and cumulative damage fraction (CDF) of cladding, fast neutron fluence and sodium void reactivity which are defined in the Japanese FBR design project. It can be concluded that it is feasible to design CADLE core by using conventional codes while satisfying some realistic engineering design constraints assuming that recladding at certain time interval is technically feasible.

An overview of zinc addition for BWR dose rate control

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marble, W.J.

1995-03-01

This paper presents an overview of the BWRs employing feedwater zinc addition to reduce primary system dose rates. It identifies which BWRs are using zinc addition and reviews the mechanical injection and passive addition hardware currently being employed. The impact that zinc has on plant chemistry, including the factor of two to four reduction in reactor water Co-60 concentrations, is discussed. Dose rate results, showing the benefits of implementing zinc on either fresh piping surfaces or on pipes with existing films are reviewed. The advantages of using zinc that is isotopically enhanced by the depletion of the Zn-64 precursor tomore » Zn-65 are identified.« less

BWRVIP-140NP: BWR Vessel and Internals Project Fracture Toughness and Crack Growth Program on Irradiated Austenitic Stainless Steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilman, J

2005-03-15

To prepare for this project, EPRI and BWRVIP conducted a workshop at Ponte Vedra Beach, Florida during February 19-21, 2003 (EPRI report 1007822). Attendees were invited to exchange relevant information on the effects of irradiation on austenitic materials in light water reactors and to produce recommendations for further work. EPRI reviewed the data, recommendations, and conclusions derived from the workshop and developed prioritized test matrices defining new data needs. Proposals were solicited, and selected proposals are the basis for the program described in this report. Results The planned test matrix for fracture toughness testing includes 21 tests on 5 materials.

Development of crawler type device using new measuring system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maruyama, T.; Sasaki, T.; Yagi, T.

1995-08-01

This paper reports the development and field application of a new device which examine shell to shell weld joints of RPV. In a BWR type nuclear power plant, there is narrow space around the Reactor Pressure Vessel (RPV) because RPV is enclosed by the Reactor Shield Wall (RSW) and thermal insulations. The developed device is characterized by a new position measuring system and magnet wheels for driving. The new position measuring system uses laser beam and ultrasonic wave. The magnet wheels make the device travel freely in the narrow space between RPV and insulation. This device is tested on mock-upsmore » and applied examination of RPVs to verify field applicability.« less

Core power and decay time limits for a disabled LOFT ECCS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Atkinson, S.A.

1978-01-09

An analysis was done to determine at what LOFT total core power (nuclear plus decay power) the ECCS could be inoperable. The criteria used for the analysis was that the maximum fuel clad temperature should not exceed 1650/sup 0/F given a loss of coolant. Calculations for natural convection cooling of the fuel by air with an inlet temperature of 580/sup 0/F determined that the limiting core power is 25 kW (discounted by 15 percent to 20 percent for potential uncertainties). Shutdown times are listed for when the LOFT ECCS can be safely bypassed or disabled.

Principled, Transformational Leadership: Analyzing the Discourse of Leadership in the Development of Librarianship's Core Competences

ERIC Educational Resources Information Center

Hicks, Deborah; Given, Lisa M.

2013-01-01

Using discourse analysis, this article explores three questions: (a) Why was "principled, transformational leadership" the leadership style added to Core Competences? (b) What was the discourse of leadership in the profession surrounding the development of the Core Competences? (c) How might this competence affect LIS education? And what measures,…

Representation and Analysis of Chemistry Core Ideas in Science Education Standards between China and the United States

ERIC Educational Resources Information Center

Wan, Yanlan; Bi, Hualin

2016-01-01

Chemistry core ideas play an important role in students' chemistry learning. On the basis of the representations of chemistry core ideas about "substances" and "processes" in the Chinese Chemistry Curriculum Standards (CCCS) and the U.S. Next Generation Science Standards (NGSS), we conduct a critical comparison of chemistry…

The Influence of the Core Practices Movement on the Teaching and Perspectives of Novice Teacher Educators

ERIC Educational Resources Information Center

DeMink-Carthew, Jessica; Grove, Rebecca; Peterson, Margaret

2017-01-01

This collaborative self-study examines the influence of engagement in the core practices movement on the course designs, instruction, and perspectives of three novice teacher educators at a large mid-Atlantic research university. Through core practices work, we integrated repeated cycles of analysis, practice, and reflection into our courses,…

Connecting Core Percolation and Controllability of Complex Networks

PubMed Central

Jia, Tao; Pósfai, Márton

2014-01-01

Core percolation is a fundamental structural transition in complex networks related to a wide range of important problems. Recent advances have provided us an analytical framework of core percolation in uncorrelated random networks with arbitrary degree distributions. Here we apply the tools in analysis of network controllability. We confirm analytically that the emergence of the bifurcation in control coincides with the formation of the core and the structure of the core determines the control mode of the network. We also derive the analytical expression related to the controllability robustness by extending the deduction in core percolation. These findings help us better understand the interesting interplay between the structural and dynamical properties of complex networks. PMID:24946797

Numerical analysis and experiment to identify origin of buckling in rapid cycling synchrotron core

NASA Astrophysics Data System (ADS)

Morita, Y.; Kageyama, T.; Akoshima, M.; Torizuka, S.; Tsukamoto, M.; Yamashita, S.; Yoshikawa, N.

2013-11-01

The accelerating cavities used in the rapid cycling synchrotron (RCS) of the Japan Proton Accelerator Research Complex (J-PARC) are loaded with magnetic alloy (MA) cores. Over lengthly periods of RCS operation, significant reductions in the impedance of the cavities resulting from the buckling of the cores were observed. A series of thermal structural simulations and compressive strength tests showed that the buckling can be attributed to the low-viscosity epoxy resin impregnation of the MA core that causes the stiffening of the originally flexible MA-ribbon-wound core. Our results showed that thermal stress can be effectively reduced upon using a core that is not epoxy-impregnated.

Qualitative Analysis of Common Definitions for Core Advanced Pharmacy Practice Experiences

PubMed Central

Danielson, Jennifer; Weber, Stanley S.

2014-01-01

Objective. To determine how colleges and schools of pharmacy interpreted the Accreditation Council for Pharmacy Education’s (ACPE’s) Standards 2007 definitions for core advanced pharmacy practice experiences (APPEs), and how they differentiated community and institutional practice activities for introductory pharmacy practice experiences (IPPEs) and APPEs. Methods. A cross-sectional, qualitative, thematic analysis was done of survey data obtained from experiential education directors in US colleges and schools of pharmacy. Open-ended responses to invited descriptions of the 4 core APPEs were analyzed using grounded theory to determine common themes. Type of college or school of pharmacy (private vs public) and size of program were compared. Results. Seventy-one schools (72%) with active APPE programs at the time of the survey responded. Lack of strong frequent themes describing specific activities for the acute care/general medicine core APPE indicated that most respondents agreed on the setting (hospital or inpatient) but the student experience remained highly variable. Themes were relatively consistent between public and private institutions, but there were differences across programs of varying size. Conclusion. Inconsistencies existed in how colleges and schools of pharmacy defined the core APPEs as required by ACPE. More specific descriptions of core APPEs would help to standardize the core practice experiences across institutions and provide an opportunity for quality benchmarking. PMID:24954931

SAXS analysis of single- and multi-core iron oxide magnetic nanoparticles

PubMed Central

Szczerba, Wojciech; Costo, Rocio; Morales, Maria del Puerto; Thünemann, Andreas F.

2017-01-01

This article reports on the characterization of four superparamagnetic iron oxide nanoparticles stabilized with dimercaptosuccinic acid, which are suitable candidates for reference materials for magnetic properties. Particles p1 and p2 are single-core particles, while p3 and p4 are multi-core particles. Small-angle X-ray scattering analysis reveals a lognormal type of size distribution for the iron oxide cores of the particles. Their mean radii are 6.9 nm (p1), 10.6 nm (p2), 5.5 nm (p3) and 4.1 nm (p4), with narrow relative distribution widths of 0.08, 0.13, 0.08 and 0.12. The cores are arranged as a clustered network in the form of dense mass fractals with a fractal dimension of 2.9 in the multi-core particles p3 and p4, but the cores are well separated from each other by a protecting organic shell. The radii of gyration of the mass fractals are 48 and 44 nm, and each network contains 117 and 186 primary particles, respectively. The radius distributions of the primary particle were confirmed with transmission electron microscopy. All particles contain purely maghemite, as shown by X-ray absorption fine structure spectroscopy. PMID:28381973

Preparation, process optimization and characterization of core-shell polyurethane/chitosan nanofibers as a potential platform for bioactive scaffolds.

PubMed

Maleknia, Laleh; Dilamian, Mandana; Pilehrood, Mohammad Kazemi; Sadeghi-Aliabadi, Hojjat; Hekmati, Amir Houshang

2018-06-01

In this paper, polyurethane (PU), chitosan (Cs)/polyethylene oxide (PEO), and core-shell PU/Cs nanofibers were produced at the optimal processing conditions using electrospinning technique. Several methods including SEM, TEM, FTIR, XRD, DSC, TGA and image analysis were utilized to characterize these nanofibrous structures. SEM images exhibited that the core-shell PU/Cs nanofibers were spun without any structural imperfections at the optimized processing conditions. TEM image confirmed the PU/Cs core-shell nanofibers were formed apparently. It that seems the inclusion of Cs/PEO to the shell, did not induce the significant variations in the crystallinity in the core-shell nanofibers. DSC analysis showed that the inclusion of Cs/PEO led to the glass temperature of the composition increased significantly compared to those of neat PU nanofibers. The thermal degradation of core-shell PU/Cs was similar to PU nanofibers degradation due to the higher PU concentration compared to other components. It was hypothesized that the core-shell PU/Cs nanofibers can be used as a potential platform for the bioactive scaffolds in tissue engineering. Further biological tests should be conducted to evaluate this platform as a three dimensional scaffold with the capabilities of releasing the bioactive molecules in a sustained manner.

Tank 241-T-204, core 188 analytical results for the final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nuzum, J.L.

TANK 241-T-204, CORE 188, ANALYTICAL RESULTS FOR THE FINAL REPORT. This document is the final laboratory report for Tank 241 -T-204. Push mode core segments were removed from Riser 3 between March 27, 1997, and April 11, 1997. Segments were received and extruded at 222-8 Laboratory. Analyses were performed in accordance with Tank 241-T-204 Push Mode Core Sampling and analysis Plan (TRAP) (Winkleman, 1997), Letter of instruction for Core Sample Analysis of Tanks 241-T-201, 241- T-202, 241-T-203, and 241-T-204 (LAY) (Bell, 1997), and Safety Screening Data Qual@ Objective (DO) ODukelow, et al., 1995). None of the subsamples submitted for totalmore » alpha activity (AT) or differential scanning calorimetry (DC) analyses exceeded the notification limits stated in DO. The statistical results of the 95% confidence interval on the mean calculations are provided by the Tank Waste Remediation Systems Technical Basis Group and are not considered in this report.« less

Group delay spread analysis of coupled-multicore fibers: A comparison between weak and tight bending conditions

NASA Astrophysics Data System (ADS)

Fujisawa, Takeshi; Saitoh, Kunimasa

2017-06-01

Group delay spread of coupled three-core fiber is investigated based on coupled-wave theory. The differences between supermode and discrete core mode models are thoroughly investigated to reveal applicability of both models for specific fiber bending condition. A macrobending with random twisting is taken into account for random modal mixing in the fiber. It is found that for weakly bent condition, both supermode and discrete core mode models are applicable. On the other hand, for strongly bent condition, the discrete core mode model should be used to account for increased differential modal group delay for the fiber without twisting and short correlation length, which were experimentally observed recently. Results presented in this paper indicate the discrete core mode model is superior to the supermode model for the analysis of coupled-multicore fibers for various bent condition. Also, for estimating GDS of coupled-multicore fiber, it is critically important to take into account the fiber bending condition.

Analysis of Radionuclide Releases from the Fukushima Dai-Ichi Nuclear Power Plant Accident Part I

NASA Astrophysics Data System (ADS)

Le Petit, G.; Douysset, G.; Ducros, G.; Gross, P.; Achim, P.; Monfort, M.; Raymond, P.; Pontillon, Y.; Jutier, C.; Blanchard, X.; Taffary, T.; Moulin, C.

2014-03-01

Part I of this publication deals with the analysis of fission product releases consecutive to the Fukushima Dai-ichi accident. Reactor core damages are assessed relying on radionuclide detections performed by the CTBTO radionuclide network, especially at the particulate station located at Takasaki, 210 km away from the nuclear power plant. On the basis of a comparison between the reactor core inventory at the time of reactor shutdowns and the fission product activities measured in air at Takasaki, especially 95Nb and 103Ru, it was possible to show that the reactor cores were exposed to high temperature for a prolonged time. This diagnosis was confirmed by the presence of 113Sn in air at Takasaki. The 133Xe assessed release at the time of reactor shutdown (8 × 1018 Bq) turned out to be in the order of 80 % of the amount deduced from the reactor core inventories. This strongly suggests a broad meltdown of reactor cores.

Fields of Coal: An analysis of industry and sedimentology in Dolores, Texas

NASA Astrophysics Data System (ADS)

Oaden, A.; Besonen, M. R.

2013-12-01

Research was conducted on a historically significant pond located in the former mining town of Dolores, located north of Laredo, Texas. The intention of this work was, to determine the influence of local mining operations on the environment and determine the extent of coal production from the sedimentary record. The pond is located ~160 m downslope from a former coal mine and waste pile, and was therefore a likely site of coal accumulations, as well as other debris. Additionally, this pond was created only 130 years ago, in 1882, giving a distinct time frame for any sedimentary records. Field work was conducted to obtain sediment core samples from the pond, and corroborating evidence was gathered using historical documents from archives in Laredo, online resources, as well as library records and inter library loan. Sedimentary cores obtained were shorter than desired as a result of the densely packed clay, which reduceding the penetration of coring equipment, leaving the historical extent of the cores limited. The limited sedimentary record also gives little indication as to the extent of production in the nearby mine and how it may have varied over time. The split cores were scanned with a Minolta CM-2600d spectrophotometer, and the results were transformed into first derivative spectrum equivalent data to identify common sedimentary minerals according to their first derivative signatures. The spectral analysis on the cores determined a large amount of clay minerals, and also limonite/goethite according to prominent first derivative peaks centered on ~440 and 540 nm. This agrees with visual observations given the all minerals showing spectra most intense in the 625 -725 nm portion of the visible spectrum, giving the cores their largely yellowish-reddish/brown hue of the cores. Magnetic susceptibility analysis indicated great changes in mineral contentmagnetism, some possibly associated with ash from fires. Bulk density and loss-on-ignition analysis to further characterize the sediments is underway. Basic conclusions indicate the present environment to be minimally affected by the coal operations and resulting tipple pile, but with a large variance over time in mineralogy and composition of sediment, with further research necessary to determine the full effects of industry in the area.

Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

NASA Technical Reports Server (NTRS)

Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

1991-01-01

The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

Evaluation of Counter-Based Dynamic Load Balancing Schemes for Massive Contingency Analysis on Over 10,000 Cores

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Yousu; Huang, Zhenyu; Rice, Mark J.

Contingency analysis studies are necessary to assess the impact of possible power system component failures. The results of the contingency analysis are used to ensure the grid reliability, and in power market operation for the feasibility test of market solutions. Currently, these studies are performed in real time based on the current operating conditions of the grid with a set of pre-selected contingency list, which might result in overlooking some critical contingencies caused by variable system status. To have a complete picture of a power grid, more contingencies need to be studied to improve grid reliability. High-performance computing techniques holdmore » the promise of being able to perform the analysis for more contingency cases within a much shorter time frame. This paper evaluates the performance of counter-based dynamic load balancing schemes for a massive contingency analysis program on 10,000+ cores. One million N-2 contingency analysis cases with a Western Electricity Coordinating Council power grid model have been used to demonstrate the performance. The speedup of 3964 with 4096 cores and 7877 with 10240 cores are obtained. This paper reports the performance of the load balancing scheme with a single counter and two counters, describes disk I/O issues, and discusses other potential techniques for further improving the performance.« less

Evaluating genetic diversity and constructing core collections of Chinese Lentinula edodes cultivars using ISSR and SRAP markers.

PubMed

Liu, Jun; Wang, Zhuo-Ren; Li, Chuang; Bian, Yin-Bing; Xiao, Yang

2015-06-01

Genetic diversity among 89 Chinese Lentinula edodes cultivars was analyzed by inter-simple sequence repeat (ISSR) and sequence-related amplified polymorphism (SRAP) markers. A 123 out of 126 ISSR loci (97.62%) and 108 out of 129 SRAP loci (83.73%) were polymorphic between two or more strains. A dendrogram constructed by cluster analysis based on the ISSR and SRAP markers separated the L. edodes strains into two major groups, of which group B was further divided into five subgroups. Clustering results also showed a positive correlation with the main agronomic traits of the strains, and that strains with similar traits clustered together into the same groups or subgroups in most cases. The average coefficient of pairwise genetic similarity was 0.820 (range: 0.576-0.988). Compared to the wild strains, Chinese L. edodes cultivars indicated a lower level of genetic diversity. Two preliminary core collections of L. edodes, Core1 and Core2, were established based on the ISSR and SRAP data, respectively. Core1 was constructed by the advanced M (maximization) strategy using the PowerCore version 1.0 software and contained 21 strains, whereas Core2 was created by the allele preferred sampling strategy using the cluster method and contained 18 strains. Both core collections were highly representative of the genetic diversity of the original germplasm, as confirmed by the values of Na (observed number of alleles), Ne (effective number of alleles), H (Nei's gene diversity) and I (Shannon's information index), as well as results of principal coordinate analysis. The loci retention ratio of Core1 (99.61%) was higher than that of Core2 (97.65%). Moreover, Core1 contained strains with more types of agronomic traits than those in Core2. This study builds the basis for further effective protection, management and use of L. edodes germplasm resource. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

CLEAR: Cross-Layer Exploration for Architecting Resilience

DTIC Science & Technology

2017-03-01

benchmark analysis, also provides cost-effective solutions (~1% additional energy cost for the same 50× improvement). This paper addresses the...core (OoO-core) [Wang 04], across 18 benchmarks . Such extensive exploration enables us to conclusively answer the above cross-layer resilience...analysis of the effects of soft errors on application benchmarks , provides a highly effective soft error resilience approach. 3. The above

Entrapment of carbon dioxide with chitosan-based core-shell particles containing changeable cores.

PubMed

Dong, Yanrui; Fu, Yinghao; Lin, Xia; Xiao, Congming

2016-08-01

Water-soluble chitosan-based core-shell particles that contained changeable cores were successfully applied to anchor carbon dioxide. The entrapment capacity of the particles for carbon dioxide (EC) depended on the cores. It was found that EC of the particles contained aqueous cores was higher than that of the beads with water-soluble chitosan gel cores, which was confirmed with thermogravimetric analysis. In addition, calcium ions and sodium hydroxide were introduced within the particles to examine their effect on the entrapment. EC of the particles was enhanced with sodium hydroxide when the cores were WSC gel. The incorporation of calcium ions was helpful for stabilizing carbon dioxide through the formation of calcium carbonate, which was verified with Fourier transform infrared spectra and scanning electron microscopy/energy-dispersive spectrometry. This phenomenon meant the role of calcium ions for fixating carbon dioxide was significant. Copyright © 2016 Elsevier B.V. All rights reserved.

Diagnostic accuracy of 22/25-gauge core needle in endoscopic ultrasound-guided sampling: systematic review and meta-analysis.

PubMed

Oh, Hyoung-Chul; Kang, Hyun; Lee, Jae Young; Choi, Geun Joo; Choi, Jung Sik

2016-11-01

To compare the diagnostic accuracy of endoscopic ultrasound-guided core needle aspiration with that of standard fine-needle aspiration by systematic review and meta-analysis. Studies using 22/25-gauge core needles, irrespective of comparison with standard fine needles, were comprehensively reviewed. Pooled sensitivity, specificity, diagnostic odds ratio (DOR), and summary receiver operating characteristic curves for the diagnosis of malignancy were used to estimate the overall diagnostic efficiency. The pooled sensitivity, specificity, and DOR of the core needle for the diagnosis of malignancy were 0.88 (95% confidence interval [CI], 0.84 to 0.90), 0.99 (95% CI, 0.96 to 1), and 167.37 (95% CI, 65.77 to 425.91), respectively. The pooled sensitivity, specificity, and DOR of the standard needle were 0.84 (95% CI, 0.79 to 0.88), 1 (95% CI, 0.97 to 1), and 130.14 (95% CI, 34.00 to 495.35), respectively. The area under the curve of core and standard needle in the diagnosis of malignancy was 0.974 and 0.955, respectively. The core and standard needle were comparable in terms of pancreatic malignancy diagnosis. There was no significant difference in procurement of optimal histologic cores between core and standard needles (risk ratio [RR], 0.545; 95% CI, 0.187 to 1.589). The number of needle passes for diagnosis was significantly lower with the core needle (standardized mean difference, -0.72; 95% CI, -1.02 to -0.41). There were no significant differences in overall complications (RR, 1.26; 95% CI, 0.34 to 4.62) and technical failure (RR, 5.07; 95% CI, 0.68 to 37.64). Core and standard needles were comparable in terms of diagnostic accuracy, technical performance, and safety profile.

The effects of substrate size, surface area, and density on coat thickness of multi-particulate dosage forms.

PubMed

Heinicke, Grant; Matthews, Frank; Schwartz, Joseph B

2005-01-01

Drugs layering experiments were performed in a fluid bed fitted with a rotor granulator insert using diltiazem as a model drug. The drug was applied in various quantities to sugar spheres of different mesh sizes to give a series of drug-layered sugar spheres (cores) of different potency, size, and weight per particle. The drug presence lowered the bulk density of the cores in proportion to the quantity of added drug. Polymer coating of each core lot was performed in a fluid bed fitted with a Wurster insert. A series of polymer-coated cores (pellets) was removed from each coating experiment. The mean diameter of each core and each pellet sample was determined by image analysis. The rate of change of diameter on polymer addition was determined for each starting size of core and compared to calculated values. The core diameter was displaced from the line of best fit through the pellet diameter data. Cores of different potency with the same size distribution were made by layering increasing quantities of drug onto sugar spheres of decreasing mesh size. Equal quantities of polymer were applied to the same-sized core lots and coat thickness was measured. Weight/weight calculations predict equal coat thickness under these conditions, but measurable differences were found. Simple corrections to core charge weight in the Wurster insert were successfully used to manufacture pellets having the same coat thickness. The sensitivity of the image analysis technique in measuring particle size distributions (PSDs) was demonstrated by measuring a displacement in PSD after addition of 0.5% w/w talc to a pellet sample.

Gains in efficiency and scientific potential of continental climate reconstruction provided by the LRC LacCore Facility, University of Minnesota

NASA Astrophysics Data System (ADS)

Noren, A.; Brady, K.; Myrbo, A.; Ito, E.

2007-12-01

Lacustrine sediment cores comprise an integral archive for the determination of continental paleoclimate, for their potentially high temporal resolution and for their ability to resolve spatial variability in climate across vast sections of the globe. Researchers studying these archives now have a large, nationally-funded, public facility dedicated to the support of their efforts. The LRC LacCore Facility, funded by NSF and the University of Minnesota, provides free or low-cost assistance to any portion of research projects, depending on the specific needs of the project. A large collection of field equipment (site survey equipment, coring devices, boats/platforms, water sampling devices) for nearly any lacustrine setting is available for rental, and Livingstone-type corers and drive rods may be purchased. LacCore staff can accompany field expeditions to operate these devices and curate samples, or provide training prior to device rental. The Facility maintains strong connections to experienced shipping agents and customs brokers, which vastly improves transport and importation of samples. In the lab, high-end instrumentation (e.g., multisensor loggers, high-resolution digital linescan cameras) provides a baseline of fundamental analyses before any sample material is consumed. LacCore staff provide support and training in lithological description, including smear-slide, XRD, and SEM analyses. The LRC botanical macrofossil reference collection is a valuable resource for both core description and detailed macrofossil analysis. Dedicated equipment and space for various subsample analyses streamlines these endeavors; subsamples for several analyses may be submitted for preparation or analysis by Facility technicians for a fee (e.g., carbon and sulfur coulometry, grain size, pollen sample preparation and analysis, charcoal, biogenic silica, LOI, freeze drying). The National Lacustrine Core Repository now curates ~9km of sediment cores from expeditions around the world, and stores metadata and analytical data for all cores processed at the facility. Any researcher may submit sample requests for material in archived cores. Supplies for field (e.g., polycarbonate pipe, endcaps), lab (e.g., sample containers, pollen sample spike), and curation (e.g., D-tubes) are sold at cost. In collaboration with facility users, staff continually develop new equipment, supplies, and procedures as needed in order to provide the best and most comprehensive set of services to the research community.

Determination and Variation of Core Bacterial Community in a Two-Stage Full-Scale Anaerobic Reactor Treating High-Strength Pharmaceutical Wastewater.

PubMed

Ma, Haijun; Ye, Lin; Hu, Haidong; Zhang, Lulu; Ding, Lili; Ren, Hongqiang

2017-10-28

Knowledge on the functional characteristics and temporal variation of anaerobic bacterial populations is important for better understanding of the microbial process of two-stage anaerobic reactors. However, owing to the high diversity of anaerobic bacteria, close attention should be prioritized to the frequently abundant bacteria that were defined as core bacteria and putatively functionally important. In this study, using MiSeq sequencing technology, the core bacterial community of 98 operational taxonomic units (OTUs) was determined in a two-stage upflow blanket filter reactor treating pharmaceutical wastewater. The core bacterial community accounted for 61.66% of the total sequences and accurately predicted the sample location in the principal coordinates analysis scatter plot as the total bacterial OTUs did. The core bacterial community in the first-stage (FS) and second-stage (SS) reactors were generally distinct, in that the FS core bacterial community was indicated to be more related to a higher-level fermentation process, and the SS core bacterial community contained more microbes in syntrophic cooperation with methanogens. Moreover, the different responses of the FS and SS core bacterial communities to the temperature shock and influent disturbance caused by solid contamination were fully investigated. Co-occurring analysis at the Order level implied that Bacteroidales, Selenomonadales, Anaerolineales, Syneristales, and Thermotogales might play key roles in anaerobic digestion due to their high abundance and tight correlation with other microbes. These findings advance our knowledge about the core bacterial community and its temporal variability for future comparative research and improvement of the two-stage anaerobic system operation.

Structural, evolutionary and genetic analysis of the histidine biosynthetic "core" in the genus Burkholderia.

PubMed

Papaleo, Maria Cristiana; Russo, Edda; Fondi, Marco; Emiliani, Giovanni; Frandi, Antonio; Brilli, Matteo; Pastorelli, Roberta; Fani, Renato

2009-12-01

In this work a detailed analysis of the structure, the expression and the organization of his genes belonging to the core of histidine biosynthesis (hisBHAF) in 40 newly determined and 13 available sequences of Burkholderia strains was carried out. Data obtained revealed a strong conservation of the structure and organization of these genes through the entire genus. The phylogenetic analysis showed the monophyletic origin of this gene cluster and indicated that it did not undergo horizontal gene transfer events. The analysis of the intergenic regions, based on the substitution rate, entropy plot and bendability suggested the existence of a putative transcription promoter upstream of hisB, that was supported by the genetic analysis that showed that this cluster was able to complement Escherichia colihisA, hisB, and hisF mutations. Moreover, a preliminary transcriptional analysis and the analysis of microarray data revealed that the expression of the his core was constitutive. These findings are in agreement with the fact that the entire Burkholderiahis operon is heterogeneous, in that it contains "alien" genes apparently not involved in histidine biosynthesis. Besides, they also support the idea that the proteobacterial his operon was piece-wisely assembled, i.e. through accretion of smaller units containing only some of the genes (eventually together with their own promoters) involved in this biosynthetic route. The correlation existing between the structure, organization and regulation of his "core" genes and the function(s) they perform in cellular metabolism is discussed.

The MARTE VNIR imaging spectrometer experiment: design and analysis.

PubMed

Brown, Adrian J; Sutter, Brad; Dunagan, Stephen

2008-10-01

We report on the design, operation, and data analysis methods employed on the VNIR imaging spectrometer instrument that was part of the Mars Astrobiology Research and Technology Experiment (MARTE). The imaging spectrometer is a hyperspectral scanning pushbroom device sensitive to VNIR wavelengths from 400-1000 nm. During the MARTE project, the spectrometer was deployed to the Río Tinto region of Spain. We analyzed subsets of three cores from Río Tinto using a new band modeling technique. We found most of the MARTE drill cores to contain predominantly goethite, though spatially coherent areas of hematite were identified in Core 23. We also distinguished non Fe-bearing minerals that were subsequently analyzed by X-ray diffraction (XRD) and found to be primarily muscovite. We present drill core maps that include spectra of goethite, hematite, and non Fe-bearing minerals.

The MARTE VNIR Imaging Spectrometer Experiment: Design and Analysis

NASA Astrophysics Data System (ADS)

Brown, Adrian J.; Sutter, Brad; Dunagan, Stephen

2008-10-01

We report on the design, operation, and data analysis methods employed on the VNIR imaging spectrometer instrument that was part of the Mars Astrobiology Research and Technology Experiment (MARTE). The imaging spectrometer is a hyperspectral scanning pushbroom device sensitive to VNIR wavelengths from 400-1000 nm. During the MARTE project, the spectrometer was deployed to the Río Tinto region of Spain. We analyzed subsets of three cores from Río Tinto using a new band modeling technique. We found most of the MARTE drill cores to contain predominantly goethite, though spatially coherent areas of hematite were identified in Core 23. We also distinguished non Fe-bearing minerals that were subsequently analyzed by X-ray diffraction (XRD) and found to be primarily muscovite. We present drill core maps that include spectra of goethite, hematite, and non Fe-bearing minerals.

Quantification of oil and water in preserved reservoir rock by NMR spectroscopy and imaging.

PubMed

Davies, S; Hardwick, A; Roberts, D; Spowage, K; Packer, K J

1994-01-01

Reservoir rock analysis by proton NMR requires separation of the response into brine and crude oil components. Tests on preserved core from a North Sea chalk reservoir show that spin-lattice relaxation time distributions can be used to distinguish the two fluids. NMR estimates of oil and water saturations for 1.5" diameter core examined in a 10 MHz Bruker Minispec spectrometer closely match fluid contents determined by distillation. The spin-lattice relaxation contrast mechanism developed for core samples can be applied in the quantitative analysis of NMR images. The relaxation data are compared with data from chemical shift imaging on the same core sample. The results indicate that it will be possible to monitor changes in fluid distributions, in this and similar systems, under dynamic conditions such as in a waterflood.

Photoreactive, core-shell cross-linked/hollow microspheres prepared by delayed addition of cross-linker in dispersion polymerization for antifouling and immobilization of protein.

PubMed

Wang, Shengliu; Yue, Kai; Liu, Lianying; Yang, Wantai

2013-01-01

When dispersion polymerization of styrene (St) had run for 3h, after particle rapidly growing stage, 4,4'-dimethacryloyloxybenzophenone (DMABP) cross-linker was added to reaction system and photoreactive, core(PSt)-shell(Poly(St-co-DMABP)) particles with rich benzophenone (BP) groups on surface were prepared. Polymerization of DMABP could occurred mainly on the preformed core of PSt because its diffusion could be impeded by (1) compactness of particles formed at the moment of cross-linker addition (more than 80% of monomer had been consumed, particles were no longer fully swollen by monomer), (2) reduced polarity of continuous phase, and (3) immediate occurrence of cross-linking. Subsequently, photoreactive, cross-linked hollow particles were yielded by removal of uncross-linked core in THF. SEM and TEM observation demonstrated the formation of core-shell structure and improvement of shell thickness when DMABP content increased. UV-vis spectra analysis on polymer dissolved in THF indicated that there is no polymer of DMABP in core. FTIR spectra analysis and XPS measurement further revealed that BP component on particle surface was enriched when amount of DMABP increased. Finally, an anti-fouling polymer (poly (ethylene glycol), PEG) and protein of mouse IgG was immobilized on particle surface under UV irradiation, as confirmed by FTIR spectra analysis, SEM observation and TMB color reaction. Crown Copyright © 2012. Published by Elsevier Inc. All rights reserved.

Monitoring of NMR porosity changes in the full-size core salvage through the drying process

NASA Astrophysics Data System (ADS)

Fattakhov, Artur; Kosarev, Victor; Doroginitskii, Mikhail; Skirda, Vladimir

2015-04-01

Currently the principle of nuclear magnetic resonance (NMR) is one of the most popular technologies in the field of borehole geophysics and core analysis. Results of NMR studies allow to calculate the values of the porosity and permeability of sedimentary rocks with sufficient reliability. All standard tools for the study of core salvage on the basis of NMR have significant limitations: there is considered only long relaxation times corresponding to the mobile formation fluid. Current trends in energy obligate to move away from conventional oil to various alternative sources of energy. One of these sources are deposits of bitumen and high-viscosity oil. In Kazan (Volga Region) Federal University (Russia) there was developed a mobile unit for the study of the full-length core salvage by the NMR method ("NMR-Core") together with specialists of "TNG-Group" (a company providing maintenance services to oil companies). This unit is designed for the study of core material directly on the well, after removing it from the core receiver. The maximum diameter of the core sample may be up to 116 mm, its length (or length of the set of samples) may be up to 1000 mm. Positional precision of the core sample relative to the measurement system is 1 mm, and the spatial resolution along the axis of the core is 10 mm. Acquisition time of the 1 m core salvage varies depending on the mode of research and is at least 20 minutes. Furthermore, there is implemented a special investigation mode of the core samples with super small relaxation times (for example, heavy oil) is in the tool. The aim of this work is tracking of the NMR porosity changes in the full-size core salvage in time. There was used a water-saturated core salvage from the shallow educational well as a sample. The diameter of the studied core samples is 93 mm. There was selected several sections length of 1m from the 200-meter coring interval. The studied core samples are being measured several times. The time interval between the measurements is from 1 hour to 48 hours. Making the measurements it possible to draw conclusions about that the processes of NMR porosity changes in time as a result of evaporation of the part of fluid from the surface layer of the core salvage and suggest a core analysis technique directly on the well. This work is supported by the grant of Ministry of Education and Science of the Russian Federation (project No. 02.G25.31.0029).

Optical Methods for Identifying Hard Clay Core Samples During Petrophysical Studies

NASA Astrophysics Data System (ADS)

Morev, A. V.; Solovyeva, A. V.; Morev, V. A.

2018-01-01

X-ray phase analysis of the general mineralogical composition of core samples from one of the West Siberian fields was performed. Electronic absorption spectra of the clay core samples with an added indicator were studied. The speed and availability of applying the two methods in petrophysical laboratories during sample preparation for standard and special studies were estimated.

Methods for the Compilation of a Core List of Journals in Toxicology.

ERIC Educational Resources Information Center

Kuch, T. D. C.

Previously reported methods for the compilation of core lists of journals in multidisciplinary areas are first examined, with toxicology used as an example of such an area. Three approaches to the compilation of a core list of journals in toxicology were undertaken and the results analyzed with the aid of models. Analysis of the results of the…

Modal analysis and acoustic transmission through offset-core honeycomb sandwich panels

NASA Astrophysics Data System (ADS)

Mathias, Adam Dustin

The work presented in this thesis is motivated by an earlier research that showed that double, offset-core honeycomb sandwich panels increased thermal resistance and, hence, decreased heat transfer through the panels. This result lead to the hypothesis that these panels could be used for acoustic insulation. Using commercial finite element modeling software, COMSOL Multiphysics, the acoustical properties, specifically the transmission loss across a variety of offset-core honeycomb sandwich panels, is studied for the case of a plane acoustic wave impacting the panel at normal incidence. The transmission loss results are compared with those of single-core honeycomb panels with the same cell sizes. The fundamental frequencies of the panels are also computed in an attempt to better understand the vibrational modes of these particular sandwich-structured panels. To ensure that the finite element analysis software is adequate for the task at hand, two relevant benchmark problems are solved and compared with theory. Results from these benchmark results compared well to those obtained from theory. Transmission loss results from the offset-core honeycomb sandwich panels show increased transmission loss, especially for large cell honeycombs when compared to single-core honeycomb panels.

Bose-Einstein condensate & degenerate Fermi cored dark matter halos

NASA Astrophysics Data System (ADS)

Chung, W.-J.; Nelson, L. A.

2018-06-01

There has been considerable interest in the last several years in support of the idea that galaxies and clusters could have highly condensed cores of dark matter (DM) within their central regions. In particular, it has been suggested that dark matter could form Bose-Einstein condensates (BECs) or degenerate Fermi cores. We examine these possibilities under the assumption that the core consists of highly condensed DM (either bosons or fermions) that is embedded in a diffuse envelope (e.g., isothermal sphere). The novelty of our approach is that we invoke composite polytropes to model spherical collisionless structures in a way that is physically intuitive and can be generalized to include other equations of state (EOSs). Our model is very amenable to the analysis of BEC cores (composed of ultra-light bosons) that have been proposed to resolve small-scale CDM anomalies. We show that the analysis can readily be applied to bosons with or without small repulsive self-interactions. With respect to degenerate Fermi cores, we confirm that fermionic particle masses between 1—1000 keV are not excluded by the observations. Finally, we note that this approach can be extended to include a wide range of EOSs in addition to multi-component collisionless systems.

Core-Shell Columns in High-Performance Liquid Chromatography: Food Analysis Applications

PubMed Central

Preti, Raffaella

2016-01-01

The increased separation efficiency provided by the new technology of column packed with core-shell particles in high-performance liquid chromatography (HPLC) has resulted in their widespread diffusion in several analytical fields: from pharmaceutical, biological, environmental, and toxicological. The present paper presents their most recent applications in food analysis. Their use has proved to be particularly advantageous for the determination of compounds at trace levels or when a large amount of samples must be analyzed fast using reliable and solvent-saving apparatus. The literature hereby described shows how the outstanding performances provided by core-shell particles column on a traditional HPLC instruments are comparable to those obtained with a costly UHPLC instrumentation, making this novel column a promising key tool in food analysis. PMID:27143972

Testing a laser-induced breakdown spectroscopy technique on the Arctic sediments

NASA Astrophysics Data System (ADS)

Han, D.; Nam, S. I.

2017-12-01

Physical and geochemical investigations coupled with the Laser-induced Breakdown Spectroscopy (LIBS) were performed on three surface sediment cores (ARA03B/24BOX, ARA02B/01(A)MUC, ARA02B/02MUC and ARA02B/03(A)MUC) recovered from the western Arctic Ocean (Chukchi Sea) during IBRV ARON expeditions in 2012. The LIBS technique was applied to carry out elemental chemical analysis of the Arctic sediments and compared with that measured by ITRAX X-ray fuorescence (XRF) core scanning. LIBS and XRF have shown similar elemental composition within each sediment core. In this study, mineral composition (XRD), grain size distribution and organic carbon content as well as elemental composition (LIBS) were all considered to understand paleoenvironmental changes (ocean circulation, sea-ice drift, iceberg discharge, and etc.) recorded in the Arctic Holocene sediment. Quantitative LIBS analysis shows a gradually varying distribution of the elements along the sampled core and clear separation between the cores. The cores are geochemically characterized by elevated Mn profile. The gradient of mineral composition and grain sizes among the cores shows regional distribution and variation in sedimentary condition due to geological distance between East Siberian and North America. The present study reveals that a LIBS technique can be employed for in-situ sediment analyses for the Arctic Ocean. Furthermore, LIBS does not require costly equipment, trained operators, and complicated sample pre-treatment processes compared to Atomic absorption spectroscopy (AAS) and inductively coupled plasma emission spectroscopy (ICP), and also known to show relatively high levels of sensitivity, precision, and distinction than XRF analysis, scanning electron microscopy-energy dispersive spectrometry (SEM-EDS), and electron probe X-ray microanalysis (EPMA).

Angiographic Validation of the American College of Cardiology Foundation–The Society of Thoracic Surgeons Collaboration on the Comparative Effectiveness of Revascularization Strategies Study

PubMed Central

Chakrabarti, Anjan K.; Grau-Sepulveda, Maria V.; O’Brien, Sean; Abueg, Cassandra; Ponirakis, Angelo; Delong, Elizabeth; Peterson, Eric; Klein, Lloyd W.; Garratt, Kirk N.; Weintraub, William S.; Gibson, C. Michael

2017-01-01

Background The goal of this study was to compare angiographic interpretation of coronary arteriograms by sites in community practice versus those made by a centralized angiographic core laboratory. Methods and Results The study population consisted of 2013 American College of Cardiology–National Cardiovascular Data Registry (ACC–NCDR) records with 2- and 3- vessel coronary disease from 54 sites in 2004 to 2007. The primary analysis compared Registry (NCDR)-defined 2- and 3-vessel disease versus those from an angiographic core laboratory analysis. Vessel-level kappa coefficients suggested moderate agreement between NCDR and core laboratory analysis, ranging from kappa=0.39 (95% confidence intervals, 0.32–0.45) for the left anterior descending artery to kappa=0.59 (95% confidence intervals, 0.55–0.64) for the right coronary artery. Overall, 6.3% (n=127 out of 2013) of those patients identified with multivessel disease at NCDR sites had had 0- or 1-vessel disease by core laboratory reading. There was no directional bias with regard to overcall, that is, 12.3% of cases read as 3-vessel disease by the sites were read as <3-vessel disease by the core laboratory, and 13.9% of core laboratory 3-vessel cases were read as <3-vessel by the sites. For a subset of patients with left main coronary disease, registry overcall was not linked to increased rates of mortality or myocardial infarction. Conclusions There was only modest agreement between angiographic readings in clinical practice and those from an independent core laboratory. Further study will be needed because the implications for patient management are uncertain. PMID:24496239

Angiographic validation of the American College of Cardiology Foundation-the Society of Thoracic Surgeons Collaboration on the Comparative Effectiveness of Revascularization Strategies study.

PubMed

Chakrabarti, Anjan K; Grau-Sepulveda, Maria V; O'Brien, Sean; Abueg, Cassandra; Ponirakis, Angelo; Delong, Elizabeth; Peterson, Eric; Klein, Lloyd W; Garratt, Kirk N; Weintraub, William S; Gibson, C Michael

2014-02-01

The goal of this study was to compare angiographic interpretation of coronary arteriograms by sites in community practice versus those made by a centralized angiographic core laboratory. The study population consisted of 2013 American College of Cardiology-National Cardiovascular Data Registry (ACC-NCDR) records with 2- and 3- vessel coronary disease from 54 sites in 2004 to 2007. The primary analysis compared Registry (NCDR)-defined 2- and 3-vessel disease versus those from an angiographic core laboratory analysis. Vessel-level kappa coefficients suggested moderate agreement between NCDR and core laboratory analysis, ranging from kappa=0.39 (95% confidence intervals, 0.32-0.45) for the left anterior descending artery to kappa=0.59 (95% confidence intervals, 0.55-0.64) for the right coronary artery. Overall, 6.3% (n=127 out of 2013) of those patients identified with multivessel disease at NCDR sites had had 0- or 1-vessel disease by core laboratory reading. There was no directional bias with regard to overcall, that is, 12.3% of cases read as 3-vessel disease by the sites were read as <3-vessel disease by the core laboratory, and 13.9% of core laboratory 3-vessel cases were read as <3-vessel by the sites. For a subset of patients with left main coronary disease, registry overcall was not linked to increased rates of mortality or myocardial infarction. There was only modest agreement between angiographic readings in clinical practice and those from an independent core laboratory. Further study will be needed because the implications for patient management are uncertain.

Structural and Geophysical Characterization of Oklahoma Basement

NASA Astrophysics Data System (ADS)

Morgan, C.; Johnston, C. S.; Carpenter, B. M.; Reches, Z.

2017-12-01

Oklahoma has experienced a large increase in seismicity since 2009 that has been attributed to wastewater injection. Most earthquakes, including four M5+ earthquakes, nucleated at depths > 4 km, well within the pre-Cambrian crystalline basement, even though wastewater injection occurred almost exclusively in the sedimentary sequence above. To better understand the structural characteristics of the rhyolite and granite that makeup the midcontinent basement, we analyzed a 150 m long core recovered from a basement borehole (Shads 4) in Rogers County, NE Oklahoma. The analysis of the fracture network in the rhyolite core included measurements of fracture inclination, aperture, and density, the examination fracture surface features and fill minerology, as well as x-ray diffraction analysis of secondary mineralization. We also analyzed the highly fractured and faulted segments of the core with a portable gamma-ray detector, magnetometer, and rebound hammer. The preliminary analysis of the fractures within the rhyolite core showed: (1) Fracture density increasing with depth by a factor of 10, from 4 fractures/10m in the upper core segment to 40 fracture/10m at 150 m deeper. (2) The fractures are primarily sub-vertical, inclined 10-20° from the axis of the vertical core. (3) The secondary mineralization is dominated by calcite and epidote. (4) Fracture aperture ranges from 0.35 to 2.35mm based on the thickness of secondary filling. (5) About 8% of the examined fractures display slickenside striations. (6) Increases of elasticity (by rebound hammer) and gamma-ray emissions are systematically correlated with a decrease in magnetic susceptibility in core segments of high fracture density and/or faulting; this observation suggests diagenetic fracture re-mineralization.

Vertical velocity in oceanic convection off tropical Australia

NASA Technical Reports Server (NTRS)

Lucas, Christopher; Zipser, Edward J.; Lemone, Margaret A.

1994-01-01

Time series of 1-Hz vertical velocity data collected during aircraft penetrations of oceanic cumulonimbus clouds over the western Pacific warm pool as part of the Equatorial Mesoscale Experiment (EMEX) are analyzed for updraft and downdraft events called cores. An updraft core is defined as occurring whenever the vertical velocity exceeds 1 m/sec for at least 500 m. A downdraft core is defined analogously. Over 19,000 km of straight and level flight legs are used in the analysis. Five hundred eleven updraft cores and 253 downdraft cores are included in the dataset. Core properties are summarized as distributions of average and maximum vertical velocity, diameter, and mass flux in four altitude intervals between 0.2 and 5.8 km. Distributions are approximately lognormal at all levels. Examination of the variation of the statistics with height suggests a maximum in vertical velocity between 2 and 3 km; slightly lower or equal vertical velocity is indicated at 5 km. Near the freezing level, virtual temperature deviations are found to be slightly positive for both updraft and downdraft cores. The excess in updraft cores is much smaller than that predicted by parcel theory. Comparisons with other studies that use the same analysis technique reveal that EMEX cores have approximately the same strength as cores of other oceanic areas, despite warmer sea surface temperatures. Diameter and mass flux are greater than those in the Global Atmospheric Research Program (GATE) but smaller than those in hurricane rainbands. Oceanic cores are much weaker and appear to be slightly smaller than those observed over land during the Thunderstorm Project. The markedly weaker oceanic vertical velocities below 5.8 km (compared to the continental cores) cannot be attributed to smaller total convective available potential energy or to very high water loading. Rather, it is suggested that water loading, although less than adiabatic, is more effective in reducing buoyancy of oceanic cores because of the smaller potential buoyancy below 5.8 km. Entrainment appears to be more effective in reducing buoyancy to well below adiabatic values in oceanic cores, a result consistent with the smaller oceanic core diameters in the lower cloud layer. It is speculated further that core diameters are related to boundary layer depth, which is clearly smaller over the oceans.

Characterization of carbon-14 generated by the nuclear power industry. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eabry, S.; Vance, J.N.; Cline, J.E.

1995-11-01

This report describes an evaluation of C-14 production rates in light-water reactors (LWRs) and characterization of its chemical speciation and environmental behavior. The study estimated the total production rate of the nuclide in operating PWRs and BWRs along with the assessment of the C-14 content of solid radwaste. The major source of production of C-14 in both PWR`s and BWRs was the activation of 0-17 in the water molecule and of N-14 dissolved in reactor coolant. The production of C-14 was estimated to range from 7 Ci/GW(e)-year to 11 Ci/GW(e)-year. The estimated range of the quantity of C-14 in LLWmore » was 1-2 Ci/ reactor-year which compares favorably with data obtained from shipping manifests. The environmental behavior of C-14 associated with low-level waste (LLW) disposal is greatly dependent upon its chemical speciation. This scoping study was performed to help identify the occurrence of inorganic and organic forms of C-14 in reactor coolant water and in primary coolant demineralization resins. These represent the major source for C-14 in LLW from nuclear power stations. Also, the behavior of inorganic and two of the organic forms of C-14 on soil uptake was determined by measuring distribution coefficients (Kd`s) on two soil types and a cement, using two different groundwater types. This study confirms that C-14 concentrations are significantly higher in the primary coolant from PWR stations compared to BWR stations. The C-14 followed trends of Co-60 generation during primary coolant demineralization at all but one of the stations examined. However, the C-14/Co-60 activity ratios measured by this study in resin samples through which samples of coolant were drawn were about 8 to 42 times higher than those reported for waste samples in the industry data base for PWR stations, and 15 to 730 times lower for the BWR stations.« less

3D modeling of missing pellet surface defects in BWR fuel

DOE PAGES

Spencer, B. W.; Williamson, R. L.; Stafford, D. S.; ...

2016-07-26

One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can bemore » used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed in this paper. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Finally, parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding adjacent to the defect.« less

Dynamic interaction between fetal adversity and a genetic score reflecting dopamine function on developmental outcomes at 36 months

PubMed Central

Pokhvisneva, Irina; Léger, Étienne; Gaudreau, Hélène; Steiner, Meir; Kennedy, James L.; O’Donnell, Kieran J.; Diorio, Josie; Meaney, Michael J.; Silveira, Patrícia P.

2017-01-01

Background Fetal adversity, evidenced by poor fetal growth for instance, is associated with increased risk for several diseases later in life. Classical cut-offs to characterize small (SGA) and large for gestational age (LGA) newborns are used to define long term vulnerability. We aimed at exploring the possible dynamism of different birth weight cut-offs in defining vulnerability in developmental outcomes (through the Bayley Scales of Infant and Toddler Development), using the example of a gene vs. fetal adversity interaction considering gene choices based on functional relevance to the studied outcome. Methods 36-month-old children from an established prospective birth cohort (Maternal Adversity, Vulnerability, and Neurodevelopment) were classified according to birth weight ratio (BWR) (SGA ≤0.85, LGA >1.15, exploring a wide range of other cut-offs) and genotyped for polymorphisms associated with dopamine signaling (TaqIA-A1 allele, DRD2-141C Ins/Ins, DRD4 7-repeat, DAT1-10- repeat, Met/Met-COMT), composing a score based on the described function, in which hypofunctional variants received lower scores. Results There were 251 children (123 girls and 128 boys). Using the classic cut-offs (0.85 and 1.15), there were no statistically significant interactions between the neonatal groups and the dopamine genetic score. However, when changing the cut-offs, it is possible to see ranges of BWR that could be associated with vulnerability to poorer development according to the variation in the dopamine function. Conclusion The classic birth weight cut-offs to define SGA and LGA newborns should be seen with caution, as depending on the outcome in question, the protocols for long-term follow up could be either too inclusive—therefore most costly, or unable to screen true vulnerabilities—and therefore ineffective to establish early interventions and primary prevention. PMID:28505190

Planktic foraminifer census data from Northwind Ridge cores PI-88-AP P3, PI-88-AR P7 and PI-88-AR P9, Arctic Ocean

USGS Publications Warehouse

Foley, Kevin M.; Poore, Richard Z.

1993-01-01

The U.S. Geological Survey recovered 9 piston cores from the Northwind Ridge in the Canada Basin of the Arctic Ocean from a cruise of the USCGC Polar Star during 1988. Preliminary analysis of the cores suggests sediments deposited on Northwind Ridge preserve a detailed record of glacial and interglacial cycles for the last few hundred-thousand to one million years. This report includes quantitative data on foraminifers and selected sediment size-fraction data in 98 samples from Northwind Ridge core PI-88AR P3, 51 samples from core PI-88-AR P7 and 117 samples from core PI-88-AR P9.

Preliminary Analysis of the BASALA-H Experimental Programme

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blaise, Patrick; Fougeras, Philippe; Philibert, Herve

2002-07-01

This paper is focused on the preliminary analysis of results obtained on the first cores of the first phase of the BASALA (Boiling water reactor Advanced core physics Study Aimed at mox fuel Lattice) programme, aimed at studying the neutronic parameters in ABWR core in hot conditions, currently under investigation in the French EOLE critical facility, within the framework of a cooperation between NUPEC, CEA and Cogema. The first 'on-line' analysis of the results has been made, using a new preliminary design and safety scheme based on the French APOLLO-2 code in its 2.4 qualified version and associated CEA-93 V4more » (JEF-2.2) Library, that will enable the Experimental Physics Division (SPEx) to perform future core designs. It describes the scheme adopted and the results obtained in various cases, going to the critical size determination to the reactivity worth of the perturbed configurations (voided, over-moderated, and poisoned with Gd{sub 2}O{sub 3}-UO{sub 2} pins). A preliminary study on the experimental results on the MISTRAL-4 is resumed, and the comparison of APOLLO-2 versus MCNP-4C calculations on these cores is made. The results obtained show very good agreements between the two codes, and versus the experiment. This work opens the way to the future full analysis of the experimental results of the qualifying teams with completely validated schemes, based on the new 2.5 version of the APOLLO-2 code. (authors)« less

Dansgaard-Oeschger cycles observed in the Greenland ReCAP ice core project

NASA Astrophysics Data System (ADS)

Kjær, Helle Astrid; Vallelonga, Paul; Vinther, Bo; Simonsen, Marius; Maffezzoli, Niccoló; Gkinis, Vasileios; Svensson, Anders; Jensen, Camilla Marie; Dallmayr, Remi; Spolaor, Andrea; Edwards, Ross

2017-04-01

The new REnland ice CAP (RECAP) ice core was drilled in summer 2015 in Greenland and measured by means of Continuous flow analysis (CFA) during the last 3 months of 2015. The Renland ice core was obtained as part of the ReCAP project, extending 584.11 meters to the bottom of the Renland ice cap located in east Greenland. The unique position on a mountain saddle above 2000 meters altitude, but close to the coast, ensures that the Renland ice core offers high accumulation, but also reaches far back in time. Results show that despite the short length the RECAP ice core holds ice all the way back to the past warm interglacial period, the Eemian. The glacial section is strongly thinned and covers on 20 meters of the ReCAP core, but nonetheless due to the high resolution of the measurements all 25 expected DO events could be identified. The record was analyzed for multiple elements including the water isotopes, forest fire tracers NH4+ and black carbon, insoluble dust particles by means of Abakus laser particle counter and the dust ion Ca2+, sea salt Na+, and sea ice proxies as well as acidity useful for finding volcanic layers to date the core. Below the glacial section another 20 meters of warm Eemian ice have been analysed. Here we present the chemistry results as obtained by continuous flow analysis (CFA) and compare the glacial section with the chemistry profile from other Greenland ice cores.

Coding, Constant Comparisons, and Core Categories: A Worked Example for Novice Constructivist Grounded Theorists.

PubMed

Giles, Tracey M; de Lacey, Sheryl; Muir-Cochrane, Eimear

2016-01-01

Grounded theory method has been described extensively in the literature. Yet, the varying processes portrayed can be confusing for novice grounded theorists. This article provides a worked example of the data analysis phase of a constructivist grounded theory study that examined family presence during resuscitation in acute health care settings. Core grounded theory methods are exemplified, including initial and focused coding, constant comparative analysis, memo writing, theoretical sampling, and theoretical saturation. The article traces the construction of the core category "Conditional Permission" from initial and focused codes, subcategories, and properties, through to its position in the final substantive grounded theory.

Study of low-velocity impact response of sandwich panels with shear-thickening gel cores

NASA Astrophysics Data System (ADS)

Wang, Yunpeng; Gong, Xinglong; Xuan, Shouhu

2018-06-01

The low-velocity impact response of sandwich panels with shear-thickening gel cores was studied. The impact tests indicated that the sandwich panels with shear-thickening gel cores showed excellent properties of energy dissipation and stress distribution. In comparison to the similar sandwich panels with chloroprene rubber cores and ethylene-propylene-diene monomer cores, the shear-thickening gel cores led to the obviously smaller contact forces and the larger energy absorptions. Numerical modelling with finite element analysis was used to investigate the stress distribution of the sandwich panels with shear-thickening gel cores and the results agreed well with the experimental results. Because of the unique mechanical property of the shear-thickening gel, the concentrated stress on the front facesheets were distributed to larger areas on the back facesheets and the peak stresses were reduced greatly.

Feasibility analysis of reciprocating magnetic heat pumps

NASA Technical Reports Server (NTRS)

Larson, A. V.; Hartley, J. G.; Shelton, S. V.; Smith, M. M.

1986-01-01

The conceptual design selected for detailed system analysis and optimization is the reciprocating gadolinium core in a regenerative fluid column within the bore of a superconducting magnet. The thermodynamic properties of gadolinium are given. A computerized literature search for relevant papers was conducted and is being analyzed. Contact was made with suppliers of superconducting magnets and accessories, magnetic materials, and various types of hardware. A description of the model for the thermal analysis of the core and regenerator fluids is included.

High Resolution Continuous Flow Analysis System for Polar Ice Cores

NASA Astrophysics Data System (ADS)

Dallmayr, Remi; Azuma, Kumiko; Yamada, Hironobu; Kjær, Helle Astrid; Vallelonga, Paul; Azuma, Nobuhiko; Takata, Morimasa

2014-05-01

In the last decades, Continuous Flow Analysis (CFA) technology for ice core analyses has been developed to reconstruct the past changes of the climate system 1), 2). Compared with traditional analyses of discrete samples, a CFA system offers much faster and higher depth resolution analyses. It also generates a decontaminated sample stream without time-consuming sample processing procedure by using the inner area of an ice-core sample.. The CFA system that we have been developing is currently able to continuously measure stable water isotopes 3) and electrolytic conductivity, as well as to collect discrete samples for the both inner and outer areas with variable depth resolutions. Chemistry analyses4) and methane-gas analysis 5) are planned to be added using the continuous water stream system 5). In order to optimize the resolution of the current system with minimal sample volumes necessary for different analyses, our CFA system typically melts an ice core at 1.6 cm/min. Instead of using a wire position encoder with typical 1mm positioning resolution 6), we decided to use a high-accuracy CCD Laser displacement sensor (LKG-G505, Keyence). At the 1.6 cm/min melt rate, the positioning resolution was increased to 0.27mm. Also, the mixing volume that occurs in our open split debubbler is regulated using its weight. The overflow pumping rate is smoothly PID controlled to maintain the weight as low as possible, while keeping a safety buffer of water to avoid air bubbles downstream. To evaluate the system's depth-resolution, we will present the preliminary data of electrolytic conductivity obtained by melting 12 bags of the North Greenland Eemian Ice Drilling (NEEM) ice core. The samples correspond to different climate intervals (Greenland Stadial 21, 22, Greenland Stadial 5, Greenland Interstadial 5, Greenland Interstadial 7, Greenland Stadial 8). We will present results for the Greenland Stadial -8, whose depths and ages are between 1723.7 and 1724.8 meters, and 35.520 to 35.636 kyr b2k 7), respectively. The results show the conductivity measured upstream and downstream of the debubbler. We will calculate the depth resolution of our system and compare it with earlier studies. 1) Bigler at al, "Optimization of High-Resolution Continuous Flow Analysis For Transient Climate Signals in Ice Cores". Environ. Sci. Technol. 2011, 45, 4483-4489 2) Kaufmann et al, "An Improved Continuous Flow Analysis System for High Resolution Field Measurements on Ice Cores". Environmental Environ. Sci. Technol. 2008, 42, 8044-8050 3) Gkinis, V., T. J. Popp, S. J. Johnsen and T, Blunier, 2010: A continuous stream flash evaporator for the calibration of an IR cavity ring down spectrometer for the isotopic analysis of water. Isotopes in Environmental and Health Studies, 46(4), 463-475. 4) McConnell et al, "Continuous ice-core chemical analyses using inductively coupled plasma mass spectrometry. Environ. Sci. Technol. 2002, 36, 7-11 5) Rhodes et al, "Continuous methane measurements from a late Holocene Greenland ice core : Atmospheric and in-situ signals" Earth and Planetary Science Letters. 2013, 368, 9-19 6) Breton et al, "Quantifying Signal Dispersion in a Hybrid Ice Core Melting System". Environ. Sci. Technol. 2012, 46, 11922-11928 7) Rasmussen et al, " A first chronology for the NEEM ice core". Climate of the Past. 2013, 9, 2967--3013

Constraining Genome-Scale Models to Represent the Bow Tie Structure of Metabolism for 13C Metabolic Flux Analysis

PubMed Central

Ando, David; Singh, Jahnavi; Keasling, Jay D.; García Martín, Héctor

2018-01-01

Determination of internal metabolic fluxes is crucial for fundamental and applied biology because they map how carbon and electrons flow through metabolism to enable cell function. 13C Metabolic Flux Analysis (13C MFA) and Two-Scale 13C Metabolic Flux Analysis (2S-13C MFA) are two techniques used to determine such fluxes. Both operate on the simplifying approximation that metabolic flux from peripheral metabolism into central “core” carbon metabolism is minimal, and can be omitted when modeling isotopic labeling in core metabolism. The validity of this “two-scale” or “bow tie” approximation is supported both by the ability to accurately model experimental isotopic labeling data, and by experimentally verified metabolic engineering predictions using these methods. However, the boundaries of core metabolism that satisfy this approximation can vary across species, and across cell culture conditions. Here, we present a set of algorithms that (1) systematically calculate flux bounds for any specified “core” of a genome-scale model so as to satisfy the bow tie approximation and (2) automatically identify an updated set of core reactions that can satisfy this approximation more efficiently. First, we leverage linear programming to simultaneously identify the lowest fluxes from peripheral metabolism into core metabolism compatible with the observed growth rate and extracellular metabolite exchange fluxes. Second, we use Simulated Annealing to identify an updated set of core reactions that allow for a minimum of fluxes into core metabolism to satisfy these experimental constraints. Together, these methods accelerate and automate the identification of a biologically reasonable set of core reactions for use with 13C MFA or 2S-13C MFA, as well as provide for a substantially lower set of flux bounds for fluxes into the core as compared with previous methods. We provide an open source Python implementation of these algorithms at https://github.com/JBEI/limitfluxtocore. PMID:29300340

Low back pain in 17 countries, a Rasch analysis of the ICF core set for low back pain.

PubMed

Røe, Cecilie; Bautz-Holter, Erik; Cieza, Alarcos

2013-03-01

Previous studies indicate that a worldwide measurement tool may be developed based on the International Classification of Functioning Disability and Health (ICF) Core Sets for chronic conditions. The aim of the present study was to explore the possibility of constructing a cross-cultural measurement of functioning for patients with low back pain (LBP) on the basis of the Comprehensive ICF Core Set for LBP and to evaluate the properties of the ICF Core Set. The Comprehensive ICF Core Set for LBP was scored by health professionals for 972 patients with LBP from 17 countries. Qualifier levels of the categories, invariance across age, sex and countries, construct validity and the ordering of the categories in the components of body function, body structure, activities and participation were explored by Rasch analysis. The item-trait χ2-statistics showed that the 53 categories in the ICF Core Set for LBP did not fit the Rasch model (P<0.001). The main challenge was the invariance in the responses according to country. Analysis of the four countries with the largest sample sizes indicated that the data from Germany fit the Rasch model, and the data from Norway, Serbia and Kuwait in terms of the components of body functions and activities and participation also fit the model. The component of body functions and activity and participation had a negative mean location, -2.19 (SD 1.19) and -2.98 (SD 1.07), respectively. The negative location indicates that the ICF Core Set reflects patients with a lower level of function than the present patient sample. The present results indicate that it may be possible to construct a clinical measure of function on the basis of the Comprehensive ICF Core Set for LBP by calculating country-specific scores before pooling the data.

Transforming Schools or Tinkering? An Analysis of CCSSO's Model Core Teaching Standards. A Center Policy & Practice Analysis Report

ERIC Educational Resources Information Center

Center for Mental Health in Schools at UCLA, 2010

2010-01-01

This report continues a series analyzing proposals and blueprints for transforming schools from the perspective of how well they delineate ways to enable equity of opportunity for "all" students to succeed at school. The focus of this report is on the July 2010 draft of CCSSO's (Council of Chief State School Office's) "Model Core Teaching…

Gender and Age Differences in the Core Triad of Impairments in Autism Spectrum Disorders: A Systematic Review and Meta-Analysis

ERIC Educational Resources Information Center

Wijngaarden-Cremers, Patricia J. M.; van Eeten, Evelien; Groen, Wouter B.; Van Deurzen, Patricia A.; Oosterling, Iris J.; Van der Gaag, Rutger Jan

2014-01-01

Autism is an extensively studied disorder in which the gender disparity in prevalence has received much attention. In contrast, only a few studies examine gender differences in symptomatology. This systematic review and meta-analysis of 22 peer reviewed original publications examines gender differences in the core triad of impairments in autism.…

Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS

DOE PAGES

Brown, C. S.; Zhang, Hongbin

2016-05-24

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surfacemore » temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.« less

Real-time oil-saturation monitoring in rock cores with low-field NMR.

PubMed

Mitchell, J; Howe, A M; Clarke, A

2015-07-01

Nuclear magnetic resonance (NMR) provides a powerful suite of tools for studying oil in reservoir core plugs at the laboratory scale. Low-field magnets are preferred for well-log calibration and to minimize magnetic-susceptibility-induced internal gradients in the porous medium. We demonstrate that careful data processing, combined with prior knowledge of the sample properties, enables real-time acquisition and interpretation of saturation state (relative amount of oil and water in the pores of a rock). Robust discrimination of oil and brine is achieved with diffusion weighting. We use this real-time analysis to monitor the forced displacement of oil from porous materials (sintered glass beads and sandstones) and to generate capillary desaturation curves. The real-time output enables in situ modification of the flood protocol and accurate control of the saturation state prior to the acquisition of standard NMR core analysis data, such as diffusion-relaxation correlations. Although applications to oil recovery and core analysis are demonstrated, the implementation highlights the general practicality of low-field NMR as an inline sensor for real-time industrial process control. Copyright © 2015 Elsevier Inc. All rights reserved.

Weather Research and Forecasting Model Sensitivity Comparisons for Warm Season Convective Initiation

NASA Technical Reports Server (NTRS)

Watson, Leela R.

2007-01-01

This report describes the work done by the Applied Meteorology Unit (AMU) in assessing the success of different model configurations in predicting warm season convection over East-Central Florida. The Weather Research and Forecasting Environmental Modeling System (WRF EMS) software allows users to choose among two dynamical cores - the Advanced Research WRF (ARW) and the Non-hydrostatic Mesoscale Model (NMM). There are also data assimilation analysis packages available for the initialization of the WRF model - the Local Analysis and Prediction System (LAPS) and the Advanced Regional Prediction System (ARPS) Data Analysis System (ADAS). Besides model core and initialization options, the WRF model can be run with one- or two-way nesting. Having a series of initialization options and WRF cores, as well as many options within each core, creates challenges for local forecasters, such as determining which configuration options are best to address specific forecast concerns. This project assessed three different model intializations available to determine which configuration best predicts warm season convective initiation in East-Central Florida. The project also examined the use of one- and two-way nesting in predicting warm season convection.

Estimating a Preference-Based Index from the Clinical Outcomes in Routine Evaluation–Outcome Measure (CORE-OM)

PubMed Central

Brazier, John E.; Rowen, Donna; Barkham, Michael

2013-01-01

Background. The Clinical Outcomes in Routine Evaluation–Outcome Measure (CORE-OM) is used to evaluate the effectiveness of psychological therapies in people with common mental disorders. The objective of this study was to estimate a preference-based index for this population using CORE-6D, a health state classification system derived from the CORE-OM consisting of a 5-item emotional component and a physical item, and to demonstrate a novel method for generating states that are not orthogonal. Methods. Rasch analysis was used to identify 11 emotional health states from CORE-6D that were frequently observed in the study population and are, thus, plausible (in contrast, conventional statistical design might generate implausible states). Combined with the 3 response levels of the physical item of CORE-6D, they generate 33 plausible health states, 18 of which were selected for valuation. A valuation survey of 220 members of the public in South Yorkshire, United Kingdom, was undertaken using the time tradeoff (TTO) method. Regression analysis was subsequently used to predict values for all possible states described by CORE-6D. Results. A number of multivariate regression models were built to predict values for the 33 health states of CORE-6D, using the Rasch logit value of the emotional state and the response level of the physical item as independent variables. A cubic model with high predictive value (adjusted R2 = 0.990) was selected to predict TTO values for all 729 CORE-6D health states. Conclusion. The CORE-6D preference-based index will enable the assessment of cost-effectiveness of interventions for people with common mental disorders using existing and prospective CORE-OM data sets. The new method for generating states may be useful for other instruments with highly correlated dimensions. PMID:23178639

Time Spent Teaching Core Academic Subjects in Elementary Schools. Comparisons across Community, School, Teacher, and Student Characteristics. Statistical Analysis Report.

ERIC Educational Resources Information Center

Perie, Marianne; And Others

The proportion of time that elementary school teachers use to teach core academic subjects (English/reading/language arts, mathematics, social studies, science) is an important aspect of instruction. Spending a large proportion of time teaching core curriculum subjects may be important not only in terms of school quality, but also in terms of…

Air-stable n-type semiconductor: core-perfluoroalkylated perylene bisimides.

PubMed

Li, Yan; Tan, Lin; Wang, Zhaohui; Qian, Hualei; Shi, Yubai; Hu, Wenping

2008-02-21

A series of core-perfluoroalkylated perylene bisimides (PBIs) have been efficiently synthesized by copper-mediated perfluoroalkylation of dibrominated PBIs. Their aromatic cores are highly twisted due to the steric encumbrance in the bay regions as revealed by single-crystal X-ray analysis. The organic field-effect transistors (OFETs) incorporating these new n-type semiconductors show remarkable air-stability and good field effect mobility.

Common Core English and Language Arts K-1 Exemplar Text Set: A Critical Content Analysis of Cultural Representations

ERIC Educational Resources Information Center

McCaffrey, Megan

2014-01-01

With the implementation of Common Core State Standards (CCSS) in over forty states, teachers are putting into practice the CCSS text exemplars of text complexity. Of particular concern for the purpose of this research are the kindergarten and first grade (K-1) read aloud and independent text exemplar lists. While not intended as core reading…

An analysis of the New York State Earth Science Curriculum with respect to standards, classroom practices, and the Regents Examination

NASA Astrophysics Data System (ADS)

Contino, Julie Anna

In a standards-based system, it is important for all components of the system to align in order to achieve the intended goals. In New York State, standards are provided to the teachers who then create individual curricula that will lead to student success on the state assessment. This mixed methods study presents an analysis of the alignment between the National Science Education Standards (NSES), New York State Physical Setting/Earth Science Core Curriculum (Core Curriculum), and New York State Earth Science Regents Examination (Regents)---the sources teachers use for creating Earth Science curricula in New York State. The NSES were found to have a 49% overlap with the Core Curriculum and a 27% overlap with the Regents. The Core Curriculum and Regents, represented by matrices consisting of performance indicators and cognitive demands, were compared using the Porter alignment index. The alignment was 0.35, categorized as slightly aligned, due to the different emphases on cognitive levels (the Core Curriculum focused on Understand and Apply while the Regents focused on Apply followed by Understand and Remember). Additionally, a purposeful sample of experienced and innovative teachers were surveyed and interviewed to gain insight on how NYS Earth Science teachers organize their scope and sequences, align their lessons with the Core Curriculum, establish internal lesson coherence, and prepare their students for the Regents Exam. Teachers' scope and sequences were well-aligned with the Core Curriculum and Regents but misalignment was found between their lessons and the Core Curriculum as well as between the stated objectives for their students and evaluation of those objectives. Based on the findings, it is suggested that the NSES be revised and the Core Curriculum updated to include quantifiable emphasis on the major understandings such as percentage of time, as well as an emphasis on alignment principles. Teacher professional development focused on alignment issues relative to the state standards and enhancing internal lesson coherence should also be provided. The insights gained from this analysis of the NYS system may be helpful to other states as they move toward standards-based systems.

Shape evolution of a core-shell spherical particle under hydrostatic pressure.

PubMed

Colin, Jérôme

2012-03-01

The morphological evolution by surface diffusion of a core-shell spherical particle has been investigated theoretically under hydrostatic pressure when the shear modulii of the core and shell are different. A linear stability analysis has demonstrated that depending on the pressure, shear modulii, and radii of both phases, the free surface of the composite particle may be unstable with respect to a shape perturbation. A stability diagram finally emphasizes that the roughness development is favored in the case of a hard shell with a soft core.

Planktic foraminifer census data from Northwind Ridge Core 5, Arctic Ocean

USGS Publications Warehouse

Foley, Kevin M.; Poore, Richard Z.

1991-01-01

The U.S. Geological Survey recovered 9 piston cores from the Northwind Ridge in the Canada Basin of the Arctic Ocean from a cruise of the USCGC Polar Star during 1988. Preliminary analysis of the cores suggests sediments deposited on Northwind Ridge preserve a detailed record of glacial and interglacial cycles for the last few hundred-thousand to one million years. This report includes quantitative data on foraminifers and selected sediment size-fraction data in samples from Northwind Ridge core PI-88AR P5.

Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop

NASA Technical Reports Server (NTRS)

Clark, John S. (Editor)

1991-01-01

Papers presented at the joint NASA/DOE/DOD workshop on nuclear thermal propulsion are compiled. The following subject areas are covered: nuclear thermal propulsion programs; Rover/NERVA and NERVA systems; Low Pressure Nuclear Thermal Rocket (LPNTR); particle bed reactor nuclear rocket; hybrid propulsion systems; wire core reactor; pellet bed reactor; foil reactor; Droplet Core Nuclear Rocket (DCNR); open cycle gas core nuclear rockets; vapor core propulsion reactors; nuclear light bulb; Nuclear rocket using Indigenous Martian Fuel (NIMF); mission analysis; propulsion and reactor technology; development plans; and safety issues.

Atomic configurations at InAs partial dislocation cores associated with Z-shape faulted dipoles.

PubMed

Li, Luying; Gan, Zhaofeng; McCartney, Martha R; Liang, Hanshuang; Yu, Hongbin; Gao, Yihua; Wang, Jianbo; Smith, David J

2013-11-15

The atomic arrangements of two types of InAs dislocation cores associated by a Z-shape faulted dipole are observed directly by aberration-corrected high-angle annular-dark-field imaging. Single unpaired columns of different atoms in a matrix of dumbbells are clearly resolved, with observable variations of bonding lengths due to excess Coulomb force from bare ions at the dislocation core. The corresponding geometric phase analysis provides confirmation that the dislocation cores serve as origins of strain field inversion while stacking faults maintain the existing strain status.

Role of CT scanning in formation evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bergosh, J.L.; Dibona, B.G.

1988-01-01

The use of the computerized tomographic (CT) scanner in formation evaluation of difficult to analyze core samples has moved from the research and development phase to daily, routine use in the core-analysis laboratory. The role of the CT scanner has become increasingly important as geologists try to obtain more representative core material for accurate formation evaluation. The most common problem facing the core analyst when preparing to measure petrophysical properties is the selection of representative and unaltered core samples for routine and special core testing. Recent data have shown that heterogeneous reservoir rock can be very difficult, if not impossible,more » to assess correctly when using standard core examination procedures, because many features, such as fractures, are not visible on the core surface. Another problem is the invasion of drilling mud into the core sample. Flushing formation oil and water from the core can greatly alter the saturation and distribution of fluids and lead to serious formation evaluation problems. Because the quality and usefulness of the core date are directly tied to proper sample selection, it has become imperative that the CT scanner be used whenever possible.« less

Evaluation of solar flares and electron precipitation by nitrate distribution in Antarctica

NASA Astrophysics Data System (ADS)

Dreschhoff, Gisela A.; Zeller, Edward J.

1991-10-01

Most of the time devoted to project research was spent in Antarctica. A firm core was drilled by hand to a depth of 29 meters at Windless Bight on the Ross Ice Shelf. The main result is that all of the major peaks identified as resulting from ionization caused by SPEs that were found in the 1988-89 core could also be identified in the analytical sequence from the 1990-91 core. Following the Antarctic field season, a set of snow samples were obtained that had been collected by the International Trans-Antarctica Expedition. The analysis of these samples showed nitrate flux that correlates closely with known spatial distribution of electron precipitation in the south polar region. A new apparatus has been build for field analysis on a continuous basis of nitrate and conductivity in a melt derived from the vertical melting of ice cores.

Pb(core)/ZnO(shell) nanowires obtained by microwave-assisted method

PubMed Central

2011-01-01

In this study, Pb-filled ZnO nanowires [Pb(core)/ZnO(shell)] were synthesized by a simple and novel one-step vapor transport and condensation method by microwave-assisted decomposition of zinc ferrite. The synthesis was performed using a conventional oven at 1000 W and 5 min of treatment. After synthesis, a spongy white cotton-like material was obtained in the condensation zone of the reaction system. HRTEM analysis revealed that product consists of a Pb-(core) with (fcc) cubic structure that preferentially grows in the [111] direction and a hexagonal wurtzite ZnO-(Shell) that grows in the [001] direction. Nanowire length was more than 5 μm and a statistical analysis determined that the shell and core diameters were 21.00 ± 3.00 and 4.00 ± 1.00 nm, respectively. Experimental, structural details, and synthesis mechanism are discussed in this study. PMID:21985637