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Sample records for analysis identifies core

  1. Numerical analysis and experiment to identify origin of buckling in rapid cycling synchrotron core

    International Nuclear Information System (INIS)

    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. -- Highlights: • Study to identify the origin of buckling in the MA cores is presented. • Thermal stress simulations and compressive strength tests were carried out. • Results show that thermal stress is the origin of core buckling. • Thermal stress can be reduced by using cores without epoxy impregnation

  2. Identifying a core set of medical informatics serials: an analysis using the MEDLINE database.

    OpenAIRE

    Sittig, D. F.

    1996-01-01

    A study was undertaken to test the hypothesis that a core set of medical informatics serials could be identified by using standard bibliometric techniques. All journal articles indexed by the National Library of Medicine between 1990 and 1994 were included. Articles were identified by using the "MEDICAL INFORMATICS" Medical Subject Heading (MeSH) term. Each serial title containing articles was then ranked according to (1) the total number of medical informatics journal articles indexed and (2...

  3. Analysis of pan-genome to identify the core genes and essential genes of Brucella spp.

    Science.gov (United States)

    Yang, Xiaowen; Li, Yajie; Zang, Juan; Li, Yexia; Bie, Pengfei; Lu, Yanli; Wu, Qingmin

    2016-04-01

    Brucella spp. are facultative intracellular pathogens, that cause a contagious zoonotic disease, that can result in such outcomes as abortion or sterility in susceptible animal hosts and grave, debilitating illness in humans. For deciphering the survival mechanism of Brucella spp. in vivo, 42 Brucella complete genomes from NCBI were analyzed for the pan-genome and core genome by identification of their composition and function of Brucella genomes. The results showed that the total 132,143 protein-coding genes in these genomes were divided into 5369 clusters. Among these, 1710 clusters were associated with the core genome, 1182 clusters with strain-specific genes and 2477 clusters with dispensable genomes. COG analysis indicated that 44 % of the core genes were devoted to metabolism, which were mainly responsible for energy production and conversion (COG category C), and amino acid transport and metabolism (COG category E). Meanwhile, approximately 35 % of the core genes were in positive selection. In addition, 1252 potential essential genes were predicted in the core genome by comparison with a prokaryote database of essential genes. The results suggested that the core genes in Brucella genomes are relatively conservation, and the energy and amino acid metabolism play a more important role in the process of growth and reproduction in Brucella spp. This study might help us to better understand the mechanisms of Brucella persistent infection and provide some clues for further exploring the gene modules of the intracellular survival in Brucella spp. PMID:26724943

  4. Comparative Genomic Analysis of Meningitis- and Bacteremia-Causing Pneumococci Identifies a Common Core Genome.

    Science.gov (United States)

    Kulohoma, Benard W; Cornick, Jennifer E; Chaguza, Chrispin; Yalcin, Feyruz; Harris, Simon R; Gray, Katherine J; Kiran, Anmol M; Molyneux, Elizabeth; French, Neil; Parkhill, Julian; Faragher, Brian E; Everett, Dean B; Bentley, Stephen D; Heyderman, Robert S

    2015-10-01

    Streptococcus pneumoniae is a nasopharyngeal commensal that occasionally invades normally sterile sites to cause bloodstream infection and meningitis. Although the pneumococcal population structure and evolutionary genetics are well defined, it is not clear whether pneumococci that cause meningitis are genetically distinct from those that do not. Here, we used whole-genome sequencing of 140 isolates of S. pneumoniae recovered from bloodstream infection (n = 70) and meningitis (n = 70) to compare their genetic contents. By fitting a double-exponential decaying-function model, we show that these isolates share a core of 1,427 genes (95% confidence interval [CI], 1,425 to 1,435 genes) and that there is no difference in the core genome or accessory gene content from these disease manifestations. Gene presence/absence alone therefore does not explain the virulence behavior of pneumococci that reach the meninges. Our analysis, however, supports the requirement of a range of previously described virulence factors and vaccine candidates for both meningitis- and bacteremia-causing pneumococci. This high-resolution view suggests that, despite considerable competency for genetic exchange, all pneumococci are under considerable pressure to retain key components advantageous for colonization and transmission and that these components are essential for access to and survival in sterile sites. PMID:26259813

  5. Promise and Pitfalls of Using Grain Size Analysis to Identify Glacial Sediments in Alpine Lake Cores.

    Science.gov (United States)

    Clark, D. H.

    2011-12-01

    Lakes fed by glacier outwash should have a clastic particle-size record distinct from non-glacial lakes in the same area, but do they? The unique turquoise color of alpine glacial lakes reflects the flux of suspended clastic glacial rock flour to those lakes; conversely, lakes not fed by outwash are generally clear with sediments dominated by organics or slope-wash from nearby hillslopes. This contrast in sediment types and sources should produce a distinct and measureable different in grain sizes between the two settings. Results from a variety of lakes suggest the actual situation is often more subtle and complex. I compare grain size results to other proxies to assess the value of grain size analysis for paleoglacier studies. Over the past 10 years, my colleagues and I have collected and analyzed sediment cores from a wide variety of lakes below small alpine glaciers in an attempt to constrain the timing and magnitude of alpine glaciation in those basins. The basic concept is that these lakes act as continuous catchments for any rock flour produced upstream by glacier abrasion; as a glacier grows, the flux of rock flour to the lake will also increase. If the glacier disappears entirely, rock flour deposition will also cease in short order. We have focused our research in basins with simple sedimentologic settings: mostly small, high-altitude, stripped granitic or metamorphic cirques in which the cirque glaciers are the primary source of clastic sediments. In most cases, the lakes are fed by meltwater from a modern glacier, but were ice free during the earlier Holocene. In such cases, the lake cores should record formation of and changes in activity of the glacier upstream. We used a Malvern Mastersizer 2000 laser particle size analyzer for our grain size analyses, as well as recording magnetic susceptibility, color, and organics for the same cores. The results indicate that although lakes often experience increases in silt and clay-size (<0.63 mm) clastic

  6. Identifying Domestic and Imported Core Inflation

    OpenAIRE

    Bjørnland, Hilde C.

    2000-01-01

    This paper estimates core inflation in Norway, identified as that component of inflation that has no long-run effect on GDP. The model distinguishes explicitly between domestic and imported core inflation. The results show that (domestic) core inflation is the main component of CPI inflation. CPI inflation, however, misrepresents core inflation in some periods. The differences are well explained by the other shocks identified in the model, in particular the oil price shocks of the 1970s when ...

  7. Pan-Genome Analysis of Human Gastric Pathogen H. pylori: Comparative Genomics and Pathogenomics Approaches to Identify Regions Associated with Pathogenicity and Prediction of Potential Core Therapeutic Targets

    DEFF Research Database (Denmark)

    Ali, Amjad; Naz, Anam; Soares, Siomar C.;

    2015-01-01

    -genome approach; the predicted conserved gene families (1,193) constitute similar to 77% of the average H. pylori genome and 45% of the global gene repertoire of the species. Reverse vaccinology strategies have been adopted to identify and narrow down the potential core-immunogenic candidates. Total of 28 nonhost...... homolog proteins were characterized as universal therapeutic targets against H. pylori based on their functional annotation and protein-protein interaction. Finally, pathogenomics and genome plasticity analysis revealed 3 highly conserved and 2 highly variable putative pathogenicity islands in all...

  8. PWR degraded core analysis

    International Nuclear Information System (INIS)

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  9. Operating Experience Identified Core Performance Issues

    International Nuclear Information System (INIS)

    The economic incentive for optimizing core design and the highly competitive markets for supply of fuel and control rods have resulted in extended-length fuel cycles and extended-burnup fuel applications that place increasing demands on the design performance of fuel and control rods. Design changes to gain these advantages have contributed to operating performance events that may have been avoidable. Steady improvement in the operating performance of fuel and core components has perhaps lulled the industry into complacency that has permitted design changes to be implemented with incomplete fuel testing or developmental testing that would have delayed the marketing of new fuel and control rod designs. This paper cites examples of apparent design deficiencies revealed by operating experience (undetected manufacturing defects, crud-induced local corrosion - CILC, debris fretting failures of PWR fuel, vibration-induced grid-to-rod fretting failures at low-pressure drop mid-grid locations of PWR fuel), and discusses other design proposals that could lead to similar problems. The U.S. NRC Core Performance Action Plan to improve regulatory oversight and contribute to continued excellence in core performance behavior is described

  10. TMI-2 core debris analysis

    International Nuclear Information System (INIS)

    One of the ongoing examination tasks for the damaged TMI-2 reactor is analysis of samples of debris obtained from the debris bed presently at the top of the core. This paper summarizes the results reported in the TMI-2 Core Debris Grab Sample Examination and Analysis Report, which will be available early in 1986. The sampling and analysis procedures are presented, and information is provided on the key results as they relate to the present core condition, peak temperatures during the transient, temperature history, chemical interactions, and core relocation. The results are then summarized

  11. Power excursion analysis for high burnup cores

    International Nuclear Information System (INIS)

    A study was undertaken of power excursions in high burnup cores. There were three objectives in this study. One was to identify boiling water reactor (BWR) and pressurized water reactor (PWR) transients in which there is significant energy deposition in the fuel. Another was to analyze the response of BWRs to the rod drop accident (RDA) and other transients in which there is a power excursion. The last objective was to investigate the sources of uncertainty in the RDA analysis. In a boiling water reactor, the events identified as having significant energy deposition in the fuel were a rod drop accident, a recirculation flow control failure, and the overpressure events; in a pressurized water reactor, they were a rod ejection accident and boron dilution events. The RDA analysis was done with RAMONA-4B, a computer code that models the space- dependent neutron kinetics throughout the core along with the thermal hydraulics in the core, vessel, and steamline. The results showed that the calculated maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important uncertainties in each of these categories are discussed in the report

  12. Core dimensions of recovery: a psychometric analysis.

    Science.gov (United States)

    Gordon, Sarah E; Ellis, Pete M; Siegert, Richard J; Walkey, Frank H

    2014-07-01

    Core recovery dimensions lie between the large general factor of recovery and its underlying components. Identifying these could enhance recovery frameworks, practice and research. In contrast to existing conceptually based taxonomies, we sought to empirically identify the core dimensions of recovery through further psychometric analysis of a robust eleven factor (sub-scale) consumer recovery outcome measure, My Voice, My Life. We subjected the sub-scale scores of 504 consumers to further principal components analyses, beginning with a single unrotated factor and progressing through two to nine factors with varimax rotation. We found the five-factor solution to provide an orderly intermediate configuration with the eleven recovery factors having either aligned and/or disengaged through the process to result in the following core dimensions: (1) Belonging and relating (encompassing the individual factors of spirituality, culture, and relationships); (2) Being and doing (encompassing the individual factors of physical health, day-to-day life, and quality of life); (3) Thinking and feeling (encompassing the individual factors of recovery, mental health, and hope and empowerment); (4) Resources (which maintained its independence); and (5) Satisfaction with Services (which also maintained its independence). We compare this empirical configuration with conceptually based taxonomies. PMID:23588506

  13. Identifying core domains to assess flare in rheumatoid arthritis

    DEFF Research Database (Denmark)

    Bartlett, Susan J; Hewlett, Sarah; Bingham, Clifton O;

    2012-01-01

    For rheumatoid arthritis (RA), there is no consensus on how to define and assess flare. Variability in flare definitions impairs understanding of findings across studies and limits ability to pool results. The OMERACT RA Flare Group sought to identify domains to define RA flares from patient and...

  14. PWR Core 2 Project accident analysis

    International Nuclear Information System (INIS)

    The various operations required for receipt, handling, defueling and storage of spent Shippingport PWR Core 2 fuel assemblies have been evaluated to determine the potential accidents and their consequences. These operations will introduce approximately 16,500 kilograms of depleted natural uranium (as UO2), 139 kilograms of plutonium, 2.8 megacuries of mixed fission products, and 14 kilograms of Zircaloy-4 (cladding and hardware) into the 221-T Canyon Building. Event sequences for potential accidents that were considered included (1) leaking fuel assemblies, (2) fire and explosion, (3) loss of coolant or cooling capability, (4) dropped and/or damaged fuel assemblies, and extrinsic occurrences such as loss of services, missile impact, and natural occurrences (e.g., earthquake, tornado). Accident frequencies were determined by formal analysis to be very low. Accident consequences are greatly mitigated by the safety and containment features designed into the fuel modules and shipping cask, the long cooling time since reactor discharge, and the redundant safety features designed into the facilities, equipment, and operating procedures for the PWR Core 2 Project. Possible hazards associated with the handling of these fuels have been considered and adequate safeguards and storage constraints identified. The operations of M-160 cask unloading and module storage will not involve identifiable risks as great or significantly greater than those for comparable licensed nuclear facilities, nor will hazards or risks be significantly different from comparable past 221-T Plant programs. Therefore, it is concluded that the operations required for receipt, handling, and defueling of the M-160 cask and for the storage and surveillance of the PWR Core 2 fuel assemblies at the 221-T Canyon Building can be performed without undue risk to the safety of the involved personnel, the public, the environment or the facility

  15. Structural safety analysis of HTGR core supports

    Energy Technology Data Exchange (ETDEWEB)

    Ju, F.; Bennett, J.G.; Anderson, C.A.

    1977-01-01

    In the current design of the High Temperature Gas-Cooled Reactor (HTGR), the core is made up of stacked columns of graphite fuel blocks. Structural support for the core takes the form of graphite columns beneath the core together with lateral springs, which position and restrain the core from contact with the sides of the reactor containment vessel. Each individual support column carries the dead load of several fuel columns together with the equivalent load caused by the coolant pressure drop through the core. The adequacy of the support structure to provide torsional stability of the core for both static and seismic loadings as well as long term stability of the core support structure itself is discussed. Analysis for long term stability of the core support columns involves consideration of eccentric loading (caused by damaged spherical seats) and imperfections in the form of surface cracks. Nonlinear graphite behavior must also be taken into consideration. For predictions of the core torsional seismic response, the core was represented as a right circular cylinder supported on elastic posts; the lateral support was represented by a single torsional spring. Energy losses from friction and material hysteresis were represented by viscous dampers. The coupled equations for vertical and rotational motions were integrated numerically and dynamic core response was computed fromtorsional acceleration time-histories obtained by differentiating a horizontal accelerogram and dividing by the shear wave speed for hard and soft soil conditions.

  16. Identifying Core Mobile Learning Faculty Competencies Based Integrated Approach: A Delphi Study

    Science.gov (United States)

    Elbarbary, Rafik Said

    2015-01-01

    This study is based on the integrated approach as a concept framework to identify, categorize, and rank a key component of mobile learning core competencies for Egyptian faculty members in higher education. The field investigation framework used four rounds Delphi technique to determine the importance rate of each component of core competencies…

  17. Core Intervention Components: Identifying and Operationalizing What Makes Programs Work. ASPE Research Brief

    Science.gov (United States)

    Blase, Karen; Fixsen, Dean

    2013-01-01

    This brief is part of a series that explores key implementation considerations. It focuses on the importance of identifying, operationalizing, and implementing the "core components" of evidence-based and evidence-informed interventions that likely are critical to producing positive outcomes. The brief offers a definition of "core components",…

  18. Evaluation of Analysis Techniques for Fluted-Core Sandwich Cylinders

    Science.gov (United States)

    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.

  19. Overview on Hydrate Coring, Handling and Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jon Burger; Deepak Gupta; Patrick Jacobs; John Shillinglaw

    2003-06-30

    Gas hydrates are crystalline, ice-like compounds of gas and water molecules that are formed under certain thermodynamic conditions. Hydrate deposits occur naturally within ocean sediments just below the sea floor at temperatures and pressures existing below about 500 meters water depth. Gas hydrate is also stable in conjunction with the permafrost in the Arctic. Most marine gas hydrate is formed of microbially generated gas. It binds huge amounts of methane into the sediments. Worldwide, gas hydrate is estimated to hold about 1016 kg of organic carbon in the form of methane (Kvenvolden et al., 1993). Gas hydrate is one of the fossil fuel resources that is yet untapped, but may play a major role in meeting the energy challenge of this century. In June 2002, Westport Technology Center was requested by the Department of Energy (DOE) to prepare a ''Best Practices Manual on Gas Hydrate Coring, Handling and Analysis'' under Award No. DE-FC26-02NT41327. The scope of the task was specifically targeted for coring sediments with hydrates in Alaska, the Gulf of Mexico (GOM) and from the present Ocean Drilling Program (ODP) drillship. The specific subjects under this scope were defined in 3 stages as follows: Stage 1: Collect information on coring sediments with hydrates, core handling, core preservation, sample transportation, analysis of the core, and long term preservation. Stage 2: Provide copies of the first draft to a list of experts and stakeholders designated by DOE. Stage 3: Produce a second draft of the manual with benefit of input from external review for delivery. The manual provides an overview of existing information available in the published literature and reports on coring, analysis, preservation and transport of gas hydrates for laboratory analysis as of June 2003. The manual was delivered as draft version 3 to the DOE Project Manager for distribution in July 2003. This Final Report is provided for records purposes.

  20. BWR core stability analysis with RETRAN

    International Nuclear Information System (INIS)

    A RETRAN model was developed for determining the stability of boiling water reactor. This model was benchmarked against plant data from stability tests conducted during plant operations. The stability analysis with RETRAN is demonstrated using best estimate RETRAN input data representative of the NSSS. All of the important neutronic and thermal hydraulic feedback mechanisms are taken into account through the modeling of the reactor vessel, recirculation loops and core neutronics. The analysis was performed with the RETRAN02 MOD003 computer code. The transient is initialized by a small step decrease in the steam dome pressure. The core exit (upper plenum) pressure and core power transient responses to this perturbation are transformed into frequency data and a system transfer function is then obtained. The system transfer function is fitted to a second order equation from which the decay ratio and natural frequency can be determined

  1. Application of Raman spectroscopy to identify microcalcifications and underlying breast lesions at stereotactic core needle biopsy

    OpenAIRE

    Barman, Ishan; Dingari, Narahara Chari; Saha, Anushree; McGee, Sasha; Galindo, Luis H.; Liu, Wendy; Plecha, Donna; Klein, Nina; Dasari, Ramachandra Rao; Fitzmaurice, Maryann

    2013-01-01

    Microcalcifications are a feature of diagnostic significance on a mammogram and a target for stereotactic breast needle biopsy. Here, we report development of a Raman spectroscopy technique to simultaneously identify microcalcification status and diagnose the underlying breast lesion, in real-time, during stereotactic core needle biopsy procedures. Raman spectra were obtained ex vivo from 146 tissue sites from fresh stereotactic breast needle biopsy tissue cores from 33 patients, including 50...

  2. CFD Analysis of Core Bypass Phenomena

    International Nuclear Information System (INIS)

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the one-twelfth grid can be set as a symmetry boundary

  3. CFD Analysis of Core Bypass Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Richard W. Johnson; Hiroyuki Sato; Richard R. Schultz

    2010-03-01

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the sector grid can be set as a symmetry boundary

  4. CFD Analysis of Core Bypass Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Richard W. Johnson; Hiroyuki Sato; Richard R. Schultz

    2009-11-01

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the one-twelfth grid can be set as a symmetry boundary

  5. Identifiability analysis in conceptual sewer modelling.

    Science.gov (United States)

    Kleidorfer, M; Leonhardt, G; Rauch, W

    2012-01-01

    For a sufficient calibration of an environmental model not only parameter sensitivity but also parameter identifiability is an important issue. In identifiability analysis it is possible to analyse whether changes in one parameter can be compensated by appropriate changes of the other ones within a given uncertainty range. Parameter identifiability is conditional to the information content of the calibration data and consequently conditional to a certain measurement layout (i.e. types of measurements, number and location of measurement sites, temporal resolution of measurements etc.). Hence the influence of number and location of measurement sites on the number of identifiable parameters can be investigated. In the present study identifiability analysis is applied to a conceptual model of a combined sewer system aiming to predict the combined sewer overflow emissions. Different measurement layouts are tested and it can be shown that only 13 of the most sensitive catchment areas (represented by the model parameter 'effective impervious area') can be identified when overflow measurements of the 20 highest overflows and the runoff to the waste water treatment plant are used for calibration. The main advantage of this method is very low computational costs as the number of required model runs equals the total number of model parameters. Hence, this method is a valuable tool when analysing large models with a long runtime and many parameters. PMID:22864432

  6. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals

  7. High-resolution sulfur isotopes in ice cores identify large stratospheric volcanic eruptions

    Science.gov (United States)

    Burke, Andrea; Sigl, Michael; Adkins, Jess; Paris, Guillaume; McConnell, Joe

    2016-04-01

    The record of the volcanic forcing of climate over the past 2500 years is reconstructed primarily from sulfate concentrations in ice cores. Of particular interest are stratospheric eruptions, as these afford sulfate aerosols the longest residence time and largest dispersion in the atmosphere, and thus the greatest impact on radiative forcing. Identification of stratospheric eruptions currently relies on the successful matching of the same volcanic sulphate peak in ice cores from both the Northern and Southern hemispheres (a "bipolar event"). These are interpreted to reflect the global distribution of sulfur aerosols by the stratospheric winds. Despite its recent success, this method relies on precise and accurate dating of ice cores, in order to distinguish between a true 'bipolar event' and two separate eruptions that occurred in close temporal succession. Sulfur isotopes can been used to distinguish between these two scenarios since stratospheric sulfur aerosols are exposed to UV radiation which imparts a mass independent fractionation (Baroni et al., 2007). Mass independent fractionation of sulfate in ice cores thus offers a novel method of fingerprinting stratospheric eruptions, and thus refining the historic record of explosive volcanism and its forcing of climate. Here we present new high-resolution (sub-annual) sulfur isotope data from the Tunu Ice core in Greenland over seven eruptions. Sulfur isotopes were measured by MC-ICP-MS, which substantially reduces sample size requirements and allows high temporal resolution from a single ice core. We demonstrate the efficacy of the method on recent, well-known eruptions (including Pinatubo and Katmai/Novarupta), and then apply it to unidentified sulfate peaks, allowing us to identify new stratospheric eruptions. Baroni, M., Thiemens, M. H., Delmas, R. J., & Savarino, J. (2007). Mass-independent sulfur isotopic compositions in stratospheric volcanic eruptions. Science, 315(5808), 84-87. http://doi.org/10

  8. Geologic analysis of Devonian Shale cores

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-02-01

    Cleveland Cliffs Iron Company was awarded a DOE contract in December 1977 for field retrieval and laboratory analysis of cores from the Devonian shales of the following eleven states: Michigan, Illinois, Indiana, Ohio, New York, Pennsylvania, West Virginia, Maryland, Kentucky, Tennessee and Virginia. The purpose of this project is to explore these areas to determine the amount of natural gas being produced from the Devonian shales. The physical properties testing of the rock specimens were performed under subcontract at Michigan Technological University (MTU). The study also included LANDSAT information, geochemical research, structural sedimentary and tectonic data. Following the introduction, and background of the project this report covers the following: field retrieval procedures; laboratory procedures; geologic analysis (by state); references and appendices. (ATT)

  9. Core Backbone Convergence Mechanisms and Microloops Analysis

    Directory of Open Access Journals (Sweden)

    Abdelali Ala

    2012-07-01

    Full Text Available In this article we study approaches that can be used to minimise the convergence time, we also make a focus on microloops phenomenon, analysis and means to mitigate them. The convergence time reflects the time required by a network to react to a failure of a link or a router failure itself. When all nodes (routers have updated their respective routing and forwarding databases, we can say the network has converged. This study will help in building real-time and resilient network infrastructure, the goal is to make any evenement in the core network, as transparent as possible to any sensitive and real-time flows. This study is also, a deepening of earlier works presented in [10] and [11].

  10. Tank 241-B-203 push mode core sampling and analysis plan. Revision 1

    International Nuclear Information System (INIS)

    This Sampling and Analysis Plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for two push-mode core samples from tank 241-B-203 (B-203)

  11. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO2 and MOX fuel rods, (3) analysis of isotopic composition data for UO2 and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  12. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    JNES has been developing a technical data base used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical data base, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The data base will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects (1) analysis of the measurement data of Doppler reactivity in experimental MOX core simulating LWR cores, (2) measurements of isotopic compositions of fission product nuclides on high-burnup BWR UO2 fuels and the analysis of the measurement data, and (3) neutronics analysis of the experimental data that has been obtained in the international joint programs such as FUBILA and REBUS. (author)

  13. Multi-Core Processor Memory Contention Benchmark Analysis Case Study

    Science.gov (United States)

    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.

  14. BN-600 full MOX core benchmark analysis

    International Nuclear Information System (INIS)

    As a follow-up of the BN-600 hybrid core benchmark, a full MOX core benchmark was performed within the framework of the IAEA co-ordinated research project. Discrepancies between the values of main reactivity coefficients obtained by the participants for the BN-600 full MOX core benchmark appear to be larger than those in the previous hybrid core benchmarks on traditional core configurations. This arises due to uncertainties in the proper modelling of the axial sodium plenum above the core. It was recognized that the sodium density coefficient strongly depends on the core model configuration of interest (hybrid core vs. fully MOX fuelled core with sodium plenum above the core) in conjunction with the calculation method (diffusion vs. transport theory). The effects of the discrepancies revealed between the participants results on the ULOF and UTOP transient behaviours of the BN-600 full MOX core were investigated in simplified transient analyses. Generally the diffusion approximation predicts more benign consequences for the ULOF accident but more hazardous ones for the UTOP accident when compared with the transport theory results. The heterogeneity effect does not have any significant effect on the simulation of the transient. The comparison of the transient analyses results concluded that the fuel Doppler coefficient and the sodium density coefficient are the two most important coefficients in understanding the ULOF transient behaviour. In particular, the uncertainty in evaluating the sodium density coefficient distribution has the largest impact on the description of reactor dynamics. This is because the maximum sodium temperature rise takes place at the top of the core and in the sodium plenum.

  15. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    JNES has been developing a technical data base used in reviewing validation of core analysis methods of LWRs in the occasions: (1) confirming core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical data base, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The data base will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects (1) measurements of Doppler reactivity in experimental MOX core simulating LWR cores, (2) measurement of isotopic compositions of fission product nuclides on high-burn up BWR UO2 fuels and (3) neutronics analysis of the experimental data that has been obtained in the international joint programs such as FUBILA and REBUS. (author)

  16. Identifying the core competence as a key requirement for business model innovation. The case of Airlines as a service industry

    OpenAIRE

    Nair, Sujith Krishnan Suchithra; Paulose, Hanna; Palacios Fernández, Miguel; Tafur Segura, José Javier

    2012-01-01

    Core competencies form the basis of an organization’s skills and the basic element of a successful strategic execution. Identifying and strengthening the core competencies enhances flexibility thereby strategically positioning a firm for responding to competition in the dynamic marketplace and can be the difference in quality among firms that follow the same business model. A correct understanding of the concept of business models, employing the right core competencies, organizing them ef...

  17. An Industrial Application of Resource-Based View of Firm in China: Identifying Core Competencies and Achieving Sustainability

    OpenAIRE

    Zhou, Wuzhen

    2013-01-01

    This thesis attempts to explore the theory application within one China based organisation, in this way to address the void in by studying the implications of applying resource based view, core competency and dynamic capability concepts to foreign organisation business management, by identifying the core competencies and the dynamic capabilities of the company. The theoretical framework consists of the strategy concept, resource based view, critical assessments about RBV, core competency,...

  18. The Randomised Controlled Trial in Medical Research: Using Bibliometric Methods to Identify Core Journals. A review of: Tsay, Migh-yueh, and Yen-hsu Yang. “Bibliometric Analysis of the Literature of Randomized Controlled Trials.” Journal of the Medical Library Association 93.4 (October 2005: 450-58.

    Directory of Open Access Journals (Sweden)

    John Loy

    2006-03-01

    Full Text Available Objective – To explore the characteristics and distribution of randomized controlled trials (RCTs in the medical literature. The study aims to identify the growth patterns of the RCT, key subject matter, country and language of publication, and determine a list of core journals which contain a substantial proportion of the RCT literature. Design – Retrospective analysis of RCTs. Setting – Medical journal literature. Subjects – A total of 160,213 articles published between 1965‐2001. Detailed analysis of a subset numbering 114,850 articles published from 1990‐2001. Methods – The study seeks to identify all RCTs in MEDLINE from 1965‐2001, and examines the growth rate of the RCT. The authors then do a more detailed analysis on a subset of data from 1990‐2001, using Access database and Excel spreadsheet software, and PERL programming language. The references were analyzed by five fields within MEDLINE; publication type, source, language, country of publication, and descriptor (subject index. Main results – An exponential growth rate for the RCT is demonstrated, suggesting that in the medical literature development has not yet matured and that research using this method continues to grow. A growth rate for the RCT of 11.2% per annum is identified. The most common form of publication is the journal article, making up approximately 98% of the RCT literature. Approximately 75% of the RCTs are multicentre trials indicating that this is the design of choice adopted by researchers. The United States proves to be the greatest source of RCT literature, with 39.9% of journals and 50.6% of articles originating there. After the USA, the most productive countries are England (15.8% of journals and 21.7% articles and Germany (6.5% journals and 6.1% articles. As might be expected, English is the predominant language providing 92.9% of the total publications. Of the remaining 7%, German is the most common language accounting for 2.2%. The top

  19. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author).

  20. COMPARATIVE ANALYSIS OF SINGLE-CORE AND MULTI-CORE SYSTEMS

    OpenAIRE

    Ogundairo Johnson; Omosehinmi Dinyo

    2015-01-01

    Overall performance of computer systems are better investigated and evaluated when its various components are considered, components such as the hardware, software and firmware. The comparative analysis of single-core and multi-core systems was carried out using Intel Pentium G640T 2.4GHz dualcore, Intel Pentium IV 2.4GHz single-core and Intel Pentium IV 2.8GHz single-core systems. The approach method was using hi-tech benchmarking and stress testing software(s) to examine systems...

  1. Full MOX BWR core physics experiment. Experimental and analysis results of 9x9 reference core

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) has been conducting an experimental program that is aimed to obtain a comprehensive data base for validation of core analysis methods applied to the full MOX ABWR and also for high burn up MOX fuel expected in the future. As a part of this program, JNES has been performing a MOX core physics experimental program, FUBILA, with collaboration of a French Consortium (CEA and COGEMA). The experiments has been designed to obtain the core physics data of the operating conditions of the full MOX BWR cores consisting of high burn up BWR MOX assemblies. The experiments started from January 2005 and completed in September 2006 at the EOLE critical facility of the CEA Cadarache center in France. Theoretical analysis of the experimental data has been also carried out with a continuous energy Monte Carlo calculation and a deterministic method with major nuclear data libraries. This report presents the outline of the FUBILA program, the measured data of the critical mass and the power distribution of a 9x9 Ref core, the first experimental core of the seven cores of FUBILA, and the core analysis by the continuous energy Monte Carlo calculation code MVP. (author)

  2. Burnup dependent core neutronic analysis for PBMR

    International Nuclear Information System (INIS)

    The strategy for core neutronics modeling is based on SCALE4.4 code KENOV.a module that uses Monte Carlo calculational methods. The calculations are based on detailed unit cell and detailed core modeling. The fuel pebble is thoroughly modeled by introducing unit cell modeling for the graphite matrix and the fuel kernels in the pebble. The core is then modeled by placing these pebbles randomly throughout the core, yet not loosing track of any one of them. For the burnup model, a cyclic manner is adopted by coupling the KENOV.a and ORIGEN-S modules. Shifting down one slice at each discrete time step, and inserting fresh fuel from the top, this cyclic calculation model continues until equilibrium burnup cycle is achieved. (author)

  3. Hybrid Analysis of Engine Core Noise

    Science.gov (United States)

    O'Brien, Jeffrey; Kim, Jeonglae; Ihme, Matthias

    2015-11-01

    Core noise, or the noise generated within an aircraft engine, is becoming an increasing concern for the aviation industry as other noise sources are progressively reduced. The prediction of core noise generation and propagation is especially challenging for computationalists since it involves extensive multiphysics including chemical reaction and moving blades in addition to the aerothermochemical effects of heated jets. In this work, a representative engine flow path is constructed using experimentally verified geometries to simulate the physics of core noise. A combustor, single-stage turbine, nozzle and jet are modeled in separate calculations using appropriate high fidelity techniques including LES, actuator disk theory and Ffowcs-Williams Hawkings surfaces. A one way coupling procedure is developed for passing fluctuations downstream through the flowpath. This method effectively isolates the core noise from other acoustic sources, enables straightforward study of the interaction between core noise and jet exhaust, and allows for simple distinction between direct and indirect noise. The impact of core noise on the farfield jet acoustics is studied extensively and the relative efficiency of different disturbance types and shapes is examined in detail.

  4. Identifying and Using ‘Core Competencies’ to Help Design and Assess Undergraduate Neuroscience Curricula

    OpenAIRE

    Kerchner, Michael; Hardwick, Jean C.; Thornton, Janice E.

    2012-01-01

    There has been a growing emphasis on the use of core competencies to design and inform curricula. Based on our Faculty for Undergraduate Neuroscience workshop at Pomona we developed a set of neuroscience core competencies. Following the workshop, faculty members were asked to complete an online survey to determine which core competencies are considered most essential and the results are presented. Backward Design principles are then described and we discuss how core competencies, through a ba...

  5. Identifying marker typing incompatibilities in linkage analysis.

    OpenAIRE

    Stringham, H M; Boehnke, M.

    1996-01-01

    A common problem encountered in linkage analyses is that execution of the computer program is halted because of genotypes in the data that are inconsistent with Mendelian inheritance. Such inconsistencies may arise because of pedigree errors or errors in typing. In some cases, the source of the inconsistencies is easily identified by examining the pedigree. In others, the error is not obvious, and substantial time and effort are required to identify the responsible genotypes. We have develope...

  6. Analysis and research status of severe core damage accidents

    International Nuclear Information System (INIS)

    The Severe Core Damage Research and Analysis Task Force was established in Nuclear Safety Research Center, Tokai Research Establishment, JAERI, in May, 1982 to make a quantitative analysis on the issues related with the severe core damage accident and also to survey the present status of the research and provide the required research subjects on the severe core damage accident. This report summarizes the results of the works performed by the Task Force during last one and half years. The main subjects investigated are as follows; (1) Discussion on the purposes and necessities of severe core damage accident research, (2) proposal of phenomenological research subjects required in Japan, (3) analysis of severe core damage accidents and identification of risk dominant accident sequences, (4) investigation of significant physical phenomena in severe core damage accidents, and (5) survey of the research status. (author)

  7. Application of Raman spectroscopy to identify microcalcifications and underlying breast lesions at stereotactic core needle biopsy.

    Science.gov (United States)

    Barman, Ishan; Dingari, Narahara Chari; Saha, Anushree; McGee, Sasha; Galindo, Luis H; Liu, Wendy; Plecha, Donna; Klein, Nina; Dasari, Ramachandra Rao; Fitzmaurice, Maryann

    2013-06-01

    Microcalcifications are a feature of diagnostic significance on a mammogram and a target for stereotactic breast needle biopsy. Here, we report development of a Raman spectroscopy technique to simultaneously identify microcalcification status and diagnose the underlying breast lesion, in real-time, during stereotactic core needle biopsy procedures. Raman spectra were obtained ex vivo from 146 tissue sites from fresh stereotactic breast needle biopsy tissue cores from 33 patients, including 50 normal tissue sites, 77 lesions with microcalcifications, and 19 lesions without microcalcifications, using a compact clinical system. The Raman spectra were modeled on the basis of the breast tissue components, and a support vector machine framework was used to develop a single-step diagnostic algorithm to distinguish normal tissue, fibrocystic change (FCC), fibroadenoma, and breast cancer, in the absence and presence of microcalcifications. This algorithm was subjected to leave-one-site-out cross-validation, yielding a positive predictive value, negative predictive value, sensitivity, and specificity of 100%, 95.6%, 62.5%, and 100% for diagnosis of breast cancer (with or without microcalcifications) and an overall accuracy of 82.2% for classification into specific categories of normal tissue, FCC, fibroadenoma, or breast cancer (with and without microcalcifications). Notably, the majority of breast cancers diagnosed are ductal carcinoma in situ (DCIS), the most common lesion associated with microcalcifications, which could not be diagnosed using previous Raman algorithm(s). Our study shows the potential of Raman spectroscopy to concomitantly detect microcalcifications and diagnose associated lesions, including DCIS, and thus provide real-time feedback to radiologists during such biopsy procedures, reducing nondiagnostic and false-negative biopsies. PMID:23729641

  8. Preliminaries on core image analysis using fault drilling samples; Core image kaiseki kotohajime (danso kussaku core kaisekirei)

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, T.; Ito, H. [Geological Survey of Japan, Tsukuba (Japan)

    1996-05-01

    This paper introduces examples of image data analysis on fault drilling samples. The paper describes the following matters: core samples used in the analysis are those obtained from wells drilled piercing the Nojima fault which has moved in the Hygoken-Nanbu Earthquake; the CORESCAN system made by DMT Corporation, Germany, used in acquiring the image data consists of a CCD camera, a light source and core rotation mechanism, and a personal computer, its resolution being about 5 pixels/mm in both axial and circumferential directions, and 24-bit full color; with respect to the opening fractures in core samples collected by using a constant azimuth coring, it was possible to derive values of the opening width, inclination angle, and travel from the image data by using a commercially available software for the personal computer; and comparison of this core image with the BHTV record and the hydrophone VSP record (travel and inclination obtained from the BHTV record agree well with those obtained from the core image). 4 refs., 4 figs.

  9. Tajoura reactor core conversion neutrons analysis

    International Nuclear Information System (INIS)

    This paper presents the preliminary neutronics studies and results of the Tajoura reactor core conversion calculations from currently used highly enriched (80% U235) fuel to low enriched fuel (36% U''2''3''5) by using the TAJN computer package. The compact core loading consists of 16 fuel assemblies type IRT-2M surrounded by removable and stationary beryllium reflector and ordinary water for moderation and cooling. The study was undertaken to compare results of TAJN computer package and the vendor documented results. The results of these calculations at the BOL and EOL conditions with equilibrium Xe at 10 MWt are shown. (author)

  10. Identifiable Data Files - Medicare Provider Analysis and ...

    Data.gov (United States)

    U.S. Department of Health & Human Services — The Medicare Provider Analysis and Review (MEDPAR) File contains data from claims for services provided to beneficiaries admitted to Medicare certified inpatient...

  11. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  12. Identifying MMORPG Bots: A Traffic Analysis Approach

    Directory of Open Access Journals (Sweden)

    Wen-Chin Chen

    2008-11-01

    Full Text Available Massively multiplayer online role playing games (MMORPGs have become extremely popular among network gamers. Despite their success, one of MMORPG's greatest challenges is the increasing use of game bots, that is, autoplaying game clients. The use of game bots is considered unsportsmanlike and is therefore forbidden. To keep games in order, game police, played by actual human players, often patrol game zones and question suspicious players. This practice, however, is labor-intensive and ineffective. To address this problem, we analyze the traffic generated by human players versus game bots and propose general solutions to identify game bots. Taking Ragnarok Online as our subject, we study the traffic generated by human players and game bots. We find that their traffic is distinguishable by 1 the regularity in the release time of client commands, 2 the trend and magnitude of traffic burstiness in multiple time scales, and 3 the sensitivity to different network conditions. Based on these findings, we propose four strategies and two ensemble schemes to identify bots. Finally, we discuss the robustness of the proposed methods against countermeasures of bot developers, and consider a number of possible ways to manage the increasingly serious bot problem.

  13. Full-Core Pin by Pin Subchannel Analysis of SMART

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Kyong Won; Hwang, Dae Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    MATRA-S, a subchannel analysis code was modified to analyze full-core of SMART with pin by pin subchannel model without lumping channels. The SMART core has 57 fuel assemblies of 17x17 arrays with 264 fuel rods and 25 tubes and there are total of 15,048 fuel rods and 16,780 subchannels as shown in Fig.1. Subchannel analysis of a whole reactor core usually has modeled by lumping channels. It lumps several subchannels or assemblies into a big virtual channel and reduces a number of channels and a calculation size of the problem by more than few orders of magnitude. The subchannel analysis models for evaluation of thermal margin of SMART core are 44 and 39 channels for the 1/8-symmetry core as show in Fig.2. The models were developed to be simple that they can be evaluated within reasonable time but to be conservative to a reference model. The reference model is a pin by pin 1/8-core subchannel analysis model and it has 2,333 subchannels and total of 2,119 fuel rods and tubes. The lumped models adopted several engineering factors and assumptions for conservativeness. As the computing power increases drastically, a single stage, full core pin by pin subchannel analysis has become true. This full real subchannel analysis model can obtain more operating margins than lumped models if it is applicable within reasonable time and cost

  14. Core Competence Analysis--Toyota Production System

    Institute of Scientific and Technical Information of China (English)

    钱璐宜

    2013-01-01

      Core competencies are the wel spring of new business development. It is the sharpest sword to penetrate the mature market, hold and enlarge the existing share. Toyota makes wel use of its TPS and form its own style which other car manufacturers hard to imitate.In contrast,the Chinese company---FAW only imitating the superficial aspects from Toyota and ignoring its own problems in manufacture line.

  15. Two evolved supernova remnants with newly identified Fe-rich cores in the Large Magellanic Cloud

    CERN Document Server

    Kavanagh, Patrick J; Bozzetto, Luke M; Points, Sean D; Crawford, Evan J; Dickel, John; Filipovic, Miroslav D; Haberl, Frank; Maggi, Pierre; Whelan, Emma T

    2016-01-01

    Aims. We present a multi-wavelength analysis of the evolved supernova remnants MCSNR J0506-7025 and MCSNR J0527-7104 in the Large Magellanic Cloud. Methods. We used data from XMM-Newton, the Australian Telescope Compact Array, and the Magellanic Cloud Emission Line Survey to study their broadband emission and used Spitzer and HI data to gain a picture of their environments. We performed a multi-wavelength morphological study and detailed radio and X-ray spectral analyses to determine their physical characteristics. Results. Both remnants were found to have bright X-ray cores, dominated by Fe L-shell emission, consistent with reverse shock heated ejecta with determined Fe masses in agreement with Type Ia explosion yields. A soft X-ray shell, consistent with swept-up interstellar medium, was observed in MCSNR J0506-7025, suggestive of a remnant in the Sedov phase. Using the spectral fit results and the Sedov self-similar solution, we estimated the age of MCSNR J0506-7025 to be ~16-28 kyr, with an initial explos...

  16. Comparative genomics of 12 strains of Erwinia amylovora identifies a pan-genome with a large conserved core.

    Directory of Open Access Journals (Sweden)

    Rachel A Mann

    Full Text Available The plant pathogen Erwinia amylovora can be divided into two host-specific groupings; strains infecting a broad range of hosts within the Rosaceae subfamily Spiraeoideae (e.g., Malus, Pyrus, Crataegus, Sorbus and strains infecting Rubus (raspberries and blackberries. Comparative genomic analysis of 12 strains representing distinct populations (e.g., geographic, temporal, host origin of E. amylovora was used to describe the pan-genome of this major pathogen. The pan-genome contains 5751 coding sequences and is highly conserved relative to other phytopathogenic bacteria comprising on average 89% conserved, core genes. The chromosomes of Spiraeoideae-infecting strains were highly homogeneous, while greater genetic diversity was observed between Spiraeoideae- and Rubus-infecting strains (and among individual Rubus-infecting strains, the majority of which was attributed to variable genomic islands. Based on genomic distance scores and phylogenetic analysis, the Rubus-infecting strain ATCC BAA-2158 was genetically more closely related to the Spiraeoideae-infecting strains of E. amylovora than it was to the other Rubus-infecting strains. Analysis of the accessory genomes of Spiraeoideae- and Rubus-infecting strains has identified putative host-specific determinants including variation in the effector protein HopX1(Ea and a putative secondary metabolite pathway only present in Rubus-infecting strains.

  17. Identifying the AD 1257 Salamas volcanic event from micron-size tephra composition in two East Antarctic ice cores

    Science.gov (United States)

    Petit, Jean Robert; Narcisi, Biancamaria; Batanova, Valentina G.; Joël, Savarino; Komorowski, Jean Christophe; Michel, Agnes; Metrich, Nicole; Besson, Pascale; Vidal, Celine; Sobolev, Alexander V.

    2016-04-01

    A wealth of valuable data about the history of explosive volcanic history can be extracted from polar ice successions. Both the volatile by-products and the solid silicate (tephra) components of volcanic plumes can be incorporated into snow layers, providing tools for chronostratigraphic correlations and for interpretation of climate-volcanism interactions. Volcanic events from low-latitude regions are of particular interest as the related sulphate aerosol travelling through the stratosphere can reach the polar sheets forming inter-hemispheric (Greenland and Antarctica) signals preserved in the ice. Within the glaciological record of globally significant volcanic markers, the AD1259 signal represents one of most prominent events over the last thousands years. Its source has been long debated. On the basis of recent field investigations (Lavigne et al., 2013; Vidal et al., 2015), it has been proposed that Mount Samalas on Lombok Island (Indonesia) represents the source responsible for the polar event. With the goal of bringing distal tephrochronological evidence to source identification, we have attempted to identify volcanic ash associated to the AD 1259 sulphate pulse. To this purpose we used firn and ice-core samples from two East Antarctic Plateau sites: Concordia-Dome C (75°06' S, 123°20' E, 3233 m) and Talos Dome (72°49'S, 159°11'E, 2315 m). Our high-resolution studies included sample processing in a Class 100 clean room using established ultra-clean procedures for insoluble microparticle analyses, Coulter counter grain size measurements, scanning electron microscope observations and the geochemical (major elements) composition from the recently set ISTERRE Jeol JXA 8230 Superprobe and calibrated for small particles analysis. Despite the difficulty of studying such minute fragments, within both cores we located and characterised multiple tiny (micron-size) glass shards concomitant with the volcanic peak. We present preliminary results alongside comparison

  18. Analysis for core conversion HEU-LEU) of PARR-2

    International Nuclear Information System (INIS)

    Calculational methodology for conversion of Miniature Neutron Source Reactor (MNSR) from HEU to LEU was validated by doing analysis of HEU fuel (90.2% enriched). On the basis of HEU based reactor model, analysis of LEU (UO/sub 2/ fuel) core gives results, which qualify the UO/sub 2/ fuel for future LEU core of MNSR. However for LEU fuel, neutron flux at irradiation sites is slightly lower for the reactor operating at 30 kW power. Therefore reactor power will have to be increased to a level of 33 kW to get the same thermal flux values as obtained for HEU core. Use of the same control rod as being used in the current HEU core gives lower values of shut down margin and control rod worth. But the slightly increased diameter of control rod improves shut down margin to a value that is comparable to the corresponding value for HEU core. LEU (UO/sub 2/ fuelled) core with following characteristics provides replica of the currently operating HEU core: 'Enrichment: 12.46%' Guide tube and grid plate material: Zr-4 'Reactor power: 3.3kW' Cladding material of fuel pin: Zr-4/' Control rod absorber (cadmium) thickness: 4.5 mm All other materials and structures have been assumed to be same as are being used in the presently operating HEU core. There is no significant difference between the dose values for HEU and prospected LEU fuel. Therefore existing HEU core and prospected LEU core of MNSR are considered to be safe for the public even in case of an accident releasing radioactive gases from the fuel. (orig./A.B.)

  19. Nuclear Core Analysis Technology Development for Future Fuel

    International Nuclear Information System (INIS)

    As for the supporting work for the irradiation test of MOX fuel rods, the fuel irradiation test data including the reactor operating conditions and the linear power generation rates for each MOX fuel rods for cycle 4 through 8 were collected. The estimated fuel burnup of test fuel rods and the radial power distribution within the fuel pellet were calculated and supplied to the fuel rod design group in order to assess the fuel rod behavior. The determination of the irradiation conditions of the test fuel rods for succeeding cycles were also carried out. As for the development of core analysis methodology, the nuclear analysis code verification data were acquired through the participation in the international programs organized by NEA. The acquired data includes the MOX core transient benchmark problem, the draft benchmark specification of VENUS-7, 9 and 17, and the B and W critical experiment data. The experiment data from the commercial power reactor operation with MOX fuel rods were also acquired. The capability of predicting the static and transient behavior of the MOX fueled core of the nuclear core analysis code, NUREC, was validated against the above the verification data. This core analysis technology was transferred to the industry related with nuclear core design

  20. Analysis and Application of the Series Core Snubber

    Science.gov (United States)

    Xie, Fei; Li, Ge; Cheng, Desheng; Chen, Qiangjian

    2013-05-01

    The transformer core snubber (CS), as one of the most important components in the EAST (experimental advanced superconducting tokamak) NBI (neutral beam injector) system, is designed to limit grid damage and protect the ion source during periods of electrical breakdowns. A transformer core snubber is analyzed in detail in this paper. Several kinds of soft magnetic cores are presented and compared. With analysis and experiment on the basic characteristics of the cores, the most suitable materials are suggested. The circuit simulation code is established which could simulate faulty conditions with concentrated and distributed CS concepts. Based on the above work, an ion source CS is developed with series type of distributed topology. The CS has been subjected to experimental validation at 80 kV with a peak short-current of approximately 400 A in a real NBI system, which proves the accuracy of the adopted assumptions and the analysis method.

  1. Analysis and Application of the Series Core Snubber

    International Nuclear Information System (INIS)

    The transformer core snubber (CS), as one of the most important components in the EAST (experimental advanced superconducting tokamak) NBI (neutral beam injector) system, is designed to limit grid damage and protect the ion source during periods of electrical breakdowns. A transformer core snubber is analyzed in detail in this paper. Several kinds of soft magnetic cores are presented and compared. With analysis and experiment on the basic characteristics of the cores, the most suitable materials are suggested. The circuit simulation code is established which could simulate faulty conditions with concentrated and distributed CS concepts. Based on the above work, an ion source CS is developed with series type of distributed topology. The CS has been subjected to experimental validation at 80 kV with a peak short-current of approximately 400 A in a real NBI system, which proves the accuracy of the adopted assumptions and the analysis method. (fusion engineering)

  2. Core conversion effects on the safety analysis of research reactors

    International Nuclear Information System (INIS)

    The safety related parameters of the 5 MW Democritus research reactor that will be affected by the scheduled core conversion to use LEU instead of HEU are considered. The analysis of the safety related items involved in such a core conversion, mainly the consequences due to MCA, DBA, etc., is of a general nature and can, therefore, be applied to other similar pool type reactors as well. (T.A.)

  3. TMI-2 accident: core heat-up analysis

    International Nuclear Information System (INIS)

    This report summarizes NSAC study of reactor core thermal conditions during the accident at Three Mile Island, Unit 2. The study focuses primarily on the time period from core uncovery (approximately 113 minutes after turbine trip) through the initiation of sustained high pressure injection (after 202 minutes). The transient analysis is based upon established sequences of events; plant data; post-accident measurements; interpretation or indirect use of instrument responses to accident conditions

  4. Development of VHTR Core Analysis and Verification Methodology

    International Nuclear Information System (INIS)

    The primary objective of this project is to develop a three dimensional cylindrical geometry code to analyze PBRs using the Analytic Function Expansion Nodal (AFEN) method, and the second objective is to produce the numerical data and to verify the deterministic code from commercial PBR core and prism reactor core using the Monte Carlo method. We developed the TOPS code and verified its validity with various benchmark problems for stead-state and transient conditions. Considering the pebble flow and temperature distribution within the core, the core analysis for commercial pebble-type reactor was carried out by using the Monte Carlo method and the spatial-dependent Dancoff factors, and also was evaluated with the Monte Carlo method. And the optimization of the decay chain model, and implementation of the multi-group cross section processing of DeCART using McCARD for double heterogeneity effect. The TOPS code can be used in VHTR's design and reactor core characteristics evaluation, and the Monte Carlo results of core analysis can be used to the verification of the deterministic code. Furthermore, they are expected that the analysis method can be installed in the deterministic code

  5. Development of VHTR Core Analysis and Verification Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Han Gyu; Kim, Chang Hyo; Park, Ho Jin [Seoul National University, Seoul (Korea, Republic of)] (and others)

    2009-03-15

    The primary objective of this project is to develop a three dimensional cylindrical geometry code to analyze PBRs using the Analytic Function Expansion Nodal (AFEN) method, and the second objective is to produce the numerical data and to verify the deterministic code from commercial PBR core and prism reactor core using the Monte Carlo method. We developed the TOPS code and verified its validity with various benchmark problems for stead-state and transient conditions. Considering the pebble flow and temperature distribution within the core, the core analysis for commercial pebble-type reactor was carried out by using the Monte Carlo method and the spatial-dependent Dancoff factors, and also was evaluated with the Monte Carlo method. And the optimization of the decay chain model, and implementation of the multi-group cross section processing of DeCART using McCARD for double heterogeneity effect. The TOPS code can be used in VHTR's design and reactor core characteristics evaluation, and the Monte Carlo results of core analysis can be used to the verification of the deterministic code. Furthermore, they are expected that the analysis method can be installed in the deterministic code.

  6. Transient analysis for PWR reactor core using neural networks predictors

    International Nuclear Information System (INIS)

    In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions

  7. Thermal hydraulic analysis of the JMTR improved LEU-core

    Energy Technology Data Exchange (ETDEWEB)

    Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

    2003-01-01

    After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

  8. HTGR core seismic analysis using an array processor

    International Nuclear Information System (INIS)

    A Floating Point Systems array processor performs nonlinear dynamic analysis of the high-temperature gas-cooled reactor (HTGR) core with significant time and cost savings. The graphite HTGR core consists of approximately 8000 blocks of various shapes which are subject to motion and impact during a seismic event. Two-dimensional computer programs (CRUNCH2D, MCOCO) can perform explicit step-by-step dynamic analyses of up to 600 blocks for time-history motions. However, use of two-dimensional codes was limited by the large cost and run times required. Three-dimensional analysis of the entire core, or even a large part of it, had been considered totally impractical. Because of the needs of the HTGR core seismic program, a Floating Point Systems array processor was used to enhance computer performance of the two-dimensional core seismic computer programs, MCOCO and CRUNCH2D. This effort began by converting the computational algorithms used in the codes to a form which takes maximum advantage of the parallel and pipeline processors offered by the architecture of the Floating Point Systems array processor. The subsequent conversion of the vectorized FORTRAN coding to the array processor required a significant programming effort to make the system work on the General Atomic (GA) UNIVAC 1100/82 host. These efforts were quite rewarding, however, since the cost of running the codes has been reduced approximately 50-fold and the time threefold. The core seismic analysis with large two-dimensional models has now become routine and extension to three-dimensional analysis is feasible. These codes simulate the one-fifth-scale full-array HTGR core model. This paper compares the analysis with the test results for sine-sweep motion

  9. TRACE analysis of Phenix core response to an increase of the core inlet sodium temperature

    International Nuclear Information System (INIS)

    This work presents the analysis, using the TRACE code, of the Phenix core response to an inlet sodium temperature increase. The considered experiment was performed in the frame of the Phenix End-Of-Life (EOL) test program of the CEA, prior to the final shutdown of the reactor. It corresponds to a transient following a 40°C increase of the core inlet temperature, which leads to a power decrease of 60%. This work focuses on the first phase of the transient, prior to the reactor scram and pump trip. First, the thermal-hydraulic TRACE model of the core developed for the present analysis is described. The kinetic parameters and feedback coefficients for the point kinetic model were first derived from a 3D static neutronic ERANOS model developed in a former study. The calculated kinetic parameters were then optimized, before use, on the basis of the experimental reactivity in order to minimize the error on the power calculation. The different reactivity feedbacks taken into account include various expansion mechanisms that have been specifically implemented in TRACE for analysis of fast-neutron spectrum systems. The point kinetic model has been used to study the sensitivity of the core response to the different feedback effects. The comparison of the calculated results with the experimental data reveals the need to accurately calculate the reactivity feedback coefficients. This is because the reactor response is very sensitive to small reactivity changes. This study has enabled us to study the sensitivity of the power change to the different reactivity feedbacks and define the most important parameters. As such, it furthers the validation of the FAST code system, which is being used to gain a more in-depth understanding of SFR core behavior during accidental transients. (author)

  10. Assumed mode approach to fast reactor core seismic analysis

    International Nuclear Information System (INIS)

    The need for a time history approach, rather than a response spectrum approach, to the seismic analysis of fast breeder reactor core structures is described. The use of a Rayleigh-Ritz/Assumed Mode formalism for developing mathematical models of reactor cores is presented. Various factors including structural nonlinearity, fluid inertia, and impact which necessitate abandonment of response spectrum methods are discussed. The use of the assumed mode formalism is described in some detail as it applies to reactor core seismic analysis. To illustrate the use of this formal approach to mathematical modeling, a sample reactor problem with increasing complexities of modeling is presented. Finally, several problem areas--fluid inertia, fluid damping, coulomb friction, impact, and modal choice--are discussed with emphasis on research needs for use in fast reactor seismic analysis

  11. Core thermal hydraulic analysis for TNR power uprating

    International Nuclear Information System (INIS)

    This paper presents preliminary results of a study undertaken to investigate the possibility of raising the power of the Tajura Nuclear Research Reactor (TNRR) from 10 to 20 MWt keeping the same core configuration and with minimum changes in the primary cooling circuit. The study was carried out for a fresh core, with compact load (16 assemblies) under normal operation conditions. A computer program, TAJT, was used to simulate the core and perform the necessary thermal hydraulic analysis. The results obtained show that the reactor power could be raised to 15 MWt safely and with no changes in the primary cooling circuit. To raise the power to 20 MWt will require changes in the core configuration and primary circuit

  12. Tank 241-B-203 Push Mode Core Sampling and Analysis Plan

    Energy Technology Data Exchange (ETDEWEB)

    Jo, J. [Westinghouse Hanford Co., Richland, WA (United States)

    1995-05-16

    This Sampling and Analysis Plan (SAP) will identify characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements in accordance with the Tank Safety Screening Data Quality Objective (Babad and Redus 1994). This Data Quality Objective (DQO) is described in the Tank Characterization Plan (Jo, 1995) for tank 241-B-203 (B-203). This SAP will also identify procedures and requirements for collecting and characterizing samples from tank B-203 by the core sampling method.

  13. Analysis of Irradiation Holes of In-Core Region

    International Nuclear Information System (INIS)

    Test fuels and materials are irradiated in the in-core region in side of the chimney. The inner chimney is composed of In-Core and Out-Core regions. The In-Core region has 23 hexagonal vertical irradiation holes named from R01 to R20, CT, IR1 and IR2 and 8 cylindrical irradiation holes named from CAR1 to CAR4 and SOR1 to SOR4. The Out-Core region is composed of 8 cylindrical irradiation holes named from OR1 to OR8 which are installed near the inner shell of the reflector tank. HANARO is the multi-purpose research reactor which utilizes in-core irradiation holes, which is being used in various field. Over the past 7 years we have used CT 8 times, IR once, IR2 and OR3 twice, OR4 three times and OR5 ten times. These irradiation holes are used to perform an evaluation of the neutron irradiation properties and the tests were all completed and done successfully. HANARO has been used successfully, and it still will be used continuously in various fields such as nuclear in-pile tests, the production of radioisotopes, neutron transmutation doping, neutron activation analysis, neutron beam research, radiography, environmental science

  14. Analysis of Irradiation Holes of In-Core Region

    Energy Technology Data Exchange (ETDEWEB)

    In, Won-ho; Lee, Yong-sub; Kim, Tae-hwan; Lim, Kyoung-hwan; Ahn, Hyung-jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Test fuels and materials are irradiated in the in-core region in side of the chimney. The inner chimney is composed of In-Core and Out-Core regions. The In-Core region has 23 hexagonal vertical irradiation holes named from R01 to R20, CT, IR1 and IR2 and 8 cylindrical irradiation holes named from CAR1 to CAR4 and SOR1 to SOR4. The Out-Core region is composed of 8 cylindrical irradiation holes named from OR1 to OR8 which are installed near the inner shell of the reflector tank. HANARO is the multi-purpose research reactor which utilizes in-core irradiation holes, which is being used in various field. Over the past 7 years we have used CT 8 times, IR once, IR2 and OR3 twice, OR4 three times and OR5 ten times. These irradiation holes are used to perform an evaluation of the neutron irradiation properties and the tests were all completed and done successfully. HANARO has been used successfully, and it still will be used continuously in various fields such as nuclear in-pile tests, the production of radioisotopes, neutron transmutation doping, neutron activation analysis, neutron beam research, radiography, environmental science.

  15. Core Handling and Real-Time Non-Destructive Characterization at the Kochi Core Center: An Example of Core Analysis from the Chelungpu Fault

    Directory of Open Access Journals (Sweden)

    W. Lin

    2007-11-01

    Full Text Available As an example of core analysis carried out inactive fault drilling programs, we report the procedures of core handling on the drilling site and non-destructive characterization in the laboratory. This analysis was employed onthe core samples taken from HoleBof the Taiwan Chelungpu-fault Drilling Project (TCDP, which penetrated through the active fault that slipped during the 1999 Chi-Chi, Taiwan earthquake. We show results of the non-destructive physical property measurements carried out at the Kochi Core Center (KCC, Japan. Distinct anomalies of lower bulk density and higher magnetic susceptibilitywere recognized in all three fault zones encountered in HoleB. To keep the core samples in good condition before they are used for variousanalyses is crucial. In addition, careful planning for core handlingand core analyses is necessary for successfulinvestigations. doi:10.2204/iodp.sd.s01.35.2007

  16. Preliminary Core Analysis for Regulatory Evaluation of SFR Nuclear Designs

    International Nuclear Information System (INIS)

    In Korea, a conceptual design of SFR has been developed by Korea Atomic Energy Research Institute (KAERI). An application for the design approval of a prototype SFR is scheduled in 2017. In order to prepare the licensing of a prototype SFR, Korea Institute of Nuclear Safety (KINS) is developing the regulatory audit code system for SFR since 2012. The SFR nuclear evaluation system for regulatory verification has an object to verify core integral parameters, reactivity coefficients, and peaking power factor, etc. provided by the designers and generate reactivity coefficients and kinetic parameters for safety analyses. For these purpose, both stochastic (Monte Carlo) and deterministic tools can be used. In deterministic core analysis, the PARCS code is being considered for the whole-core steady-state and time-dependent, multi-group hexagonal diffusion calculations and additional code modules for cross-section generation and more accurate transport solution, etc. are planned to be developed from 2013 or be introduced. This paper presents the development strategy of SFR nuclear evaluation system for regulatory verification and the preliminary core analysis results of Korean SFR demonstration reactor with 600MWe (DEMO-600). The DEMO-600 core is preliminarily analyzed using PARCS, and global reactivity worths such as uniform expansion coefficients, rod worths are calculated by direct eigenvalue differences of the base and perturbed configurations. The results calculated by PARCS are compared with the DIF3D nodal solutions provided by KAERI. (author)

  17. TRR-1/M1 Core Analysis with MVP

    International Nuclear Information System (INIS)

    Full text: Since early 1990s, the in-core fuel management of TRR-1/M1 has been performed by TRIGAP. This code was specifically developed for reactor physics calculations of the TRIGA-type reactor. However, because of its limitations in geometrical and cross sectional options, the attempt of using other techniques/codes are provoked. Nowadays, the choice of using the Monte Carlo method to perform core analysis becomes more satisfaction with acceptable computational time. The MVP is one of the codes that utilize the Monte Carlo method with continuous-energy library. It is able to explicitly model the problem in 3-D geometry. It also has a burn-up calculation feature called MVP-BURN. The aim of the current work is to apply the MVP code for TRR-1/M1 core analysis. In this paper, the MVP code was verified with the experiment results for the fresh core and some burn-up cores. The calculated-eigenvalue results agree well with the experimental data within an acceptable range of statistical error

  18. Stress analysis of portable safety platform (Core Sampler Truck)

    International Nuclear Information System (INIS)

    This document provides the stress analysis and evaluation of the portable platform of the rotary mode core sampler truck No. 2 (RMCST number-sign 2). The platform comprises railing, posts, deck, legs, and a portable ladder; it is restrained from lateral motion by means of two brackets added to the drill-head service platform

  19. Methodology for reactor core physics analysis - part 2

    International Nuclear Information System (INIS)

    The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs

  20. Species-level core oral bacteriome identified by 16S rRNA pyrosequencing in a healthy young Arab population

    Directory of Open Access Journals (Sweden)

    Nezar Noor Al-hebshi

    2016-05-01

    Full Text Available Background: Reports on the composition of oral bacteriome in Arabs are lacking. In addition, the majority of previous studies on other ethnic groups have been limited by low-resolution taxonomic assignment of next-generation sequencing reads. Furthermore, there has been a conflict about the existence of a ‘core’ bacteriome. Objective: The objective of this study was to characterize the healthy core oral bacteriome in a young Arab population at the species level. Methods: Oral rinse DNA samples obtained from 12 stringently selected healthy young subjects of Arab origin were pyrosequenced (454's FLX chemistry for the bacterial 16S V1–V3 hypervariable region at an average depth of 11,500 reads. High-quality, non-chimeric reads ≥380 bp were classified to the species level using the recently described, prioritized, multistage assignment algorithm. A core bacteriome was defined as taxa present in at least 11 samples. The Chao2, abundance-based coverage estimator (ACE, and Shannon indices were computed to assess species richness and diversity. Results: Overall, 557 species-level taxa (211±42 per subject were identified, representing 122 genera and 13 phyla. The core bacteriome comprised 55 species-level taxa belonging to 30 genera and 7 phyla, namely Firmicutes, Proteobacteria, Actinobacteria, Bacteroidetes, Fusobacteria, Saccharibacteria, and SR1. The core species constituted between 67 and 87% of the individual bacteriomes. However, the abundances differed by up to three orders of magnitude among the study subjects. On average, Streptococcus mitis, Rothia mucilaginosa, Haemophilus parainfluenzae, Neisseria flavescence/subflava group, Prevotella melaninogenica, and Veillonella parvula group were the most abundant. Streptococcus sp. C300, a taxon never reported in the oral cavity, was identified as a core species. Species richness was estimated at 586 (Chao2 and 614 (ACE species, whereas diversity (Shannon index averaged at 3.99. Conclusions

  1. Magnetic resonance imaging in laboratory petrophysical core analysis

    Science.gov (United States)

    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

  2. CORE

    DEFF Research Database (Denmark)

    Krigslund, Jeppe; Hansen, Jonas; Hundebøll, Martin;

    2013-01-01

    different flows. Instead of maintaining these approaches separate, we propose a protocol (CORE) that brings together these coding mechanisms. Our protocol uses random linear network coding (RLNC) for intra- session coding but allows nodes in the network to setup inter- session coding regions where flows...... increase the benefits of XORing by exploiting the underlying RLNC structure of individual flows. This goes beyond providing additional reliability to each individual session and beyond exploiting coding opportunistically. Our numerical results show that CORE outperforms both forwarding and COPE......-like schemes in general. More importantly, we show gains of up to 4 fold over COPE-like schemes in terms of transmissions per packet in one of the investigated topologies....

  3. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  4. Performance Analysis: Work Control Events Identified January - August 2010

    Energy Technology Data Exchange (ETDEWEB)

    De Grange, C E; Freeman, J W; Kerr, C E; Holman, G; Marsh, K; Beach, R

    2011-01-14

    This performance analysis evaluated 24 events that occurred at LLNL from January through August 2010. The analysis identified areas of potential work control process and/or implementation weaknesses and several common underlying causes. Human performance improvement and safety culture factors were part of the causal analysis of each event and were analyzed. The collective significance of all events in 2010, as measured by the occurrence reporting significance category and by the proportion of events that have been reported to the DOE ORPS under the ''management concerns'' reporting criteria, does not appear to have increased in 2010. The frequency of reporting in each of the significance categories has not changed in 2010 compared to the previous four years. There is no change indicating a trend in the significance category and there has been no increase in the proportion of occurrences reported in the higher significance category. Also, the frequency of events, 42 events reported through August 2010, is not greater than in previous years and is below the average of 63 occurrences per year at LLNL since 2006. Over the previous four years, an average of 43% of the LLNL's reported occurrences have been reported as either ''management concerns'' or ''near misses.'' In 2010, 29% of the occurrences have been reported as ''management concerns'' or ''near misses.'' This rate indicates that LLNL is now reporting fewer ''management concern'' and ''near miss'' occurrences compared to the previous four years. From 2008 to the present, LLNL senior management has undertaken a series of initiatives to strengthen the work planning and control system with the primary objective to improve worker safety. In 2008, the LLNL Deputy Director established the Work Control Integrated Project Team to develop the core requirements and graded

  5. Method and device for identifying various objects, especially reactor core elements

    International Nuclear Information System (INIS)

    This invention represents an alternative to the mechanical and logistic methods for identification of objects as used so far especially in the field of reactor technology. For example, a measuring head with a primary coil and two rings of secondary coils used for determining the position of fuel elements equipped with two ferromagnetic pieces is an appropriate device to fulfil these new tasks after having some modification in the analyzing electronics. To this effect, the objects to be identified are equipped with several ferromagnetic pieces, the orientation and/or position and/or size and/or number of which are characteristic for the different objects. These features can be detected with the measuring head and the different objects can thus be identified. A certain technique is proposed in particular, determining the identity of the object by means of two azimuth angle measurements of the alignment of the connecting liner of two pairs of ferromagnetic pieces fixed to the object. (orig.)

  6. Spectroscopic modeling and analysis of plasma conditions in implosion cores

    Science.gov (United States)

    Golovkin, Igor E.

    In this dissertation we discuss the effects of opacity and plasma gradients on the analysis and interpretation of Ar K-shell line emission from Ar-doped inertial confinement fusion (ICF) experiments, and introduce a spectroscopic technique for the determination of core plasma gradients. In particular, the Ar Heβ composite spectral feature is used for core plasma temperature and density diagnostics. We present a versatile, spectroscopic-quality Non-Local-Thermodynamic- Equilibrium radiation transport model that takes into account the effects of collisional-radiative atomic kinetics, plasma gradients, Stark-broadened line shapes and radiation transport. The code computes the radiative properties of the plasma, and it can be easily adapted to treat different problems of spectra formation. We discuss the importance of high-order satellite emission in the formation of Heβ spectral feature, and the interpretation of core averaged electron temperatures and densities extracted from space integrated spectra of non- uniform plasmas. We also present an application of Genetic Algorithms to the analysis of experimental X-ray spectra. This algorithm drives the search for plasma parameters that yield the best fits to experimental spectra. We discuss the applicability of Case Injected Genetic Algorithms to accelerate analysis of spectra. Furthermore, we introduce a novel method for the determination of plasma temperature and density gradients in imploded cores. The gradients are extracted from the self-consistent analysis of time-resolved X-ray spectra and spatial emissivity distributions obtained from X-ray monochromatic images. In this case, the search in the complex parameter space of gradient functions is driven by a multi-objective Niched Pareto Genetic Algorithm. We discuss the analysis of time resolved spectra recorded during Ar-doped ICF implosions at the NOVA laser facility. Time histories of core averaged electron densities and temperatures during the collapse of the

  7. Preliminary analysis of the proposed BN-600 benchmark core

    International Nuclear Information System (INIS)

    The Indira Gandhi Centre for Atomic Research is actively involved in the design of Fast Power Reactors in India. The core physics calculations are performed by the computer codes that are developed in-house or by the codes obtained from other laboratories and suitably modified to meet the computational requirements. The basic philosophy of the core physics calculations is to use the diffusion theory codes with the 25 group nuclear cross sections. The parameters that are very sensitive is the core leakage, like the power distribution at the core blanket interface etc. are calculated using transport theory codes under the DSN approximations. All these codes use the finite difference approximation as the method to treat the spatial variation of the neutron flux. Criticality problems having geometries that are irregular to be represented by the conventional codes are solved using Monte Carlo methods. These codes and methods have been validated by the analysis of various critical assemblies and calculational benchmarks. Reactor core design procedure at IGCAR consists of: two and three dimensional diffusion theory calculations (codes ALCIALMI and 3DB); auxiliary calculations, (neutron balance, power distributions, etc. are done by codes that are developed in-house); transport theory corrections from two dimensional transport calculations (DOT); irregular geometry treated by Monte Carlo method (KENO); cross section data library used CV2M (25 group)

  8. Fault tree analysis on BWR core spray system

    International Nuclear Information System (INIS)

    Fault Trees which describe the failure modes for the Core Spray System function in the Browns Ferry Nuclear Plant (BWR 1065MWe) were developed qualitatively and quantitatively. The unavailability for the Core Spray System was estimated to be 1.2 x 10-3/demand. It was found that the miscalibration of four reactor pressure sensors or the failure to open of the two inboard valves (FCV 75-25 and 75-53) could reduce system reliability significantly. It was recommended that the pressure sensors would be calibrated independently. The introduction of the redundant inboard valves could improve the system reliability. Thus this analysis method was verified useful for system analysis. The detailed test and maintenance manual and the informations on the control logic circuits of each active component are necessary for further analysis. (author)

  9. Identifying the genetic diversity, genetic structure and a core collection of Ziziphus jujuba Mill. var. jujuba accessions using microsatellite markers

    Science.gov (United States)

    Xu, Chaoqun; Gao, Jiao; Du, Zengfeng; Li, Dengke; Wang, Zhe; Li, Yingyue; Pang, Xiaoming

    2016-01-01

    Ziziphus is a genus of spiny shrubs and small trees in the Rhamnaceae family. This group has a controversial taxonomy, with more than 200 species described, including Chinese jujube (Ziziphus jujuba Mill. var. jujuba) and Indian jujube (Z. mauritiana), as well as several other important cultivated fruit crops. Using 24 SSR markers distributed across the Chinese jujube genome, 962 jujube accessions from the two largest germplasm repositories were genotyped with the aim of analyzing the genetic diversity and structure and constructing a core collection that retain high genetic diversity. A molecular profile comparison revealed 622 unique genotypes, among which 123 genotypes were genetically identical to at least one other accessions. STRUCTURE analysis and multivariate analyses (Cluster and PCoA) roughly divided the accessions into three major groups, with some admixture among groups. A simulated annealing algorithm and a heuristic algorithm were chosen to construct the core collection. A final core of 150 accessions was selected, comprising 15.6% of the analyzed accessions and retaining more than 99.5% of the total alleles detected. We found no significant differences in allele frequency distributions or in genetic diversity parameters between the chosen core accessions and the 622 genetically unique accessions. This work contributes to the understanding of Chinese jujube diversification and the protection of important germplasm resources. PMID:27531220

  10. Core and Conal Component Analysis of Pulsar B1933+16 --- Investigation of the Segregated Modes

    CERN Document Server

    Mitra, Dipanjan; Arjunwadkar, Mihir

    2016-01-01

    Radio pulsar B1933+16 is brightest core-radiation dominated pulsar in the Arecibo sky, and here we carry out a comprehensive high resolution polarimetric study of its radiation at both 1.5 and 4.6 GHz. At 1.5 GHz, the polarization is largely compatible with a rotating-vector model with $\\alpha$ and $\\beta$ values of 125 and --1.2$^{\\circ}$, such that the core and conal regions can be identified with the primary and secondary polarization modes and plausibly with the extraordinary and ordinary propagation modes. Polarization modal segregation of profiles shows that the core is comprised of two parts which we associate with later X-mode and earlier O-mode emission. Analysis of the broad microstructures under the core shows that they have similar timescales to those of the largely conal radiation of other pulsars studied earlier. Aberration/retardation analysis was here possible for both the conal and core radiation and showed average physical emission heights of about 200 km for each. Comparison with other core...

  11. Study of reference material for NMR core analysis

    International Nuclear Information System (INIS)

    Reference Materials of NMR core experimental analysis have been studied systematically in this paper. According to the national standard criterion, a set of NMR reference materials have been made successfully. These reference materials include fluid pattern-D2O and H2O, divergence of glass grain muster, and the solid ceramics. They have been applied to the core analysis in Daqing, Xinjiang, Dagang Oilfield, etc. The results show, fluid pattern, divergence of glass grain muster as well as ceramics reference materials have different calibration results of core NMR porosity as a result of their different relaxation mechanisms. Fluid pattern is suitable for NMR porosity calibration in Berea sandstone as well as terrestrial sediment sand with little shale content and well sorted. Ceramics reference material is applied to shaly sand with average sorted. Both fluid pattern and solid ceramics material are not suitable for calibration in sandstone sample with high clay content, complex lithology with rich paramagnetic materials. It is suggested that the representative natural core sample should be selected as regional reference material to calibrate NMR porosity of complex lithology, or internal magnetic field gradient in complex rock grain/fluid system should be studied in order to get corrected NMR porosity. (authors)

  12. Analysis of core damage frequency: Surry, Unit 1 internal events

    International Nuclear Information System (INIS)

    This document contains the accident sequence analysis of internally initiated events for the Surry Nuclear Station, Unit 1. This is one of the five plant analyses conducted as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 documents the risk of a selected group of nuclear power plants. The work performed and described here is an extensive of that published in November 1986 as NUREG/CR-4450, Volume 3. It addresses comments form numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved. The context and detail of this report are directed toward PRA practitioners who need to know how the work was performed and the details for use in further studies. The mean core damage frequency at Surry was calculated to be 4.05-E-5 per year, with a 95% upper bound of 1.34E-4 and 5% lower bound of 6.8E-6 per year. Station blackout type accidents (loss of all AC power) were the largest contributors to the core damage frequency, accounting for approximately 68% of the total. The next type of dominant contributors were Loss of Coolant Accidents (LOCAs). These sequences account for 15% of core damage frequency. No other type of sequence accounts for more than 10% of core damage frequency. 49 refs., 52 figs., 70 tabs

  13. 100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT

    Energy Technology Data Exchange (ETDEWEB)

    HARRINGTON RA

    2010-01-15

    On December 15-16, 2009, a 100-KE Reactor Core Removal Project Alternative Analysis Workshop was conducted at the Washington State University Consolidated Information Center, Room 214. Colburn Kennedy, Project Director, CH2M HILL Plateau Remediation Company (CHPRC) requested the workshop and Richard Harrington provided facilitation. The purpose of the session was to select the preferred Bio Shield Alternative, for integration with the Thermal Shield and Core Removal and develop the path forward to proceed with project delivery. Prior to this workshop, the S.A. Robotics (SAR) Obstruction Removal Alternatives Analysis (565-DLV-062) report was issued, for use prior to and throughout the session, to all the team members. The multidisciplinary team consisted ofrepresentatives from 100-KE Project Management, Engineering, Radcon, Nuclear Safety, Fire Protection, Crane/Rigging, SAR Project Engineering, the Department of Energy Richland Field Office, Environmental Protection Agency, Washington State Department of Ecology, Defense Nuclear Facility Safety Board, and Deactivation and Decommission subject matter experts from corporate CH2M HILL and Lucas. Appendix D contains the workshop agenda, guidelines and expectations, opening remarks, and attendance roster going into followed throughout the workshop. The team was successful in selecting the preferred alternative and developing an eight-point path forward action plan to proceed with conceptual design. Conventional Demolition was selected as the preferred alternative over two other alternatives: Diamond Wire with Options, and Harmonic Delamination with Conventional Demolition. The teams preferred alternative aligned with the SAR Obstruction Removal Alternative Analysis report conclusion. However, the team identified several Path Forward actions, in Appendix A, which upon completion will solidify and potentially enhance the Conventional Demolition alternative with multiple options and approaches to achieve project delivery

  14. High Resolution Continuous Flow Analysis System for Polar Ice Cores

    Science.gov (United States)

    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

  15. Preliminary core mechanics analysis for KALIMER by CRAMP code

    International Nuclear Information System (INIS)

    CRAMP code is designed to solve the problem of mutually interacting and distorting sub-assemblies in a fast breeder reactor. It is the UK's main core mechanics design tool and is currently being used in the design of EFR. This report contains the results of preliminary core mechanics calculations for KALIMER core configuration by the updated version of CRAMP code. The base case calculation s on KALIMER core, and the sensitivity studies (to investigate effect of main design parameter) are carried out by the code which was updated with material subroutine in CRAMP to model the characteristics of HT9. Sensitivity studies include following cases; (1) with gaps at LRP and URP reduced to 0.4 mm at 386 dg C (2) with 0.2 mm radial clearance around both nosepiece at seals (3) with flexibility at LRP reduced by a factor of 2 (4) with stiffness of nosepiece increased by a factor of 2 (5) with reduced creep (6) combined with gap reduced 0.4 mm and 0.2 mm clearance at seals (7) with IVS position replaced to dummy ducts (8) with initial bow at every duct except S/A 1. From each calculation, the data obtained and compared are as follows; (a) contact forces between pads (b) gaps between pads (c) duct dynamic behavior of duct bowing and dilation i.e. the variation of bowing and dilation with time) (d) reactivity change (e) maximum LRP contact force, maximum URP contact force and maximum nosepiece force vs interval number for the base case. The design requirements and the specifications for KALIMER assembly ducts are reviewed, and preliminary core mechanics analysis for KALIMER core configuration are carried out. (Author). 7 refs., 2 tabs., 50 figs

  16. Preliminary Core Analysis of a Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Chang Keun; Chang, Jongwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Venneri, Francesco [Ultra Safe Nuclear Corporation, Los Alamos (United States); Hawari, Ayman [NC State Univ., Raleigh (United States)

    2014-05-15

    The Micro Modular Reactor (MMR) will be 'melt-down proof'(MDP) under all circumstances, including the complete loss of coolant, and will be easily transportable and retrievable, and suitable for use with very little site preparation and Balance of Plant (BOP) requirements for a variety of applications, from power generation and process heat applications in remote areas to grid-unattached locations, including ship propulsion. The Micro Modular Reactor design proposed in this paper has 3 meter diameter core (2 meter active core) which is suitable for 'factory manufactured' and has few tens year of service life for remote deployment. We confirmed the feasibility of long term service life by a preliminary neutronic analysis in terms of the excess reactivity, the temperature feedback coefficient, and the control margins. We are able to achieve a reasonably long core life time of 5 ∼ 10 years under typical thermal hydraulic condition of a helium cooled reactor. However, on a situation where longer service period and safety is important, we can reduce the power density to the level of typical pebble bed reactor. In this case we can design 10 MWt MMR with core diameter for 10 ∼ 40 years core life time without much loss in the economics. Several burnable poisons are studied and it is found that erbia mixed in the compact matrix seems reasonably good poison. The temperature feedback coefficients were remaining negative during lifetime. Drum type control rods at reflector region and few control rods inside core region are sufficient to control the reactivity during operation and to achieve safe cold shutdown state.

  17. Thermal Analysis of Air-Core Power Reactors

    OpenAIRE

    Zhao Yuan; Jun-jia He; Yuan Pan; Xiao-gen Yin; Can Ding; Shao-fei Ning; Hong-lei Li

    2013-01-01

    A fluid-thermal coupled analysis based on FEM is conducted. The inner structure of the coils is built with consideration of both the structural details and the simplicity; thus, the detailed heat conduction process is coupled with the computational fluid dynamics in the thermal computation of air-core reactors. According to the simulation results, 2D temperature distribution results are given and proved by the thermal test results of a prototype. Then the temperature results are used to calcu...

  18. Magnetic loss analysis in Mn-Zn ferrite cores

    International Nuclear Information System (INIS)

    Magnetic losses have been measured and analyzed upon a wide range of frequencies in Mn-Zn ferrite ring cores. Exploiting the concept of loss separation and modeling the conductivity process in the heterogeneous material as a function of frequency, the role of the different energy dissipation mechanisms has been elucidated. It is shown, in particular, that eddy current effects can be appreciated, in standard materials and cores, only on approaching and overcoming the MHz range. The basic mechanism for hysteresis and low-frequency losses is therefore identified with the domain wall relaxation engendered by spin damping processes. Resonant absorption of energy associated with magnetization rotation is in turn deemed to chiefly contribute to the loss upon the practical range of frequencies going from a few 104Hz to a few MHz

  19. Analysis of Homogeneous BFS-73-1 MA Benchmark Core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Il; Yoo, Jae Woon; Song, Hoon; Jang, Jin Wook; Kim, Yeong Il

    2007-06-15

    Analysis of BFS-73-1 critical assembly for MA transmutation has been carried out by using K-CORE system mainly, DIF3D code. All of measured data are compared with the results of analysis and sensitiveness of calculation conditions, for example, number of neutron energy groups, mesh size used, and analysis method, are assessed. Effective multiplication factor was in good agreement within experimental uncertainty in both transport and diffusion calculations. Fission rate distribution of U-235 and U-238 is also fairly good agreed with experimental results within maximum 5% in core region. But large discrepancy was seen in blanket region and it tends to increase as the location closes to core boundary. Largest error of relative reaction rate ratio was seen in Am-243 fission and U-238 capture. For the case of Am-243, the error lay on appropriate range considering the measurement uncertainty of that as 4.6%. Sample reactivity worths for scattering dominant isotope was greatly differ from the experimental results, which can be explained in terms of sample heterogeneity effect, sample self shielding and finally resonance bilinear correction effect. These effects will be evaluated as future study. C/E of effective delayed neutron fraction is within 4%, which is within the measurement uncertainty.

  20. Noise analysis of Forsmark 1 data to investigate BWR core local instability

    International Nuclear Information System (INIS)

    BWR core local instability was experienced at Forsmark 1 (F1) during reactor operation in cycle 16. The event has been studied by applying noise analysis and stability calculations to get insight into the event as well as to identify the cause of local instability. The present report is concerned with noise analysis of data collected during start-up in cycle 17. The results of the current study indicates: The F1 core is quite stable in cycle 17. The max. decay ratio (DR) value of 0.37 was obtained from the stability evaluation of an APRM (average power range monitor) and LPRM (local power range monitor) signals measured at 66% (APRM) of reactor power and 4252 Kg/s (SA-HC) of core flow. Compared with the power profile in cycle 17 (as well as in reactor F2), the core in cycle 16 had an extreme power profile with high power and bottom-shifted axial peak in the core periphery esp. at the four quadrant corners. Such a profile decreases the stability margin in the region. It is a common observation that the DR obtained from APRM tends to be higher than that from LPRM if the global instability mechanism is dominant in the core, and vice versa. The comparison of global and local DR values should be an effective method for detecting local instability during the reactor operation. In order to detect the local instability it is important to evaluate the core stability with sufficient number of LPRMs so as to cover the whole core cross section together with APRMs

  1. Monte Carlo analysis of Musashi TRIGA mark II reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Tetsuo [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan)

    1999-08-01

    The analysis of the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). Effective multiplication factors (k{sub eff}) for the several fuel-loading patterns including the initial core criticality experiment, the fuel element and control rod reactivity worth as well as the neutron flux measurements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k{sub eff} overestimated the experimental data by about 1.0%{delta}k/k for both the initial core and the several fuel-loading arrangements. The calculated reactivity worths of control rod and fuel element agree well the measured ones within the uncertainties. The comparison of neutron flux distribution was consistent with the experimental ones which were measured by activation methods at the sample irradiation tubes. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicated that the Monte Carlo model is enough to simulate the Musashi TRIGA-II reactor core. (author)

  2. Core disruptive accident analysis using ASTERIA-FBR

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) is developing a core disruptive accident analysis code, ASTERIA-FBR, which tightly couples the thermal-hydraulics and the neutronics to simulate the core behavior during core disruptive accidents (CDA) of fast breeder reactors (FBRs). ASTERIA-FBR consists of the three-dimensional thermal-hydraulics calculation module: CONCORD, the fuel pin behavior calculation module: FEMAXI-FBR, and the space-time neutronics module: Dynamic-GMVP or PARTISN/RKIN. This paper describes a comparison between characteristics of GMVP and PARTISN and summarizes the challenging issues on applying Dynamic-GMVP to the calculation against unprotected loss-of-flow (ULOF) event which is a typical initiator of core disruptive accident of FBR. It was found that Dynamic-GMVP is confirmed to be basically applicable to the CDA phenomena. It was found that, however, applying GMVP to the CDA calculation is less reasonable than PARTISN since the calculation load of GMVP is too large to meet the required calculation accuracy, although the Monte-Carlo method is based on the actual neutron behavior without any discretization of space and energy. The statistical error included in the calculation results may affect the super-prompt criticality during ULOF event and thus the amount of released energy

  3. An Efficient Analysis Methodology for Fluted-Core Composite Structures

    Science.gov (United States)

    Oremont, Leonard; Schultz, Marc R.

    2012-01-01

    The primary loading condition in launch-vehicle barrel sections is axial compression, and it is therefore important to understand the compression behavior of any structures, structural concepts, and materials considered in launch-vehicle designs. This understanding will necessarily come from a combination of test and analysis. However, certain potentially beneficial structures and structural concepts do not lend themselves to commonly used simplified analysis methods, and therefore innovative analysis methodologies must be developed if these structures and structural concepts are to be considered. This paper discusses such an analysis technique for the fluted-core sandwich composite structural concept. The presented technique is based on commercially available finite-element codes, and uses shell elements to capture behavior that would normally require solid elements to capture the detailed mechanical response of the structure. The shell thicknesses and offsets using this analysis technique are parameterized, and the parameters are adjusted through a heuristic procedure until this model matches the mechanical behavior of a more detailed shell-and-solid model. Additionally, the detailed shell-and-solid model can be strategically placed in a larger, global shell-only model to capture important local behavior. Comparisons between shell-only models, experiments, and more detailed shell-and-solid models show excellent agreement. The discussed analysis methodology, though only discussed in the context of fluted-core composites, is widely applicable to other concepts.

  4. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author)

  5. LWR core safety analysis with Areva's 3-dimensional methods

    International Nuclear Information System (INIS)

    The quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools and an extensive validation base. Sophisticated 3-dimensional core models ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. The validation base includes measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models achieve reliable and comprehensive results for a wide range of applications. As an example an overview of the application experience as well as the validation base of AREVA's 3-dimensional codes is given. The importance and necessity of the comprehensive 3-dimensional methodology is illustrated with examples of a BWR and PWR safety analysis. For BWR transient application the analysis of regional power oscillations is considered and regarding the PWR safety analysis an example referring to fast enthalpy rise and the maximum fuel temperature caused by a rod ejection accident is shown. (orig.)

  6. BIOELECTRICAL IMPEDANCE VECTOR ANALYSIS IDENTIFIES SARCOPENIA IN NURSING HOME RESIDENTS

    Science.gov (United States)

    Loss of muscle mass and water shifts between body compartments are contributing factors to frailty in the elderly. The body composition changes are especially pronounced in institutionalized elderly. We investigated the ability of single-frequency bioelectrical impedance analysis (BIA) to identify b...

  7. SEDIMENT CORE SAMPLING AND ANALYSIS OF KAW LAKE

    Directory of Open Access Journals (Sweden)

    Dejene Alemayehu

    2014-01-01

    Full Text Available The Kaw Nation and Black land Research Center in July 2012 conducted a sediment core sampling from Kaw Lake. Kaw Lake is a reservoir constructed in 1976 by the Army Corps of Engineers for the purpose of water supply and recreation. It is located 11 miles east of Ponca City, Kay County, Oklahoma. This reservoir covers approximately 17,040 acres (69 km2 and is also known to be the seventh largest lake in Oklahoma by surface area. This lake holds 428,600 acre feet (528,700, 00 m3 of water which is said to be the ninth largest lake in Oklahoma. The lake is fed by the Arkansas River that flows from Colorado, through kansas and into Kaw Lake. The Arkansas River flows through Kaw Lake shared by several small creeks and empties into the Mississippi River. The purpose of the study was to determine the rate of sediment accumulation and examine the level of nutrient and heavy metals accumulation or deposition at the bottom of the lake. Four core samples from different parts of the lake at different depth were sampled for analysis. Each core sample was sectioned into 20 cm, dried and ground into homogenous powder. Samples from each section were tested for organic carbon content and heavy metals. Organic carbon content was verified by burning through a muffle furnace, while the remaining core samples were digested into a solution and ran through an Atomic Absorption Spectrophotometer (AAS to evaluate the concentration of heavy metals. Particle size analyses were also determined. Results were organized by depth, concentration, particle size distribution and bulk density. Data showed phosphorous and some heavy metals concentrations at core 3 and 4 were higher than core 1 and 2. Phosphorous concentration at four sediment core sites ranged from 350 mg kg-1 to 550 mg kg-1. Whereas Aluminum concentration was 40,000 mg kg-1 to 70,000 mg kg-1, Barium 280 mg kg-1 to 420 mg kg-1

  8. Development of the transient analysis of the LMR core

    International Nuclear Information System (INIS)

    FX2-TH is a two-dimensional multigroup fine-mesh diffusion code for the transient analysis of the LMR core and has been designed to treat transients initiated by such mechanism as sodium voiding, motion of fuel and/or structural material, and control rod motion. It takes account of feedback effects from changes in both the fuel temperature and average coolant temperatures through Doppler broadening and change in coolant density, respectively. However, the thermal expansion of the fuel and structural material causes one of the most important reactivity feedback effects on the transient behavior of the LMR core loaded with metallic fuels. This paper describes the thermal reactivity feedback model implemented into the code and the three-dimensional, steady-state nodal diffusion code for the hexagonal-z geometry, which has been developed to serve as a standard neutronics solver of a future transient code for LMR's

  9. Identifying flares in rheumatoid arthritis: reliability and construct validation of the OMERACT RA Flare Core Domain Set

    Science.gov (United States)

    Bykerk, Vivian P; Bingham, Clifton O; Choy, Ernest H; Lin, Daming; Alten, Rieke; Christensen, Robin; Furst, Daniel E; Hewlett, Sarah; Leong, Amye; March, Lyn; Woodworth, Thasia; Boire, Gilles; Haraoui, Boulos; Hitchon, Carol; Jamal, Shahin; Keystone, Edward C; Pope, Janet; Tin, Diane; Thorne, J Carter

    2016-01-01

    Objective To evaluate the reliability of concurrent flare identification using 3 methods (patient, rheumatologist and Disease Activity Score (DAS)28 criteria), and construct validity of candidate items representing the Outcome Measures in Rheumatology Clinical Trials (OMERACT) RA Flare Core Domain Set. Methods Candidate flare questions and legacy measures were administered at consecutive visits to Canadian Early Arthritis Cohort (CATCH) patients between November 2011 and November 2014. The American College of Rheumatology (ACR) core set indicators were recorded. Concordance to identify flares was assessed using the agreement coefficient. Construct validity of flare questions was examined: convergent (Spearman's r); discriminant (mean differences between flaring/non-flaring patients); and consequential (proportions with prior treatment reductions and intended therapeutic change postflare). Results The 849 patients were 75% female, 81% white, 42% were in remission/low disease activity (R/LDA), and 16–32% were flaring at the second visit. Agreement of flare status was low–strong (κ's 0.17–0.88) and inversely related to RA disease activity level. Flare domains correlated highly (r's≥0.70) with each other, patient global (r's≥0.66) and corresponding measures (r's 0.49–0.92); and moderately highly with MD and patient-reported joint counts (r's 0.29–0.62). When MD/patients agreed the patient was flaring, mean flare domain between-group differences were 2.1–3.0; 36% had treatment reductions prior to flare, with escalation planned in 61%. Conclusions Flares are common in rheumatoid arthritis (RA) and are often preceded by treatment reductions. Patient/MD/DAS agreement of flare status is highest in patients worsening from R/LDA. OMERACT RA flare questions can discriminate between patients with/without flare and have strong evidence of construct and consequential validity. Ongoing work will identify optimal scoring and cut points to identify RA flares. PMID

  10. Seismic response analysis for prismatic fuel HTGR core

    International Nuclear Information System (INIS)

    For high-temperature gas cooled reactors (HTGR) with prismatic fuels, their resistance against an earthquake is not fully ascertained yet. Aseismic design studies and also experiments must therefore be made when such a reactor plant is to be installed in areas of high seismicity. This report describes analytical study on the seismic response of a prismatic fuel reactor core, including the following: aseismic core structure, the analysis model and calculation formulae, the effects of various design variables on response charactersitics, and the desired block shape. Three analysis models have been considered for the seismic vibration of the prismatic fuel HTGR core. The first is the impact model, the second ''the spring dashpot model'', and the third ''the dryfriction model''. The calculation has been performed with three models, and these results are nearly the same. The followings were revealed: (1) At low input-wave frequencies, the response value increases with the gap between the blocks. Beyond a certain point, however, the effect of gap is nearly negligible. (2) When the blocks are restrained horizontally by keys, the response value decreases with increase of the key stiffness. The key is thus effective in earthquake resistance. (3) The response value increases with block-stiffness, so that short massive blocks are better for earthquake resistance. (4) The response value decreases with increase of the block damping factor. But beyond a certain point, this effect is only small. (5) Stiffness and damping in the restraint structure for the reactor core do not have much effect in earthquake resistance. (author)

  11. Experimental Analysis and Modeling of the Crushing of Honeycomb Cores

    Science.gov (United States)

    Aminanda, Y.; Castanié, B.; Barrau, J.-J.; Thevenet, P.

    2005-05-01

    In the aeronautical field, sandwich structures are widely used for secondary structures like flaps or landing gear doors. The modeling of low velocity/low energy impact, which can lead to a decrease of the structure strength by 50%, remains a designer’s main problem. Since this type of impact has the same effect as quasi-static indentation, the study focuses on the behavior of honeycomb cores under compression. The crushing phenomenon has been well identified for years but its mechanism is not described explicitly and the model proposed may not satisfy industrial purposes. To understand the crushing mechanism, honeycomb test specimens made of Nomex™, aluminum alloy and paper were tested. During the crushing, a CCD camera showed that the cell walls buckled very quickly. The peak load recorded during tests corresponded to the buckling of the common edge of three honeycomb cells. Further tests on corner structures to simulate only one vertical edge of a honeycomb cell show a similar behavior. The different specimens exhibited similar load/displacement curves and the differences observed were only due to the behavior of the different materials. As a conclusion of this phenomenological study, the hypothesis that loads are mainly taken by the vertical edge can be made. So, a honeycomb core subjected to compression can be modeled by a grid of nonlinear springs. A simple analytical model was then developed and validated by tests on Nomex™ honeycomb core indented by different sized spherical indenters. A good correlation between theory and experiment was found. This result can be used to satisfactorily model using finite elements the indentation on a sandwich structure with a metallic or composite skin and honeycomb core.

  12. Analysis of core calculation schemes for advanced water reactors

    International Nuclear Information System (INIS)

    This research thesis addresses the analysis of the core control of sub-moderated water reactors with plutonium fuel and varying spectrum. Firstly, a calculation scheme is defined, based on transport theory for the three existing assembly configurations. It is based on the efficiency analysis of the control cluster and of the flow sheet shape in the assembly. Secondly, studies of the assembly with control cluster and within a theory of diffusion with homogenization or detailed assembly representation are performed by taking the environment into account in order to assess errors. Thirdly, due to the presence of a very efficient absorbent in control clusters, a deeper physical analysis requires the study of the flow gradient existing at the interface between assemblies. A parameter is defined to assess this gradient, and theoretically calculated by using finite elements. Developed software is validated

  13. Identifying clinical course patterns in SMS data using cluster analysis

    DEFF Research Database (Denmark)

    Kent, Peter; Kongsted, Alice

    2012-01-01

    whole group, by including all SMS time points in their original form. It was a 'proof of concept' study to explore the potential, clinical relevance, strengths and weakness of such an approach. METHODS: This was a secondary analysis of longitudinal SMS data collected in two randomised controlled trials...... subgroups in the outcomes of research studies. Two previous studies have investigated detailed clinical course patterns in SMS data obtained from people seeking care for low back pain. One used a visual analysis approach and the other performed a cluster analysis of SMS data that had first been transformed...... conducted simultaneously from a single clinical population (n = 322) . Fortnightly SMS data collected over a year on 'days of problematic low back pain' and on 'days of sick leave' were analysed using Two-Step (probabilistic) Cluster Analysis. RESULTS: Clinical course patterns were identified that were...

  14. Design and analysis of PCRV core cavity closure

    International Nuclear Information System (INIS)

    Design requirements and considerations for a core cavity closure which led to the choice of a concrete closure with a toggle hold-down as the design for the Gas-Cooled Fast Breeder Reactor (GCFR) plant are discussed. A procedure for preliminary stress analysis of the closure by means of a three-dimensional finite element method is described. A limited parametric study using this procedure indicates the adequacy of the present closure design and the significance of radial compression developed as a result of inclined support reaction

  15. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  16. Core analysis and CT imaging improve shale completions

    International Nuclear Information System (INIS)

    To improve hydraulic fracturing efficiency in Devonian shales, core analysis and computerized tomography (CT) can provide data for orienting perforations, determining fracture direction, and selecting deviated well trajectories. This article reports on technology tested in a West Virginia well for improving the economics of developing Devonian shale and other low permeability gas reservoirs. With slight production increase per well, Columbia Natural Resources Inc. (CNR) has determined that marginal gas well payout time can be shortened enough to encourage additional drilling. For eight wells completed by CNR in 1992, the absolute open flow (AOF) averaged 116 Mcfd before stimulation. After stimulation using long-standing fracture stimulation procedures, the AOF averaged 500 Mcfd

  17. Identifying influential factors of business process performance using dependency analysis

    Science.gov (United States)

    Wetzstein, Branimir; Leitner, Philipp; Rosenberg, Florian; Dustdar, Schahram; Leymann, Frank

    2011-02-01

    We present a comprehensive framework for identifying influential factors of business process performance. In particular, our approach combines monitoring of process events and Quality of Service (QoS) measurements with dependency analysis to effectively identify influential factors. The framework uses data mining techniques to construct tree structures to represent dependencies of a key performance indicator (KPI) on process and QoS metrics. These dependency trees allow business analysts to determine how process KPIs depend on lower-level process metrics and QoS characteristics of the IT infrastructure. The structure of the dependencies enables a drill-down analysis of single factors of influence to gain a deeper knowledge why certain KPI targets are not met.

  18. Using factor analysis to identify neuromuscular synergies during treadmill walking

    Science.gov (United States)

    Merkle, L. A.; Layne, C. S.; Bloomberg, J. J.; Zhang, J. J.

    1998-01-01

    Neuroscientists are often interested in grouping variables to facilitate understanding of a particular phenomenon. Factor analysis is a powerful statistical technique that groups variables into conceptually meaningful clusters, but remains underutilized by neuroscience researchers presumably due to its complicated concepts and procedures. This paper illustrates an application of factor analysis to identify coordinated patterns of whole-body muscle activation during treadmill walking. Ten male subjects walked on a treadmill (6.4 km/h) for 20 s during which surface electromyographic (EMG) activity was obtained from the left side sternocleidomastoid, neck extensors, erector spinae, and right side biceps femoris, rectus femoris, tibialis anterior, and medial gastrocnemius. Factor analysis revealed 65% of the variance of seven muscles sampled aligned with two orthogonal factors, labeled 'transition control' and 'loading'. These two factors describe coordinated patterns of muscular activity across body segments that would not be evident by evaluating individual muscle patterns. The results show that factor analysis can be effectively used to explore relationships among muscle patterns across all body segments to increase understanding of the complex coordination necessary for smooth and efficient locomotion. We encourage neuroscientists to consider using factor analysis to identify coordinated patterns of neuromuscular activation that would be obscured using more traditional EMG analyses.

  19. Latent cluster analysis of ALS phenotypes identifies prognostically differing groups.

    Directory of Open Access Journals (Sweden)

    Jeban Ganesalingam

    Full Text Available BACKGROUND: Amyotrophic lateral sclerosis (ALS is a degenerative disease predominantly affecting motor neurons and manifesting as several different phenotypes. Whether these phenotypes correspond to different underlying disease processes is unknown. We used latent cluster analysis to identify groupings of clinical variables in an objective and unbiased way to improve phenotyping for clinical and research purposes. METHODS: Latent class cluster analysis was applied to a large database consisting of 1467 records of people with ALS, using discrete variables which can be readily determined at the first clinic appointment. The model was tested for clinical relevance by survival analysis of the phenotypic groupings using the Kaplan-Meier method. RESULTS: The best model generated five distinct phenotypic classes that strongly predicted survival (p<0.0001. Eight variables were used for the latent class analysis, but a good estimate of the classification could be obtained using just two variables: site of first symptoms (bulbar or limb and time from symptom onset to diagnosis (p<0.00001. CONCLUSION: The five phenotypic classes identified using latent cluster analysis can predict prognosis. They could be used to stratify patients recruited into clinical trials and generating more homogeneous disease groups for genetic, proteomic and risk factor research.

  20. Analysis of a basic core performance for FBR core nuclear design. 3

    International Nuclear Information System (INIS)

    The spatial distribution of reaction rates in the ZPPR-13A, having an axially heterogeneous core, has been analyzed. The ZPPR-13A core is treated as a 2-dimensional RZ configuration consisting of a homogeneous core. The analysis is performed by utilizing both probabilistic and deterministic treatments. The probabilistic treatment is performed with the Monte Carlo Code MVP running with continuous energy variable. By comparing the results obtained by both treatments and reviewing the calculation method of effective resonance cross sections, for deterministic treatment, utilized for the reaction rate distributions, it is revealed that the present treatment of effective resonance cross sections is not accurate, since there are observed effects due to dependence on energy group number or energy group width, and on anisotropic scattering. To utilize multi-band method for calculating effective resonance cross sections, widely used by the European researchers, the computer code GROUPIE is installed and the performance of the code is confirmed. Although, in order to improve effective resonance cross sections accuracy, the thermal neutron reactor standard code system SRAC-95 was introduced last year in which the ultra-fine group spectrum calculation module PEACO worked specially under the restriction that number of nuclei having resonance cross section, in any zone, should be less than three, because collision probabilities were obtained by an interpolation method. This year, the module is improved so that these collision probabilities are directly calculated, and by this improvement the highly accurate effective resonance cross sections below the energy of 40.868 keV can be calculated for whole geometrical configurations considered. To extend the application range of the module PEACO, the cross sections of sodium and structure material nuclei are prepared so that they are also represented as ultra-fine group cross sections. By such modifications of cross section library

  1. Parameter Trajectory Analysis to Identify Treatment Effects of Pharmacological Interventions

    OpenAIRE

    Tiemann, Christian A.; Vanlier, Joep; Oosterveer, Maaike H.; Albert K Groen; Hilbers, Peter A. J.; Natal A W van Riel

    2013-01-01

    The field of medical systems biology aims to advance understanding of molecular mechanisms that drive disease progression and to translate this knowledge into therapies to effectively treat diseases. A challenging task is the investigation of long-term effects of a (pharmacological) treatment, to establish its applicability and to identify potential side effects. We present a new modeling approach, called Analysis of Dynamic Adaptations in Parameter Trajectories (ADAPT), to analyze the long-t...

  2. Three Systems of Insular Functional Connectivity Identified with Cluster Analysis

    OpenAIRE

    Deen, Ben; Pitskel, Naomi B.; Kevin A. Pelphrey

    2010-01-01

    Despite much research on the function of the insular cortex, few studies have investigated functional subdivisions of the insula in humans. The present study used resting-state functional connectivity magnetic resonance imaging (MRI) to parcellate the human insular lobe based on clustering of functional connectivity patterns. Connectivity maps were computed for each voxel in the insula based on resting-state functional MRI (fMRI) data and segregated using cluster analysis. We identified 3 ins...

  3. PWR core safety analysis with 3-dimensional methods

    International Nuclear Information System (INIS)

    Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

  4. Identifying Organizational Inefficiencies with Pictorial Process Analysis (PPA

    Directory of Open Access Journals (Sweden)

    David John Patrishkoff

    2013-11-01

    Full Text Available Pictorial Process Analysis (PPA was created by the author in 2004. PPA is a unique methodology which offers ten layers of additional analysis when compared to standard process mapping techniques.  The goal of PPA is to identify and eliminate waste, inefficiencies and risk in manufacturing or transactional business processes at 5 levels in an organization. The highest level being assessed is the process management, followed by the process work environment, detailed work habits, process performance metrics and general attitudes towards the process. This detailed process assessment and analysis is carried out during process improvement brainstorming efforts and Kaizen events. PPA creates a detailed visual efficiency rating for each step of the process under review.  A selection of 54 pictorial Inefficiency Icons (cards are available for use to highlight major inefficiencies and risks that are present in the business process under review. These inefficiency icons were identified during the author's independent research on the topic of why things go wrong in business. This paper will highlight how PPA was developed and show the steps required to conduct Pictorial Process Analysis on a sample manufacturing process. The author has successfully used PPA to dramatically improve business processes in over 55 different industries since 2004.  

  5. Equilibrium Core Analysis of Two Types of Cores for the AHR

    International Nuclear Information System (INIS)

    A preliminary conceptual design of a research reactor core employing a rod type fuel assembly, has been performed. Two types of the reactor core configuration have been developed as a basic core of the Advanced HANARO Reactor (AHR). One is the aluminum block core whose coolant channels are made inside a hexagonal aluminum block for loading the fuel assemblies, the other is a flow tube core by using the same zircaloy flow tubes as the HANARO. These cores have four control rods of a shroud type. In the control rod sites, 18-element fuel assemblies of a circular type are loaded, whereas 36-element fuel assemblies of a hexagonal type are loaded in the hexagonal sites except the central site. The AHR is designed on the basis that the thermal power is 20MW and the maximum thermal neutron flux is about 4.0x1014 n/cm2/sec in the reflector region. For these two cores, MCNP calculations have been finished for the condition of loading fresh fuels and no irradiation holes and beam tubes in the reflector region. For the depleted core, the parameters such as the cycle length, fuel burnup and maximum linear power for the equilibrium core, are evaluated by using the HANARO fuel management code system

  6. Transcriptome analysis of recurrently deregulated genes across multiple cancers identifies new pan-cancer biomarkers

    DEFF Research Database (Denmark)

    Kaczkowski, Bogumil; Tanaka, Yuji; Kawaji, Hideya; Sandelin, Albin; Andersson, Robin; Itoh, Masayoshi; Lassmann, Timo; Hayashizaki, Yoshihide; Carninci, Piero; Forrest, Alistair R

    2015-01-01

    Genes that are commonly deregulated in cancer are clinically attractive as candidate pan-diagnostic markers and therapeutic targets. To globally identify such targets, we compared Cap Analysis of Gene Expression (CAGE) profiles from 225 different cancer cell lines and 339 corresponding primary cell...... samples to identify transcripts that are deregulated recurrently in a broad range of cancer types. Comparing RNA-seq data from 4,055 tumors and 563 normal tissues profiled in the TCGA and FANTOM5 datasets, we identified a core transcript set with theranostic potential. Our analyses also revealed enhancer...... RNAs which are upregulated in cancer, defining promoters which overlap with repetitive elements (especially SINE/Alu and LTR/ERV1 elements) that are often upregulated in cancer. Lastly, we documented for the first time upregulation of multiple copies of the REP522 interspersed repeat in cancer. Overall...

  7. Radiometric dating of sediment core from waterwork reservoir Rozgrund and analysis of mercury concentration depth profile

    International Nuclear Information System (INIS)

    Radioisotope dating of lake sediments combined with analysis of chemical properties of the sediment layers allow us to study the history of the human impact on nature. Undisturbed sediment layers in the core samples serve as chronicle database with information about lake ecosystem and surrounding environment in the time of deposition. A sediment core sample from the bottom of the water-work reservoir Rozgrund was collected and separated into 2 cm thick layers. Samples were analysed by HPGe spectrometry for anthropogenous Cs-137 activity. From identified peaks corresponding to nuclear tests and Chernobyl accident the sedimentation rate was calculated and the chronology of layers established. Sub-samples from each layer were prepared separately for the analysis of the Hg concentration by atomic absorption spectrometry. The results show very small variations in Hg concentrations and there is no significant trend present in the profile. (author)

  8. Developing engineering design core competences through analysis of industrial products

    DEFF Research Database (Denmark)

    Hansen, Claus Thorp; Lenau, Torben Anker

    2011-01-01

    Most product development work carried out in industrial practice is characterised by being incremental, i.e. the industrial company has had a product in production and on the market for some time, and now time has come to design a new and upgraded variant. This type of redesign project requires...... challenge in staging a course module, in which students develop knowledge, understanding and skills, which will prepare them for being able to participate in and contribute to redesign projects in industrial practice. In the course module Product Analysis and Redesign that has run for 8 years we have...... that the engineering designers have core design competences to carry through an analysis of the existing product encompassing both a user-oriented side and a technical side, as well as to synthesise solution proposals for the new and upgraded product. The authors of this paper see an educational...

  9. Pan-genome sequence analysis using Panseq: an online tool for the rapid analysis of core and accessory genomic regions

    Directory of Open Access Journals (Sweden)

    Villegas Andre

    2010-09-01

    Full Text Available Abstract Background The pan-genome of a bacterial species consists of a core and an accessory gene pool. The accessory genome is thought to be an important source of genetic variability in bacterial populations and is gained through lateral gene transfer, allowing subpopulations of bacteria to better adapt to specific niches. Low-cost and high-throughput sequencing platforms have created an exponential increase in genome sequence data and an opportunity to study the pan-genomes of many bacterial species. In this study, we describe a new online pan-genome sequence analysis program, Panseq. Results Panseq was used to identify Escherichia coli O157:H7 and E. coli K-12 genomic islands. Within a population of 60 E. coli O157:H7 strains, the existence of 65 accessory genomic regions identified by Panseq analysis was confirmed by PCR. The accessory genome and binary presence/absence data, and core genome and single nucleotide polymorphisms (SNPs of six L. monocytogenes strains were extracted with Panseq and hierarchically clustered and visualized. The nucleotide core and binary accessory data were also used to construct maximum parsimony (MP trees, which were compared to the MP tree generated by multi-locus sequence typing (MLST. The topology of the accessory and core trees was identical but differed from the tree produced using seven MLST loci. The Loci Selector module found the most variable and discriminatory combinations of four loci within a 100 loci set among 10 strains in 1 s, compared to the 449 s required to exhaustively search for all possible combinations; it also found the most discriminatory 20 loci from a 96 loci E. coli O157:H7 SNP dataset. Conclusion Panseq determines the core and accessory regions among a collection of genomic sequences based on user-defined parameters. It readily extracts regions unique to a genome or group of genomes, identifies SNPs within shared core genomic regions, constructs files for use in phylogeny programs

  10. Global stability analysis of pressurized water reactor core nonlinear system

    International Nuclear Information System (INIS)

    Determining the global stability of a pressurized water reactor (PWR) core nonlinear system is the problem to be solved. In the paper, the core nonlinear system was modeled and the linearized model of the system was obtained via the small perturbation method. According to the distributing situation of the core nonlinearity measure in the power level range based on the equilibrium manifold, seven linear models corresponding to seven power levels respectively were chosen as local models of the core and the set of seven local models was used to approximately substitute the core system. The global stability of the PWR core nonlinear system was analyzed by utilizing Lyapunov stability theory. The calculated result shows that the core nonlinear system is globally and asymptotically stable. The modeling method of the core is effective in analyzing the global stability of a PWR core nonlinear system. (authors)

  11. Statistical analysis of dynamic parameters of the core

    International Nuclear Information System (INIS)

    The transients of various types were investigated for the cores of zero power critical facilities in RRC KI and NPP. Dynamic parameters of neutron transients were explored by tool statistical analysis. Its have sufficient duration, few channels for currents of chambers and reactivity and also some channels for technological parameters. On these values the inverse period. reactivity, lifetime of neutrons, reactivity coefficients and some effects of a reactivity are determinate, and on the values were restored values of measured dynamic parameters as result of the analysis. The mathematical means of statistical analysis were used: approximation(A), filtration (F), rejection (R), estimation of parameters of descriptive statistic (DSP), correlation performances (kk), regression analysis(KP), the prognosis (P), statistician criteria (SC). The calculation procedures were realized by computer language MATLAB. The reasons of methodical and statistical errors are submitted: inadequacy of model operation, precision neutron-physical parameters, features of registered processes, used mathematical model in reactivity meters, technique of processing for registered data etc. Examples of results of statistical analysis. Problems of validity of the methods used for definition and certification of values of statistical parameters and dynamic characteristics are considered (Authors)

  12. Rice Transcriptome Analysis to Identify Possible Herbicide Quinclorac Detoxification Genes

    Directory of Open Access Journals (Sweden)

    Wenying eXu

    2015-09-01

    Full Text Available Quinclorac is a highly selective auxin-type herbicide, and is widely used in the effective control of barnyard grass in paddy rice fields, improving the world’s rice yield. The herbicide mode of action of quinclorac has been proposed and hormone interactions affect quinclorac signaling. Because of widespread use, quinclorac may be transported outside rice fields with the drainage waters, leading to soil and water pollution and environmental health problems.In this study, we used 57K Affymetrix rice whole-genome array to identify quinclorac signaling response genes to study the molecular mechanisms of action and detoxification of quinclorac in rice plants. Overall, 637 probe sets were identified with differential expression levels under either 6 or 24 h of quinclorac treatment. Auxin-related genes such as GH3 and OsIAAs responded to quinclorac treatment. Gene Ontology analysis showed that genes of detoxification-related family genes were significantly enriched, including cytochrome P450, GST, UGT, and ABC and drug transporter genes. Moreover, real-time RT-PCR analysis showed that top candidate P450 families such as CYP81, CYP709C and CYP72A genes were universally induced by different herbicides. Some Arabidopsis genes for the same P450 family were up-regulated under quinclorac treatment.We conduct rice whole-genome GeneChip analysis and the first global identification of quinclorac response genes. This work may provide potential markers for detoxification of quinclorac and biomonitors of environmental chemical pollution.

  13. Towards a Methodology for Identifying Program Constraints During Requirements Analysis

    Science.gov (United States)

    Romo, Lilly; Gates, Ann Q.; Della-Piana, Connie Kubo

    1997-01-01

    Requirements analysis is the activity that involves determining the needs of the customer, identifying the services that the software system should provide and understanding the constraints on the solution. The result of this activity is a natural language document, typically referred to as the requirements definition document. Some of the problems that exist in defining requirements in large scale software projects includes synthesizing knowledge from various domain experts and communicating this information across multiple levels of personnel. One approach that addresses part of this problem is called context monitoring and involves identifying the properties of and relationships between objects that the system will manipulate. This paper examines several software development methodologies, discusses the support that each provide for eliciting such information from experts and specifying the information, and suggests refinements to these methodologies.

  14. A Burnup Analysis of PBMR-400MWth Reactor Core

    International Nuclear Information System (INIS)

    The purpose of this study is to analyze the burnup characteristics of 400MWth PBMR using Monte Carlo method. In the world, the deterministic method is widely used to model such that system but it still has a disadvantage which is not flexible in simulating the burnup cycle. Although this method applies some techniques to increase the accuracy of calculation results but it is necessary to model this system by a suitable computer code that can verify and validate the results of the deterministic method. A method which uses a Monte Carlo technique for simulating the burnup cycle was performed. A reactor physics computer code uses in this method is MONTEBURN 2.0 which accurately and efficiently computes the neutronic and material properties of the fuel cycle. MONTEBURN is a fully automated tool that links the MCNP Monte Carlo transport code with a radioactive decay and burnup code ORIGEN. In this model, the calculations are based on a detailed core modeling using MCNP. The fuel pebble is thoroughly modeled by introducing unit cell modeling for the graphite matrix and fuel kernels in the pebble. For the burnup model, a start-up core was studied with considering the movement of pebbles. By shifting down one layer at each discrete time step and inserting fresh fuel from the top, this cyclic calculation is continued until equilibrium burnup cycle is achieved. In this study, the time dependence of multiplication factor keff, the spatial dependence of flux profile, power distribution, burnup, and inventory of isotopes in the start up process are analyzed. The results will provide the basis data of the burnup process and be also utilized as the verified data to validate a compute code for PBMR core analysis which will be developed in near future

  15. Parameter trajectory analysis to identify treatment effects of pharmacological interventions.

    Directory of Open Access Journals (Sweden)

    Christian A Tiemann

    Full Text Available The field of medical systems biology aims to advance understanding of molecular mechanisms that drive disease progression and to translate this knowledge into therapies to effectively treat diseases. A challenging task is the investigation of long-term effects of a (pharmacological treatment, to establish its applicability and to identify potential side effects. We present a new modeling approach, called Analysis of Dynamic Adaptations in Parameter Trajectories (ADAPT, to analyze the long-term effects of a pharmacological intervention. A concept of time-dependent evolution of model parameters is introduced to study the dynamics of molecular adaptations. The progression of these adaptations is predicted by identifying necessary dynamic changes in the model parameters to describe the transition between experimental data obtained during different stages of the treatment. The trajectories provide insight in the affected underlying biological systems and identify the molecular events that should be studied in more detail to unravel the mechanistic basis of treatment outcome. Modulating effects caused by interactions with the proteome and transcriptome levels, which are often less well understood, can be captured by the time-dependent descriptions of the parameters. ADAPT was employed to identify metabolic adaptations induced upon pharmacological activation of the liver X receptor (LXR, a potential drug target to treat or prevent atherosclerosis. The trajectories were investigated to study the cascade of adaptations. This provided a counter-intuitive insight concerning the function of scavenger receptor class B1 (SR-B1, a receptor that facilitates the hepatic uptake of cholesterol. Although activation of LXR promotes cholesterol efflux and -excretion, our computational analysis showed that the hepatic capacity to clear cholesterol was reduced upon prolonged treatment. This prediction was confirmed experimentally by immunoblotting measurements of SR-B1

  16. Application of Network Analysis Method to VHTR core

    International Nuclear Information System (INIS)

    A Very High Temperature Reactor (VHTR) is currently envisioned as a promising future reactor concept because of its high-efficiency and capability of generating hydrogen. Prismatic Modular Reactor (PMR) is one of the main VHTR concepts, which consists of hexagonal prismatic fuel blocks and reflector blocks made of nuclear grade graphite. However their shape could be changed by neutron damage during the reactor operation and the shape change can makes the gaps between the blocks inducing bypass flow. Most of reactor coolant flows through the coolant channel within the fuel block, but some portion of the reactor coolant bypasses to the interstitial gaps. The vertical gap and horizontal gap are called bypass gap and cross gap, respectively. CFD simulation for the full core of VHTR might be possible but it requires vast computational cost and time. Therefore, fast, flexible and reliable code is required to predict the flow distribution corresponding to the various bypass gap distribution. Consequently in this study, the flow network analysis method is applied to analyze the core flow of VHTR. The applied method was validated by comparing with SNU VHTR multiblock experiment. As a result, the calculated results show good agreements with experimental data although computational time and cost of the developed code was very small

  17. Lidar point density analysis: implications for identifying water bodies

    Science.gov (United States)

    Worstell, Bruce B.; Poppenga, Sandra; Evans, Gayla A.; Prince, Sandra

    2014-01-01

    Most airborne topographic light detection and ranging (lidar) systems operate within the near-infrared spectrum. Laser pulses from these systems frequently are absorbed by water and therefore do not generate reflected returns on water bodies in the resulting void regions within the lidar point cloud. Thus, an analysis of lidar voids has implications for identifying water bodies. Data analysis techniques to detect reduced lidar return densities were evaluated for test sites in Blackhawk County, Iowa, and Beltrami County, Minnesota, to delineate contiguous areas that have few or no lidar returns. Results from this study indicated a 5-meter radius moving window with fewer than 23 returns (28 percent of the moving window) was sufficient for delineating void regions. Techniques to provide elevation values for void regions to flatten water features and to force channel flow in the downstream direction also are presented.

  18. Stress analysis and assessment of RPV core support pads and surround bottom head

    International Nuclear Information System (INIS)

    Background: The core support pads were used to limit circumferential rotation of core barrel, and the structure integrity of the core support pads was an important factor to the safe operation of nuclear power plant. Purpose: To ensure the structure integrity of the core support pads. Methods: Three-dimensional FEA model for bottom head, core support pads and part cylinder of CAP1000 RPV was established. Thermal analysis, static analysis, fatigue analysis and fracture analysis were performed. The analysis results were evaluated according to ASME B and PVC-III-NB-3200 and ASME B and PVC-III-1-Appendix G. Results: The evaluation indicated that the core support pads and surround bottom head could satisfy related requirements of above code. Conclusions: The analysis methodology used in this paper could also be applied to the core support pads of RPV for above 1000 MW nuclear power plant. (authors)

  19. Longitudinal Metagenomic Analysis of Hospital Air Identifies Clinically Relevant Microbes

    Science.gov (United States)

    King, Paula; Pham, Long K.; Waltz, Shannon; Sphar, Dan; Yamamoto, Robert T.; Conrad, Douglas; Taplitz, Randy; Torriani, Francesca

    2016-01-01

    We describe the sampling of sixty-three uncultured hospital air samples collected over a six-month period and analysis using shotgun metagenomic sequencing. Our primary goals were to determine the longitudinal metagenomic variability of this environment, identify and characterize genomes of potential pathogens and determine whether they are atypical to the hospital airborne metagenome. Air samples were collected from eight locations which included patient wards, the main lobby and outside. The resulting DNA libraries produced 972 million sequences representing 51 gigabases. Hierarchical clustering of samples by the most abundant 50 microbial orders generated three major nodes which primarily clustered by type of location. Because the indoor locations were longitudinally consistent, episodic relative increases in microbial genomic signatures related to the opportunistic pathogens Aspergillus, Penicillium and Stenotrophomonas were identified as outliers at specific locations. Further analysis of microbial reads specific for Stenotrophomonas maltophilia indicated homology to a sequenced multi-drug resistant clinical strain and we observed broad sequence coverage of resistance genes. We demonstrate that a shotgun metagenomic sequencing approach can be used to characterize the resistance determinants of pathogen genomes that are uncharacteristic for an otherwise consistent hospital air microbial metagenomic profile. PMID:27482891

  20. Use of discriminant analysis to identify propensity for purchasing properties

    Directory of Open Access Journals (Sweden)

    Ricardo Floriani

    2015-03-01

    Full Text Available Properties usually represent a milestone for people and families due to the high added-value when compared with family income. The objective of this study is the proposition of a discrimination model, by a discriminant analysis of people with characteristics (according to independent variables classified as potential buyers of properties, as well as to identify the interest in the use of such property, if it will be assigned to housing or leisure activities such as a cottage or beach house, and/or for investment. Thus, the following research question is proposed: What are the characteristics that better describe the profile of people which intend to acquire properties? The study justifies itself by its economic relevance in the real estate industry, as well as to the players of the real estate Market that may develop products based on the profile of potential customers. As a statistical technique, discriminant analysis was applied to the data gathered by questionnaire, which was sent via e-mail. Three hundred and thirty four responses were gathered. Based on this study, it was observed that it is possible to identify the intention for acquired properties, as well the purpose for acquiring it, for housing or investments.

  1. Cluster analysis of clinical data identifies fibromyalgia subgroups.

    Directory of Open Access Journals (Sweden)

    Elisa Docampo

    Full Text Available INTRODUCTION: Fibromyalgia (FM is mainly characterized by widespread pain and multiple accompanying symptoms, which hinder FM assessment and management. In order to reduce FM heterogeneity we classified clinical data into simplified dimensions that were used to define FM subgroups. MATERIAL AND METHODS: 48 variables were evaluated in 1,446 Spanish FM cases fulfilling 1990 ACR FM criteria. A partitioning analysis was performed to find groups of variables similar to each other. Similarities between variables were identified and the variables were grouped into dimensions. This was performed in a subset of 559 patients, and cross-validated in the remaining 887 patients. For each sample and dimension, a composite index was obtained based on the weights of the variables included in the dimension. Finally, a clustering procedure was applied to the indexes, resulting in FM subgroups. RESULTS: VARIABLES CLUSTERED INTO THREE INDEPENDENT DIMENSIONS: "symptomatology", "comorbidities" and "clinical scales". Only the two first dimensions were considered for the construction of FM subgroups. Resulting scores classified FM samples into three subgroups: low symptomatology and comorbidities (Cluster 1, high symptomatology and comorbidities (Cluster 2, and high symptomatology but low comorbidities (Cluster 3, showing differences in measures of disease severity. CONCLUSIONS: We have identified three subgroups of FM samples in a large cohort of FM by clustering clinical data. Our analysis stresses the importance of family and personal history of FM comorbidities. Also, the resulting patient clusters could indicate different forms of the disease, relevant to future research, and might have an impact on clinical assessment.

  2. Safety analysis of JMTR LEU fuel core, (3)

    International Nuclear Information System (INIS)

    Dose analysis in the safety evaluation and the site evaluation were performed for the JMTR core conversion from MEU fuel to LEU fuel. In the safety evaluation, the effective dose equivalents for the public surrounding the site were estimated in fuel handling accident and flow blockage to coolant channel which were selected as the design basis accidents with release of radioactive fission products to the environment. In the site evaluation, the flow blockage to coolant channel was selected as siting basis events, since this accident had the possibility of spreading radioactive release. Maximum exposure doses for the public were estimated assuming large amounts of fission products to release. It was confirmed that risk of radiation exposure of the public is negligible and the siting is appropriate. (author)

  3. Design study on BN-600 hybrid core. 1. Evaluation of core neutronic and thermalhydraulic characteristics by Japanese analysis methods

    International Nuclear Information System (INIS)

    A program of disposition of Russian weapon-grade plutonium by containing the plutonium in vibropacked MOX fuel subassemblies and burning them in the BN-600 hybrid reactor core has been progressed. The relevant design works on the BN-600 hybrid core were carried out under the contract between Japan Nuclear Cycle Development Institute (JNC) and OKB Mechanical Engineering (OKBM), Russian public enterprise, and completed. JNC obtained a series of design technical reports. Japanese analysis methods were adopted to evaluate neutronic and thermal-hydraulic characteristics of the BN-600 hybrid core, based on the design technical data described in the obtained reports. The evaluation results of the key core performances, such as maximum linear heat rate, core-averaged discharge burnup, sodium void reactivity, capability of disposition of weapon-grade plutonium and, and reactivity control balance, were found to satisfy the design criteria and/or targets provided by Russia, and meet the Russian rule. The results of this study showed that the core and fuel specifications determined by Russia can be considered reasonable and proper in terms of neutronic and thermal-hydraulic design, and that the Japanese analysis methods are expected to contribute to increasing reliability of the Russian design works. (author)

  4. Spatial and model-order based reactor signal analysis methodology for BWR core stability evaluation

    International Nuclear Information System (INIS)

    A new methodology for the boiling water reactor core stability evaluation from measured noise signals has been recently developed and adopted at the Paul Scherrer Institut (PSI). This methodology consists in a general reactor noise analysis where as much as possible information recorded during the tests is investigated prior to determining core representative stability parameters, i.e. the decay ratio (DR) and the resonance frequency, along with an associated estimate of the uncertainty range. A central part in this approach is that the evaluation of the core stability parameters is performed not only for a few but for ALL recorded neutron flux signals, allowing thereby the assessment of signal-related uncertainties. In addition, for each signal, three different model-order optimization methods are systematically employed to take into account the sensitivity upon the model-order. The current methodology is then applied to the evaluation of the core stability measurements performed at the Leibstadt NPP, Switzerland, during cycles 10, 13 and 19. The results show that as the core becomes very stable, the method-related uncertainty becomes the major contributor to the overall uncertainty range while for intermediate DR values, the signal-related uncertainty becomes dominant. However, as the core stability deteriorates, the method-related and signal-related spreads have similar contributions to the overall uncertainty, and both are found to be small. The PSI methodology identifies the origin of the different contributions to the uncertainty. Furthermore, in order to assess the results obtained with the current methodology, a comparative study is for completeness carried out with respect to results from previously developed and applied procedures. The results show a good agreement between the current method and the other methods

  5. Simple Sequence Repeat Analysis of Genetic Diversity in Primary Core Collection of Peach (Prunus persica)

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    In this study, the genetic diversity of 51 cultivars in the primary core collection of peach (Prunus persica (L.) Batsch) was evaluated by using simple sequence repeats (SSRs). The phylogenetic relationships and the evolutionary history among different cultivars were determined on the basis of SSR data. Twenty-two polymorphic SSR primer pairs were selected, and a total of 111 alleles were identified in the 51 cultivars, with an average of 5 alleles per locus. According to traditional Chinese classification of peach cultivars, the 51 cultivars in the peach primary core collection belong to six variety groups. The SSR analysis revealed that the levels of the genetic diversity within each variety group were ranked as Sweet peach > Crisppeach > Flat peach > Nectarine > Honey Peach > Yellow fleshed peach. The genetic diversity among the Chinese cultivars was higher than that among the introduced cultivars. Cluster analysis by the unweighted pair group method with arithmetic averaging (UPGMA)placed the 51 cultivars into five linkage clusters. Cultivar members from the same variety group were distributed in different UPGMA clusters and some members from different variety groups were placed under the same cluster. Different variety groups could not be differentiated in accordance with SSR markers. The SSR analysis revealed rich genetic diversity in the peach primary core collection, representative of genetic resources of peach.

  6. A meta-analysis of core stability exercise versus general exercise for chronic low back pain.

    Directory of Open Access Journals (Sweden)

    Xue-Qiang Wang

    Full Text Available OBJECTIVE: To review the effects of core stability exercise or general exercise for patients with chronic low back pain (LBP. SUMMARY OF BACKGROUND DATA: Exercise therapy appears to be effective at decreasing pain and improving function for patients with chronic LBP in practice guidelines. Core stability exercise is becoming increasingly popular for LBP. However, it is currently unknown whether core stability exercise produces more beneficial effects than general exercise in patients with chronic LBP. METHODS: Published articles from 1970 to October 2011 were identified using electronic searches. For this meta-analysis, two reviewers independently selected relevant randomized controlled trials (RCTs investigating core stability exercise versus general exercise for the treatment of patients with chronic LBP. Data were extracted independently by the same two individuals who selected the studies. RESULTS: From the 28 potentially relevant trials, a total of 5 trials involving 414 participants were included in the current analysis. The pooling revealed that core stability exercise was better than general exercise for reducing pain [mean difference (-1.29; 95% confidence interval (-2.47, -0.11; P = 0.003] and disability [mean difference (-7.14; 95% confidence interval (-11.64, -2.65; P = 0.002] at the time of the short-term follow-up. However, no significant differences were observed between core stability exercise and general exercise in reducing pain at 6 months [mean difference (-0.50; 95% confidence interval (-1.36, 0.36; P = 0.26] and 12 months [mean difference (-0.32; 95% confidence interval (-0.87, 0.23; P = 0.25]. CONCLUSIONS: Compared to general exercise, core stability exercise is more effective in decreasing pain and may improve physical function in patients with chronic LBP in the short term. However, no significant long-term differences in pain severity were observed between patients who engaged in core stability exercise

  7. Availability analysis of the AP600 passive core cooling system

    International Nuclear Information System (INIS)

    The reliability analysis of the AP600 Passive Core Cooling System (PXS) has been done. The fault tree analysis method was used for the quantitative analysis. The PXS can be grouped to several sub-systems i.e.: Reactor Coolant System (RCS) Injection Subsystem, Emergency Core Decay Heat Removal Subsystem, and Containment Sump pH Control Subsystem. The quantitative analysis results indicates that the system unavailability is highly dependent on the valves configuration of the Automatic Depressurization System (ADS). If the ADS valves is arranged in Option-1, the system unavailability is 2.347E-03, this means that the yearly contribution to plant down time can be estimated to be about 20.56 hours per year. Whereas, by using Option-2 of fourth stage ADS valves, the system unavailability is reduced to be 9.877E-04 or 8.65 hours per year and this value is consistent with the allocated goal value (8.0 hours per year). The ADS contributes 66.89% to the system unavailability if it is arranged in Option-1, and will reduced to be about 21.21% if its fourth stages are arranged in Option-2. If the ADS is not included as a subsystem of the PXS (relocate to RCS as a subsystem of RCS), then the PXS unavailability will be reduced to about 7.784E-04 or 6.82 hours per year and this is less then the allocated goal value. The major contributors to the system unavailability are mostly dominated by Stage-4 ADS valves (air piston operated valves and squib valves), inservice testing valves of ADS (solenoid operated valves), solenoid valves of Nitrogen Supply to Accumulator, and Passive Residual Heat Removal actuation valves (air operated valves). Therefore, it is recommended that those valves be analyzed more detail to gain the improvement in its reliability. It is also recommended that the fourth stage of ADS valves should be arranged according to Option-2, i.e. one 10-inch normally open motor operated gate valve in series with one 10-inch normally closed squib valve. (author). 13 refs, 3

  8. Imaging and analysis of microcalcifications and lipid/necrotic core calcification in fibrous cap atheroma.

    Science.gov (United States)

    Maldonado, Natalia; Kelly-Arnold, Adreanne; Laudier, Damien; Weinbaum, Sheldon; Cardoso, Luis

    2015-06-01

    The presence of microcalcifications (µCalcs) >5 µm within the cap of human fibroatheroma has been shown to produce a 200-700% increase in peak circumferential stress, which can transform a stable plaque into a vulnerable one, whereas µCalcs 5 µm based on the gross morphological features of fibroatheromas, and the correlation between the size and distribution of µCalcs in the cap and the calcification in the lipid/necrotic core beneath it. Atherosclerotic lesions (N = 72) were imaged using HR-μCT at 2.1-μm resolution for detailed analysis of atheroma morphology and composition, and validated using non-decalcified histology. At 2.1-μm resolution one observes four different patterns of calcification within the lipid/necrotic core, and is able to elucidate the 3D spatial progression of the calcification process using these four patterns. Of the gross morphological features identified, only minimum cap thickness positively correlated with the existence of µCalcs > 5 µm in the cap. We also show that µCalcs in the cap accumulate in the vicinity of the lipid/necrotic core boundary with few on the lumen side of the cap. HR-μCT enables three-dimensional assessment of soft tissue composition, lipid content, calcification patterns within lipid/necrotic cores and analysis of the axial progression of calcification within individual atheroma. The distribution of µCalcs within the cap is highly non-uniform and decreases sharply as one proceeds from the lipid pool/necrotic core boundary to the lumen. PMID:25837377

  9. Uncertainly propagation analysis for Yonggwang nuclear unit 4 by McCARD/MASTER core analysis system

    International Nuclear Information System (INIS)

    This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor (keff), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

  10. Analysis of an Image Secret Sharing Scheme to Identify Cheaters

    Directory of Open Access Journals (Sweden)

    Jung-San LEe

    2010-09-01

    Full Text Available Secret image sharing mechanisms have been widely applied to the military, e-commerce, and communications fields. Zhao et al. introduced the concept of cheater detection into image sharing schemes recently. This functionality enables the image owner and authorized members to identify the cheater in reconstructing the secret image. Here, we provide an analysis of Zhao et al.¡¦s method: an authorized participant is able to restore the secret image by him/herself. This contradicts the requirement of secret image sharing schemes. The authorized participant utilizes an exhaustive search to achieve the attempt, though, simulation results show that it can be done within a reasonable time period.

  11. Social network analysis in identifying influential webloggers: A preliminary study

    Science.gov (United States)

    Hasmuni, Noraini; Sulaiman, Nor Intan Saniah; Zaibidi, Nerda Zura

    2014-12-01

    In recent years, second generation of internet-based services such as weblog has become an effective communication tool to publish information on the Web. Weblogs have unique characteristics that deserve users' attention. Some of webloggers have seen weblogs as appropriate medium to initiate and expand business. These webloggers or also known as direct profit-oriented webloggers (DPOWs) communicate and share knowledge with each other through social interaction. However, survivability is the main issue among DPOW. Frequent communication with influential webloggers is one of the way to keep survive as DPOW. This paper aims to understand the network structure and identify influential webloggers within the network. Proper understanding of the network structure can assist us in knowing how the information is exchanged among members and enhance survivability among DPOW. 30 DPOW were involved in this study. Degree centrality and betweenness centrality measurement in Social Network Analysis (SNA) were used to examine the strength relation and identify influential webloggers within the network. Thus, webloggers with the highest value of these measurements are considered as the most influential webloggers in the network.

  12. DESIGN, DEVELOPMENT AND NUMERICAL ANALYSIS OF HONEYCOMB CORE WITH VARIABLE CRUSHING STRENGTH

    OpenAIRE

    Shabnam Sadeghi Esfahlani; Hassan Shirvani; Ayoub Shirvani; Habtom Mebrahtu; Sunny Nwaubani

    2013-01-01

    A honeycomb core with half-circular cut-away sections at the spine (the adjoining cell walls) is designed and developed and numerically tested under axial dynamic load condition. The parametric study is invoked to identify the effect of various circular cut-away dimensions. In one embodiment a half-circular shaped cuts are removed from the top of the cell where the cell is impacted and its radius decreases toward the trailing edge of the cell. Numerical (FE) analysis was performed using expli...

  13. BN-600 full MOX core benchmark analysis (PHYSOR 2004 paper)

    International Nuclear Information System (INIS)

    As a follow-up of the BN-600 hybrid core benchmark, a full MOX core benchmark was performed within the framework of the IAEA co-ordinated research project Discrepancies between the values of main reactivity coefficients obtained by the participants for the BN-600 full MOX core benchmark appear to be larger than those in the previous hybrid core benchmarks on traditional core configurations. This arises due to uncertainties in the proper modelling of the axial sodium plenum above the core. It was recognized that the sodium density coefficient strongly depends on the core model configuration of interest (hybrid core vs. fully MOX fuelled core with sodium plenum above the core) in conjunction with the calculation method (diffusion vs. transport theory). The effects of the discrepancies revealed between the participants' results on the ULOF and UTOP transient behaviours of the BN-600 full MOX core were investigated in simplified transient analyses. Generally the diffusion approximation predicts more benign consequences for the ULOF accident but more hazardous ones for the UTOP accident when compared with the transport theory results. The heterogeneity effect does not have any significant effect on the simulation of the transient The comparison of the transient analyses results concluded that the fuel Doppler coefficient and the sodium density coefficient are the two most important coefficients in understanding the ULOF transient behaviour. In particular, the uncertainty in evaluating the sodium density coefficient distribution has the largest impact on the description of reactor dynamics. This is because the maximum sodium temperature rise takes place at the top of the core and in the sodium plenum

  14. Overview of core simulation methodologies for light water reactor analysis

    International Nuclear Information System (INIS)

    The current in-core fuel management calculation methods provide a very efficient route to predict neutronics behavior of light water reactor (LWR) cores and their prediction accuracy for current generation LWRs is generally sufficient. However, since neutronics calculations for LWRs are based on various assumptions and simplifications, we should also recognize many implicit limitations that are 'embedded' in current neutronics calculation methodologies. Continuous effort for improvement of core simulation methodologies is also discussed. (author)

  15. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

  16. Self-Healing Many-Core Architecture: Analysis and Evaluation

    Directory of Open Access Journals (Sweden)

    Arezoo Kamran

    2016-01-01

    Full Text Available More pronounced aging effects, more frequent early-life failures, and incomplete testing and verification processes due to time-to-market pressure in new fabrication technologies impose reliability challenges on forthcoming systems. A promising solution to these reliability challenges is self-test and self-reconfiguration with no or limited external control. In this work a scalable self-test mechanism for periodic online testing of many-core processor has been proposed. This test mechanism facilitates autonomous detection and omission of faulty cores and makes graceful degradation of the many-core architecture possible. Several test components are incorporated in the many-core architecture that distribute test stimuli, suspend normal operation of individual processing cores, apply test, and detect faulty cores. Test is performed concurrently with the system normal operation without any noticeable downtime at the application level. Experimental results show that the proposed test architecture is extensively scalable in terms of hardware overhead and performance overhead that makes it applicable to many-cores with more than a thousand processing cores.

  17. Directional reflectance analysis for identifying counterfeit drugs: Preliminary study.

    Science.gov (United States)

    Wilczyński, Sławomir; Koprowski, Robert; Błońska-Fajfrowska, Barbara

    2016-05-30

    The WHO estimates that up to 10% of drugs on the market may be counterfeit. In order to prevent intensification of the phenomenon of drug counterfeiting, the methods for distinguishing genuine medicines from fake ones need to be developed. The aim of this study was to try to develop simple, reproducible and inexpensive method for distinguishing between original and counterfeit medicines based on the measurement of directional reflectance. The directional reflectance of 6 original Viagra(®) tablets (Pfizer) and 24 (4 different batches) counterfeit tablets (imitating Viagra(®)) was examined in six spectral bands: from 0.9 to 1.1 μm, from 1.9 to 2.6 μm, from 3.0 to 4.0 μm, from 3.0 to 5.0 μm, from 4.0 to 5.0 μm, from 8.0 to 12.0 μm, and for two angles of incidence, 20° and 60°. Directional hemispherical reflectometer was applied to measure directional reflectance. Significant statistical differences between the directional reflectance of the original Viagra(®) and counterfeit tablets were registered. Any difference in the value of directional reflectance for any spectral band or angle of incidence identifies the drug as a fake one. The proposed method of directional reflectance analysis enables to differentiate between the real Viagra(®) and fake tablets. Directional reflectance analysis is a fast (measurement time under 5s), cheap and reproducible method which does not require expensive equipment or specialized laboratory staff. It also seems to be an effective method, however, the effectiveness will be assessed after the extension of research. PMID:26977587

  18. Real time analysis of light water core neutronics

    International Nuclear Information System (INIS)

    A method is described for determining the neutronics parameters of a reactor core comprising the steps of: representing the reactor core as a plurality of nodes having a coarse nodal representation; monitoring selected neutronic parameters of the reactor core; providing time-dependent two group neutron diffusion equations coupled to delayed neutron precursor concentrations that have been subjected to space-time factorization by shape and amplitude functions in response to the plurality of nodes; sensing the monitored parameters; and determining the core neutronics parameters in response to the sensed parameters and the provided two group neutron diffusion equations in constant time steps for sensing the monitored parameters and determining the core neutronics parameters in a real-time environment, the time steps being not less than one quarter second

  19. Unsteady thermal analysis of gas-cooled fast reactor core

    International Nuclear Information System (INIS)

    This thesis presents numerical analysis of transient heat transfer in an equivalent coolant-fuel rod cell of a typical gas cooled, fast nuclear reactor core. The transient performance is assumed to follow a complete sudden loss of coolant starting from steady state operation. Steady state conditions are obtained from solving a conduction problem in the fuel rod and a parabolic turbutent convection problem in the coolant section. The coupling between the two problems is accomplished by ensuring continuity of the thermal conditions at the interface between the fuel rod and the coolant. to model turbulence, the mixing tenght theory is used. Various fuel rod configurations have been tested for optimal transient performance. Actually, the loss of coolant accident occurs gradually at an exponential rate. Moreover, a time delay before shutting down the reactor by insertion of control rods usually exists. It is required to minimize maximum steady state cladding temperature so that the time required to reach its limiting value during transient state is maximum. This will prevent the escape of radioactive gases that endanger the environment and the public. However, the case considered here is a limiting case representing what could actually happen in the worst probable accident. So, the resutls in this thesis are very indicative regarding selection of the fuel rode configuration for better transient performance in case of accidents in which complete loss of collant occurs instantaneously

  20. Homogeneous protein analysis by magnetic core-shell nanorod probes

    KAUST Repository

    Schrittwieser, Stefan

    2016-03-29

    Studying protein interactions is of vital importance both to fundamental biology research and to medical applications. Here, we report on the experimental proof of a universally applicable label-free homogeneous platform for rapid protein analysis. It is based on optically detecting changes in the rotational dynamics of magnetically agitated core-shell nanorods upon their specific interaction with proteins. By adjusting the excitation frequency, we are able to optimize the measurement signal for each analyte protein size. In addition, due to the locking of the optical signal to the magnetic excitation frequency, background signals are suppressed, thus allowing exclusive studies of processes at the nanoprobe surface only. We study target proteins (soluble domain of the human epidermal growth factor receptor 2 - sHER2) specifically binding to antibodies (trastuzumab) immobilized on the surface of our nanoprobes and demonstrate direct deduction of their respective sizes. Additionally, we examine the dependence of our measurement signal on the concentration of the analyte protein, and deduce a minimally detectable sHER2 concentration of 440 pM. For our homogeneous measurement platform, good dispersion stability of the applied nanoprobes under physiological conditions is of vital importance. To that end, we support our measurement data by theoretical modeling of the total particle-particle interaction energies. The successful implementation of our platform offers scope for applications in biomarker-based diagnostics as well as for answering basic biology questions.

  1. Global secretome analysis identifies novel mediators of bone metastasis

    Institute of Scientific and Technical Information of China (English)

    Mario Andres Blanco; Gary LeRoy; Zia Khan; Ma(s)a Ale(c)kovi(c); Barry M Zee; Benjamin A Garcia; Yibin Kang

    2012-01-01

    Bone is the one of the most common sites of distant metastasis of solid tumors.Secreted proteins are known to influence pathological interactions between metastatic cancer cells and the bone stroma.To comprehensively profile secreted proteins associated with bone metastasis,we used quantitative and non-quantitative mass spectrometry to globally analyze the secretomes of nine cell lines of varying bone metastatic ability from multiple species and cancer types.By comparing the secretomes of parental cells and their bone metastatic derivatives,we identified the secreted proteins that were uniquely associated with bone metastasis in these cell lines.We then incorporated bioinformatic analyses of large clinical metastasis datasets to obtain a list of candidate novel bone metastasis proteins of several functional classes that were strongly associated with both clinical and experimental bone metastasis.Functional validation of selected proteins indicated that in vivo bone metastasis can be promoted by high expression of (1) the salivary cystatins CST1,CST2,and CST4; (2) the plasminogen activators PLAT and PLAU; or (3) the collagen functionality proteins PLOD2 and COL6A1.Overall,our study has uncovered several new secreted mediators of bone metastasis and therefore demonstrated that secretome analysis is a powerful method for identification of novel biomarkers and candidate therapeutic targets.

  2. A Sensitivity Analysis Approach to Identify Key Environmental Performance Factors

    Directory of Open Access Journals (Sweden)

    Xi Yu

    2014-01-01

    Full Text Available Life cycle assessment (LCA is widely used in design phase to reduce the product’s environmental impacts through the whole product life cycle (PLC during the last two decades. The traditional LCA is restricted to assessing the environmental impacts of a product and the results cannot reflect the effects of changes within the life cycle. In order to improve the quality of ecodesign, it is a growing need to develop an approach which can reflect the changes between the design parameters and product’s environmental impacts. A sensitivity analysis approach based on LCA and ecodesign is proposed in this paper. The key environmental performance factors which have significant influence on the products’ environmental impacts can be identified by analyzing the relationship between environmental impacts and the design parameters. Users without much environmental knowledge can use this approach to determine which design parameter should be first considered when (redesigning a product. A printed circuit board (PCB case study is conducted; eight design parameters are chosen to be analyzed by our approach. The result shows that the carbon dioxide emission during the PCB manufacture is highly sensitive to the area of PCB panel.

  3. Global secretome analysis identifies novel mediators of bone metastasis.

    Science.gov (United States)

    Blanco, Mario Andres; LeRoy, Gary; Khan, Zia; Alečković, Maša; Zee, Barry M; Garcia, Benjamin A; Kang, Yibin

    2012-09-01

    Bone is the one of the most common sites of distant metastasis of solid tumors. Secreted proteins are known to influence pathological interactions between metastatic cancer cells and the bone stroma. To comprehensively profile secreted proteins associated with bone metastasis, we used quantitative and non-quantitative mass spectrometry to globally analyze the secretomes of nine cell lines of varying bone metastatic ability from multiple species and cancer types. By comparing the secretomes of parental cells and their bone metastatic derivatives, we identified the secreted proteins that were uniquely associated with bone metastasis in these cell lines. We then incorporated bioinformatic analyses of large clinical metastasis datasets to obtain a list of candidate novel bone metastasis proteins of several functional classes that were strongly associated with both clinical and experimental bone metastasis. Functional validation of selected proteins indicated that in vivo bone metastasis can be promoted by high expression of (1) the salivary cystatins CST1, CST2, and CST4; (2) the plasminogen activators PLAT and PLAU; or (3) the collagen functionality proteins PLOD2 and COL6A1. Overall, our study has uncovered several new secreted mediators of bone metastasis and therefore demonstrated that secretome analysis is a powerful method for identification of novel biomarkers and candidate therapeutic targets. PMID:22688892

  4. Identifying redundancy and exposing provenance in crowdsourced data analysis.

    Science.gov (United States)

    Willett, Wesley; Ginosar, Shiry; Steinitz, Avital; Hartmann, Björn; Agrawala, Maneesh

    2013-12-01

    We present a system that lets analysts use paid crowd workers to explore data sets and helps analysts interactively examine and build upon workers' insights. We take advantage of the fact that, for many types of data, independent crowd workers can readily perform basic analysis tasks like examining views and generating explanations for trends and patterns. However, workers operating in parallel can often generate redundant explanations. Moreover, because workers have different competencies and domain knowledge, some responses are likely to be more plausible than others. To efficiently utilize the crowd's work, analysts must be able to quickly identify and consolidate redundant responses and determine which explanations are the most plausible. In this paper, we demonstrate several crowd-assisted techniques to help analysts make better use of crowdsourced explanations: (1) We explore crowd-assisted strategies that utilize multiple workers to detect redundant explanations. We introduce color clustering with representative selection--a strategy in which multiple workers cluster explanations and we automatically select the most-representative result--and show that it generates clusterings that are as good as those produced by experts. (2) We capture explanation provenance by introducing highlighting tasks and capturing workers' browsing behavior via an embedded web browser, and refine that provenance information via source-review tasks. We expose this information in an explanation-management interface that allows analysts to interactively filter and sort responses, select the most plausible explanations, and decide which to explore further. PMID:24051786

  5. Analysis of core damage frequency: Surry Power Station, Unit 1 external events

    International Nuclear Information System (INIS)

    This report presents the analysis of external events (earthquakes, fires, floods, etc.) performed for the Surry Power Station as part of the USNRC-sponsored NUREG-1150 program. Both the internal and external events analyses make full use of recent insights and developments in risk assessment methods. In addition, the external event analyses make use of newly-developed simplified methods. As a first step, a screening analysis was performed which showed that all external events were negligible except for fires and seismic events. Subsequent detailed analysis of fires resulted in a total (mean) core damage frequency of 1.13E-5 per year. The seismic analysis resulted in a total (mean) core damage frequency of 1.16E-4 per year using hazard curves developed by Lawrence Livermore National Laboratory and 2.50E-5 per year using hazard curves developed by the Electric Power Research Institute. Uncertainty analyses were performed, and dominant components and sources of uncertainty were identified. 71 refs., 61 figs., 59 tabs

  6. Aspects of cell calculations in deterministic reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Varvayanni, M. [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Savva, P., E-mail: savvapan@ipta.demokritos.gr [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N. [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    {Tau}he capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available

  7. Simulation of Core Support Barrel Vibration Monitoring Using Ex-Core Neutron Noise Analysis and Fuzzy Logic Algorithm

    International Nuclear Information System (INIS)

    The application of ex-core Neutron Noise Analysis (NNA) to monitor the vibration characteristics of a reactor Core Support Barrel (CSB) was studied. Ex-core flux data was obtained using a non-analog Monte Carlo neutron transport method in a simulated CSB model. The implicit capture and Russian Roulette technique was optimized through a sensitivity study to simulate the neutron transport. A combination of two-dimensional and three-dimensional beam and shell mode vibration of CSB was modelled. Parallel processing was employed to reduce the simulation time. An NNA module was developed to analyze the ex-core flux data based on its time variation, Normalized Power Spectral Density (NPSD), Normalized Cross-Power Spectral Density (NCPSD), Coherence and phase differences. The data was then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core flux signal fluctuation was directly proportional to the CSB's vibration observed at 8 and 15 Hz in the beam mode vibration, and at 8 Hz in the shell mode vibration. The Coherence result between flux pairs was unity at the vibration peak frequencies. A set of out-of-phase and in-phase unique pattern of phase differences was observed for each of the vibration models. The fuzzy logic module successfully recognized the correct vibration frequencies, modes, orders, directions, and phase differences within 4.1 milliseconds for the three-dimensional beam and shell mode vibrations

  8. Simulation of Core Support Barrel Vibration Monitoring Using Ex-Core Neutron Noise Analysis and Fuzzy Logic Algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Christian, Robby; Song, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Seon, Ho [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-08-15

    The application of ex-core Neutron Noise Analysis (NNA) to monitor the vibration characteristics of a reactor Core Support Barrel (CSB) was studied. Ex-core flux data was obtained using a non-analog Monte Carlo neutron transport method in a simulated CSB model. The implicit capture and Russian Roulette technique was optimized through a sensitivity study to simulate the neutron transport. A combination of two-dimensional and three-dimensional beam and shell mode vibration of CSB was modelled. Parallel processing was employed to reduce the simulation time. An NNA module was developed to analyze the ex-core flux data based on its time variation, Normalized Power Spectral Density (NPSD), Normalized Cross-Power Spectral Density (NCPSD), Coherence and phase differences. The data was then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core flux signal fluctuation was directly proportional to the CSB's vibration observed at 8 and 15 Hz in the beam mode vibration, and at 8 Hz in the shell mode vibration. The Coherence result between flux pairs was unity at the vibration peak frequencies. A set of out-of-phase and in-phase unique pattern of phase differences was observed for each of the vibration models. The fuzzy logic module successfully recognized the correct vibration frequencies, modes, orders, directions, and phase differences within 4.1 milliseconds for the three-dimensional beam and shell mode vibrations.

  9. Analysis of the seismic response of a fast reactor core

    International Nuclear Information System (INIS)

    This report deals with the methods to apply for a correct evaluation of the reactor core seismic response. Reference is made to up-to-date design data concerning the PEC core, taking into account the presence of the core-restraint plate located close to the PEC core elements top and applying the optimized iterative procedure between the vessel linear calculation and the non-linear ones limited to the core, which had been described in a previous report. It is demonstrated that the convergence of this procedure is very fast, similar to what obtained in the calculations of the cited report, carried out with preliminary data, and it is shown that the cited methods allow a reliable evaluation of the excitation time histories for the experimental tests in support of the seismic verification of the shutdown system and the core of a fast reactor, as well as relevant data for the experimental, structural and functional, verification of the core elements in the case of seismic loads

  10. HEU-LEU mixed core analysis for TR-2

    International Nuclear Information System (INIS)

    Core conversion calculation have been carried out for different core loadings of the TR-2 reactor in order to find out the optimum design for radioisotope production. Using HEU and LEU fuel elements in the mixed core also introduced additional peaking problems to be eliminated. Five group structure is used for the burnup dependent cross-section libraries that are generated by EPRI-CELL code. 20 diffusion-depletion code GEREBUS is used for the reactivity and burnup calculations. New graphite reflectors have been added to the periphery of the core to enhance the reactivity and the discharge burnup levels. Two water boxes have been placed inside reactor core in order to increase the radioisotope production. The activity levels of the irradiation samples, core excess reactivities, power peaking factors, and the anti-reactivities of the control blades have been calculated for various loadings. After the optimization studies, it is found that these modifications have been yielded higher production rates and an uniform distribution in the activity levels of the irradiation samples. One irradiation and two standard LEU fuel elements have already been loaded to the TR-2 core without any operational or safety related problems.The agreement between the calculation and the experiments are quite good for the operated 13 cycles

  11. European ERANOS formulaire for fast reactor core analysis

    International Nuclear Information System (INIS)

    ERANOS code scheme was developed within the European collaboration on fast reactors. It contains all the functions required to calculate a complete set of core, shielding and fuel cycle parameters for LMFR cores. Nuclear data are taken from recent evaluations (JEF2.2) and adjusted on integral experiments (ERALIB1). Calculational scheme uses the ECCO cell code to generate cross section data. Whole core calculations are carried out using the spatial modules BISTRO (Sn) and TGVNARIANT (nodal method). Validation is based on integral and power reactor experiments. Integral experiments are also used for adjustment of nuclear data

  12. Methods adopted to account for vessel-core seismic interactions in the PEC analysis

    International Nuclear Information System (INIS)

    In the PEC reactor a restraint was provided around the core in order to control the top-core displacements under seismic conditions. Clearances between adjacent elements and between the external elements and the core-restraint ring introduce some amount of non-linearity in the seismic motion of the core. On the other hand a linear analysis shows that the core strongly affects the dynamical behaviour of the vessel. Therefore a method was developed by ENEA and Ansaldo to take into account the effects of non-linearities in the decoupled detailed analysis of the core and reactor block. Studies were also performed in order to account for the possibility of influences of fluid coupling on the core vessel interaction. This paper briefly describes the adopted methods and provides an overview of the main results. (author). 2 refs, 6 figs, 1 tab

  13. Analysis of dismantling possibility and unloading efforts of fuel assemblies from core of WWER

    International Nuclear Information System (INIS)

    The computation methods of optimal dismantling sequence of fuel assemblies (FA) from core of WWER after different operating periods and accident conditions are considered. The algorithms of fuel dismantling sequence are constructed both on the basis of analysis of mutual spacer grid overlaps of adjacent fuel assemblies and numerical structure analysis of efforts required for FA removal as FA heaving from the core. Computation results for core dismantling sequence after 3-year operating period and LB LOCA are presented in the paper

  14. Tank 241-SY-101 push mode core sampling and analysis plan

    International Nuclear Information System (INIS)

    This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for push mode core samples from tank 241-SY-101 (SY-101). It is written in accordance with Data Quality Objective to Support Resolution of the Flammable Gas Safety Issue (Bauer 1998), Low Activity Waste Feed Data Quality Objectives (Wiemers and Miller 1997 and DOE 1998), Data Quality Objectives for TWRS Privatization Phase I: Confirm Tank T is an Appropriate Feed Source for Low-Activity Waste Feed Batch X (Certa 1998), and the Tank Safety Screening Data Quality Objective (Dukelow et al. 1995). The Tank Characterization Technical Sampling Basis document (Brown et al. 1998) indicates that these issues apply to tank SY-101 for this sampling event. Brown et al. also identifies high-level waste, regulatory, pretreatment and disposal issues as applicable issues for this tank. However, these issues will not be addressed via this sampling event

  15. Design study on BN-600 hybrid core. 2. Evaluation of fuel integrity and core neutronic characteristics by Japanese analysis methods

    International Nuclear Information System (INIS)

    A program of disposal of Russian surplus weapon-grade plutonium by containing the plutonium in vibropacked MOX fuel subassemblies and burning them in the BN-600 hybrid reactor core has been progressed. The relevant design works on the BN-600 hybrid core have been carried out under the contract between Japan Nuclear Cycle Development Institute (JNC) and OKB Mechanical Engineering (OKBM), Russian public enterprise. JNC obtained a series of design technical reports. Japanese analysis methods were adopted to evaluate fuel integrity in the design basis transients and neutronic characteristics of the BN-600 hybrid core, based on the design technical data described in the obtained reports. The evaluation results of the key performances, such as maximum cladding and fuel temperatures, coolant (sodium) void reactivity, reactivity coefficient, were found to satisfy the design criteria and/or target provided by Russia, and meet the Russian rule. The results of this study showed that the core and fuel specifications determined by Russia can be considered reasonable and proper from the viewpoint of safety and neutronic designs, and that the Japanese analysis methods are expected to contribute to increasing reliability of the Russian design works. (author)

  16. Transport criticality analysis for FBR MONJU initial critical core in whole core simulation by NSHEX and GMVP

    International Nuclear Information System (INIS)

    FBR MONJU Initial Critical Core (ICC) criticality problem has been solved by deterministic and Monte Carlo transport methods by the codes NSHEX and GMVP. The analysis has been carried out in different energy-groups approximations. As a result the effect of cross-section (XS) condensation from 70 into few energy-group structures by different collapsing methods has been evaluated. The 3D discrete-ordinate code NSHEX has been applied for wide range of core simulations-from whole core, considering the fissile, fertile and shielding regions to simplified models that simulate an increased neutron leakage. It has been found that there is room for improvement in the assessment of the neutron leakage in the few energy-group approximations. The good agreement between NSHEX and GMVP results, especially without XS collapsing, is pointed out as a conformation for the applicability of the code NSHEX in FBR 3D whole core calculations. Some practical conclusions have been extracted that are important for the implementation of the code NSHEX in the standard criticality analysis. (author)

  17. Identifying Phytoplankton Classes In California Reservoirs Using HPLC Pigment Analysis

    Science.gov (United States)

    Siddiqui, S.; Peacock, M. B.; Kudela, R. M.; Negrey, K.

    2014-12-01

    Few bodies of water are routinely monitored for phytoplankton composition due to monetary and time constraints, especially the less accessible bodies of water in central and southern California. These lakes and estuaries are important for economic reasons such as tourism and fishing. This project investigated the composition of phytoplankton present using pigment analysis to identify dominant phytoplankton groups. A total of 28 different sites with a wide range of salinity (0 - 60) in central and southern California were examined. These included 13 different bodies of water in central California: 6 in the Sierras, 7 in the San Francisco Bay Estuary, and 15 from southern California. The samples were analyzed using high-performance liquid-chromatography (HPLC) to quantify the pigments present (using retention time and the spectral thumbprint). Diagnostic pigments were used to indicate the phytoplankton class composition, focusing on diatoms, dinoflagellates, cryptophytes, and cyanobacteria - all key phytoplankton groups indicative of the health of the sampled reservoir. Our results indicated that cyanobacteria dominated four of the seven bodies of central California water (Mono Lake, Bridgeport Reservoir, Steamboat Slough, and Pinto Lake); cryptophytes and nannoflagellates dominated two of the central California bodies of water (Mare Island Strait and Topaz Lake); and diatoms and dinoflagellates dominated one central California body of water, Oakland Inner Harbor, comprising more than 70% of the phytoplankton present. We expect the bodies of water from Southern California to be as disparate. Though this data is only a snapshot, it has significant implications in comparing different ecosystems across California, and it has the potential to provide valuable insight into the composition of phytoplankton communities.

  18. Plastic-dynamic analysis on shock absorber of reactor core

    International Nuclear Information System (INIS)

    The plastic-dynamic process under the condition of impact is studied for the shock absorbing device. The safety of the reactor core and vessel can be ensured by reasonably selecting the dimensions to lessen the dynamic loading factor

  19. Definition of a core module for the nuclear retrograde response to altered organellar gene expression identifies GLK overexpressors as gun mutants.

    Science.gov (United States)

    Leister, Dario; Kleine, Tatjana

    2016-07-01

    Retrograde signaling can be triggered by changes in organellar gene expression (OGE) induced by inhibitors such as lincomycin (LIN) or mutations that perturb OGE. Thus, an insufficiency of the organelle-targeted prolyl-tRNA synthetase PRORS1 in Arabidopsis thaliana activates retrograde signaling and reduces the expression of nuclear genes for photosynthetic proteins. Recently, we showed that mTERF6, a member of the so-called mitochondrial transcription termination factor (mTERF) family, is involved in the formation of chloroplast (cp) isoleucine-tRNA. To obtain further insights into its functions, co-expression analysis of MTERF6, PRORS1 and two other genes for organellar aminoacyl-tRNA synthetases was conducted. The results suggest a prominent role of mTERF6 in aminoacylation activity, light signaling and seed storage. Analysis of changes in whole-genome transcriptomes in the mterf6-1 mutant showed that levels of nuclear transcripts for cp OGE proteins were particularly affected. Comparison of the mterf6-1 transcriptome with that of prors1-2 showed that reduced aminoacylation of proline (prors1-2) and isoleucine (mterf6-1) tRNAs alters retrograde signaling in similar ways. Database analyses indicate that comparable gene expression changes are provoked by treatment with LIN, norflurazon or high light. A core OGE response module was defined by identifying genes that were differentially expressed under at least four of six conditions relevant to OGE signaling. Based on this module, overexpressors of the Golden2-like transcription factors GLK1 and GLK2 were identified as genomes uncoupled mutants. PMID:26876646

  20. TREAT Transient Analysis Benchmarking for the HEU Core

    International Nuclear Information System (INIS)

    This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used to determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term reported values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core's performance.

  1. Knowledge Economy Core Journals: Identification through LISTA Database Analysis

    OpenAIRE

    Nouri, Rasool; Karimi, Saeed; Ashrafi-rizi, Hassan; Nouri, Azadeh

    2013-01-01

    Background Knowledge economy has become increasingly broad over the years and identification of core journals in this field can be useful for librarians in journal selection process and also for researchers to select their studies and finding Appropriate Journal for publishing their articles. Present research attempts to determine core journals of Knowledge Economy indexed in LISTA (Library and Information Science and Technology). Methods The research method was bibliometric and research popu...

  2. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries

  3. Performance analysis of the 840 MWt PRISM reference burner core

    International Nuclear Information System (INIS)

    The General Electric PRISM (Power Reactor, Innovative Small Module) is a modular, pool-type sodium-cooled fast reactor employing innovative, passive features to provide an extremely high level of public safety. A PRISM power block consists of two 840 MWt reactor modules, each with a vessel diameter of 9.15 m (30 ft), tied to a turbine generator and producing 622 MWe. A full-size plant consists of three power blocks producing 1866 MWe of electrical power. Two core configurations have been analyzed. The reference is a 'burner' core (conversion ratio of 0.8) and the alternative is a breakeven' core (plutonium consumption balanced by plutonium generation). The core nuclear designs are largely governed by passive safety and reactivity control issues. The key features employed to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters (rod stops) and gas expansion modules (GEMs). A passive reactor vessel auxiliary cooling system (RVACS) assures safety-grade decay heat removal. This paper summarizes the operational and safety performance of the 840 MWt PRISM modular reactor, with emphasis on the reference burner core. (author)

  4. Single assembly preliminary analysis for horizontal seismic analysis on fast breeder reactor core

    International Nuclear Information System (INIS)

    Seismic analysis is one of important parts of fast breeder reactor (FBR) core design. It is necessary for structural integrity assessment and analysis of variation of reactivity under the earthquake. Moreover some important data for qualification of the scram capability of the control rods during the earthquake. Moreover some important data for qualification of the scram capability of the control rods during the earthquake could be provided. In the paper, FINAS, one finite element code developed by Japanese engineers, was used. The calculation model and method were studied on single assembly in China Experimental Fast Reactor (CEFR), as an example. Some preliminary analyses were carried out, which prepare for the seismic analysis on multiple assemblies in FBR core. (authors)

  5. TREAT Transient Analysis Benchmarking for the HEU Core

    Energy Technology Data Exchange (ETDEWEB)

    Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-05-01

    This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used to determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term “reported” values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core’s performance.

  6. Integrating subpathway analysis to identify candidate agents for hepatocellular carcinoma.

    Science.gov (United States)

    Wang, Jiye; Li, Mi; Wang, Yun; Liu, Xiaoping

    2016-01-01

    Hepatocellular carcinoma (HCC) is the second most common cause of cancer-associated death worldwide, characterized by a high invasiveness and resistance to normal anticancer treatments. The need to develop new therapeutic agents for HCC is urgent. Here, we developed a bioinformatics method to identify potential novel drugs for HCC by integrating HCC-related and drug-affected subpathways. By using the RNA-seq data from the TCGA (The Cancer Genome Atlas) database, we first identified 1,763 differentially expressed genes between HCC and normal samples. Next, we identified 104 significant HCC-related subpathways. We also identified the subpathways associated with small molecular drugs in the CMap database. Finally, by integrating HCC-related and drug-affected subpathways, we identified 40 novel small molecular drugs capable of targeting these HCC-involved subpathways. In addition to previously reported agents (ie, calmidazolium), our method also identified potentially novel agents for targeting HCC. We experimentally verified that one of these novel agents, prenylamine, induced HCC cell apoptosis using 3-(4,5-dimethylthiazol-2-yl)-2,5-diphenyltetrazolium bromide, an acridine orange/ethidium bromide stain, and electron microscopy. In addition, we found that prenylamine not only affected several classic apoptosis-related proteins, including Bax, Bcl-2, and cytochrome c, but also increased caspase-3 activity. These candidate small molecular drugs identified by us may provide insights into novel therapeutic approaches for HCC. PMID:27022281

  7. Structural Studies of Lipopolysaccharide-defective Mutants from Brucella melitensis Identify a Core Oligosaccharide Critical in Virulence.

    Science.gov (United States)

    Fontana, Carolina; Conde-Álvarez, Raquel; Ståhle, Jonas; Holst, Otto; Iriarte, Maite; Zhao, Yun; Arce-Gorvel, Vilma; Hanniffy, Seán; Gorvel, Jean-Pierre; Moriyón, Ignacio; Widmalm, Göran

    2016-04-01

    The structures of the lipooligosaccharides fromBrucella melitensismutants affected in the WbkD and ManBcoreproteins have been fully characterized using NMR spectroscopy. The results revealed that disruption ofwbkDgives rise to a rough lipopolysaccharide (R-LPS) with a complete core structure (β-d-Glcp-(1→4)-α-Kdop-(2→4)[β-d-GlcpN-(1→6)-β-d-GlcpN-(1→4)[β-d-GlcpN-(1→6)]-β-d-GlcpN-(1→3)-α-d-Manp-(1→5)]-α-Kdop-(2→6)-β-d-GlcpN3N4P-(1→6)-α-d-GlcpN3N1P), in addition to components lacking one of the terminal β-d-GlcpN and/or the β-d-Glcpresidues (48 and 17%, respectively). These structures were identical to those of the R-LPS fromB. melitensisEP, a strain simultaneously expressing both smooth and R-LPS, also studied herein. In contrast, disruption ofmanBcoregives rise to a deep-rough pentasaccharide core (β-d-Glcp-(1→4)-α-Kdop-(2→4)-α-Kdop-(2→6)-β-d-GlcpN3N4P-(1→6)-α-d-GlcpN3N1P) as the major component (63%), as well as a minor tetrasaccharide component lacking the terminal β-d-Glcpresidue (37%). These results are in agreement with the predicted functions of the WbkD (glycosyltransferase involved in the biosynthesis of the O-antigen) and ManBcoreproteins (phosphomannomutase involved in the biosynthesis of a mannosyl precursor needed for the biosynthesis of the core and O-antigen). We also report that deletion ofB. melitensis wadCremoves the core oligosaccharide branch not linked to the O-antigen causing an increase in overall negative charge of the remaining LPS inner section. This is in agreement with the mannosyltransferase role predicted for WadC and the lack of GlcpN residues in the defective core oligosaccharide. Despite carrying the O-antigen essential inB. melitensisvirulence, the core deficiency in thewadCmutant structure resulted in a more efficient detection by innate immunity and attenuation, proving the role of the β-d-GlcpN-(1→6)-β-d-GlcpN-(1→4)[β-d-GlcpN-(1→6)]-β-d-GlcpN-(1→3)-α-d-Manp-(1→5) structure

  8. Eulerian two-phase computational fluid dynamics for boiling water reactor core analysis

    International Nuclear Information System (INIS)

    Traditionally, the analysis of two-phase boiling flows has relied on experimentally-derived correlations. This approach provides accurate predictions of channel-averaged temperatures and void fractions and even peak assembly temperatures within an assembly. However, it lacks the resolution needed to predict the detailed intra-channel distributions of temperature, void fraction and steaming rates that are needed to address the fuel reliability concerns which result from longer refueling cycles and higher burnup fuels, particularly for the prediction of potential fuel pin cladding failures resulting from growth of tenacious crud. As part of the ongoing effort to develop a high-fidelity, full-core, pin-by-pin, fully-coupled neutronic and thermal hydraulic simulation package for reactor core analysis], capabilities for Eulerian-Eulerian two-phase simulation within the commercial Computational Fluid Dynamics code Star-CD are being extended and validated for application to Boiling Water Reactor (BWR) cores. The extension of the existing capability includes the development of wall heat partitioning and bubble growth models, implementation of a topology map based approach that provides the necessary capability to switch between the liquid and vapor as the continuous phase on a cell-by-cell basis and the development of appropriate models for the inter-phase forces that influence the movement of bubbles and droplets. Two applications have been identified as an initial demonstration and validation of the implemented methodology. First, the model is being applied to an Atrium-10 fuel assembly from Cycle 11 of the River Bend Nuclear Power Plant. Second, the model is being applied to an international benchmark problem for validation of BWR assembly analysis methods. (authors)

  9. The reactor core analysis code CITATION-1000VP for High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Reactor core analysis with full core model has been necessary for the High Temperature Engineering Test Reactor (HTTR) design. The CITATION-1000VP code has been developed to enable reactor core analysis of HTTR with full core model through extending the number of zones and meshes, and enhancing the calculation speed of CITATION code. This report describes the program changes for extending the number of zones and meshes, and for vectorization. The maximum number of zones and meshes becomes 999 and 500, respectively. The calculation speed is enhanced up to 21 times. (author)

  10. The integrated code system CASCADE-3D for advanced core design and safety analysis

    International Nuclear Information System (INIS)

    The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

  11. An industrial application of the resource based view: towards a universal methodology for identifying core competencies within an SME

    OpenAIRE

    Maddocks, Thomas

    2011-01-01

    Since the early 1990’s the resource based view and core competency concepts have been the focus of strategic management within academia and industry alike. However, it is largely documented that the vast majority of the research which has been conducted has been centred on large complex firms, often operating multibusiness and/ or multinational business strategies. This thesis addresses this void in the body of literature by studying the implications of applying the resource based view and co...

  12. Breeding ratio analysis for the improved Flower-SCWFR core

    International Nuclear Information System (INIS)

    Supercritical Water-cooled Fast Reactor (SCWFR) presents features of combining fast and light water reactor characteristics in one design. The coolant mass flow rate is just as 1/8 as in the BWR, and the neutron energy is harder than in PWR, so it would has the breeding ability. In this paper, using different models of improved Flower-SCWFR core, the void reactivity effect, power distribution,and breeding ratio are analyzed by core zoning scheme, axial coolant densities zoning,seed and blanket assembly with suitable P/D value, MOX fuel with different design and enrichment zoning, and solid uranium matrix cooled by internal clad channels in blanket assembly. Finally, an optimized modal of the improved SCWFR cores-'Flower type' is obtained. (authors)

  13. Analysis of RA-8 critical facility core in some configurations

    International Nuclear Information System (INIS)

    The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

  14. analysis of reactivity accidents in MTR for various protection system parameters and core condition

    International Nuclear Information System (INIS)

    Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

  15. Extraction of trapped gases in ice cores for isotope analysis

    International Nuclear Information System (INIS)

    The use of ice cores for paleoclimatic investigations is discussed in terms of their application for dating, temperature indication, spatial time marker synchronization, trace gas fluxes, solar variability indication and changes in the Dole effect. The different existing techniques for the extraction of gases from ice cores are discussed. These techniques, all to be carried out under vacuum, are melt-extraction, dry-extraction methods and the sublimation technique. Advantages and disadvantages of the individual methods are listed. An extensive list of references is provided for further detailed information. (author)

  16. Recent Development of the Inter-Assembly Flow Analysis Tools for SFR Core Thermal Hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. G.; Kim, E. K.; Lee, Y. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    A typical SFR core is generally comprised of hundreds of hexagonal type ducted subassemblies. And these subassemblies have hundreds of fuel rods with a triangular channel arrangement forming a closed circuit by themselves without any flow path between them. Subchannel analysis is considered to be the most suitable method for the LMR subassembly analysis when considering the geometrical complexities and computational resources needs. MATRA-LMR was developed as an analysis code to predict flow and temperature fields in SFR subassemblies. In the SFR core, flow redistribution can be occurred in the inter-assembly region of the core. The hotter counter flow from the upper center region of the LMR core may have a significant effect on the thermo-mechanical integrity of the duct wall. This paper describes the recent development of the inter-assembly flow analysis tools for SFR core thermal hydraulics and shows a few calculation results.

  17. Magnetic, Structural, and Particle Size Analysis of Single- and Multi-Core Magnetic Nanoparticles

    DEFF Research Database (Denmark)

    Ludwig, Frank; Kazakova, Olga; Barquin, Luis Fernandez;

    2014-01-01

    We have measured and analyzed three different commercial magnetic nanoparticle systems, both multi-core and single-core in nature, with the particle (core) size ranging from 20 to 100 nm. Complementary analysis methods and same characterization techniques were carried out in different labs and the...... results are compared with each other. The presented results primarily focus on determining the particle size—both the hydrodynamic size and the individual magnetic core size—as well as magnetic and structural properties. The used analysis methods include transmission electron microscopy, static and...... dynamic magnetization measurements, and Mössbauer spectroscopy. We show that particle (hydrodynamic and core) size parameters can be determined from different analysis techniques and the individual analysis results agree reasonably well. However, in order to compare size parameters precisely determined...

  18. McCARD for neutronics design and analysis of research reactor cores

    International Nuclear Information System (INIS)

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO research reactor, and YALINA subcritical facility. (authors)

  19. Verification of JUPITER standard analysis method for upgrading Joyo MK-III core design and management

    International Nuclear Information System (INIS)

    The verification of calculated core characteristics of the Joyo MK-III core using the JUPITER fast reactor standard analysis method was conducted by comparing with the measured values through the core physics tests. The purpose is to upgrade the core performance to increase the driver fuel burn-up and to increase the excess reactivity necessary for conducting various irradiation tests in the core region. Most of the C/Es are within 5% of unity. Through the comparisons, the calculation accuracy of the JUPITER standard analysis method for a small size sodium cooled fast reactor with a hard neutron spectrum like Joyo was clarified. As a result of this study, more irradiation tests can be performed with appropriate safety margin and the efficient core and fuel management can be achieved to save the number of refueling. (author)

  20. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  1. Market Analysis Identifies Community and School Education Goals.

    Science.gov (United States)

    Lindle, Jane C.

    1989-01-01

    Principals must realize the positive effects that marketing can have on improving schools and building support for them. Market analysis forces clarification of the competing needs and interests present in the community. The four marketing phases are needs assessment, analysis, goal setting, and public relations and advertising. (MLH)

  2. Structural parameter identifiability analysis for dynamic reaction networks

    DEFF Research Database (Denmark)

    Davidescu, Florin Paul; Jørgensen, Sten Bay

    2008-01-01

    A fundamental problem in model identification is to investigate whether unknown parameters in a given model structure potentially can be uniquely recovered from experimental data. This issue of global or structural identifiability is essential during nonlinear first principles model development...... where for a given set of measured variables it is desirable to investigate which parameters may be estimated prior to spending computational effort on the actual estimation. This contribution addresses the structural parameter identifiability problem for the typical case of reaction network models. The...

  3. Beacon: A three-dimensional structural analysis code for bowing history of fast breeder reactor cores

    International Nuclear Information System (INIS)

    The core elements of an LMFBR are bowed due to radial gradients of both temperature and neutron flux in the core. Since all hexagonal elements are multiply supported by adjacent elements or the restraint system, restraint forces and bending stresses are induced. In turn, these forces and stresses are relaxed by irradiation enhanced creep of the material. The analysis of the core bowing behavior requires a three-dimensional consideration of the mechanical interactions among the core elements, because the core consists of different kinds of elements and of fuel assemblies with various burnup histories. A new computational code BEACON has been developed for analyzing the bowing behavior of an LMFBR's core in three dimensions. To evaluate mechanical interactions among core elements, the code uses the analytical method of the earlier SHADOW code. BEACON analyzes the mechanical interactions in three directions, which form angles of 600 with one another. BEACON is applied to the 600 sector of a typical LMFBR's core for analyzing the bowing history during one equilibrium cycle. 120 core elements are treated, assuming the boundary condition of rotational symmetry. The application confirms that the code can be an effective tool for parametric studies as well as for detailed structural analysis of LMFBR's core. (orig.)

  4. Analysis of a multigroup stylized CANDU half-core benchmark

    International Nuclear Information System (INIS)

    Highlights: → This paper provides a benchmark that is a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. → An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core CANDU benchmark problem. → Reference eigenvalues and selected pin and bundle fission rates are included. → 2-, 4- and 47-group Monte Carlo solutions are compared to analyze homogenization-free transport approximations that result from energy condensation. - Abstract: An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem. Reference eigenvalues and selected pin and bundle fission rates are also included. This benchmark is intended to provide computational reactor physicists and methods developers with a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. In addition to transport theory code verification, the 8-group energy structure provides reactor physicist with an ideal problem for examining cross section homogenization and collapsing effects in a full-core environment. To this end, additional 2-, 4- and 47-group full-core Monte Carlo benchmark solutions are compared to analyze homogenization-free transport approximations incurred as a result of energy group condensation.

  5. Modelling and analysis of the behavior of LWRs at severe core accidents

    International Nuclear Information System (INIS)

    With respect to the assessment of the consequences of severe accidents in light water reactors from the initiation of the accident up to the thermal failure of the reactor pressure vessel (RPV), a modular program system has been developed. Experimental results will be considered with respect to the modeling of the fuel rod behavior, e.g. deformation of the fuel rod, metal water reaction and the melting of the fuel rods. The fuel and core models allow to estimate the coolability of fuel rods and core as well as the consequences of core meltdown accidents at various pressure levels. After partial failure of the lower core retention structure, the core material will drop into the lower plenum and heat up the RPV. This strong interaction between the thermal behavior of the remaining core and the partially dropped core material has been modeled because of an accident sequence analysis. The analyses described here show, that not the entire core will fail, but a partial drop of core material into the lower plenum is likely to occur. With respect to the validation of the program system, comparison calculations with the fuel rod behavior and melt models SSYST and EXMEL will be performed. Moreover, the program system will be applied to the bundle behavior in meltdown experiments, the TMI-2 core behavior and the course of a core meltdown accident in risk studies. (orig.)

  6. 2011 annual report of validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    This report summarizes results obtained by the validation study of core analysis methods for Full MOX BWR. Validation Study of Core Analysis Methods for Full MOX BWR is aimed at compiling technical data base, which is used to assess calculated characteristics of MOX cores from initial (MOX-one-third-loaded core) to full MOX cores in a full MOX BWR nuclear plant and review the validation of nuclear design methods in safety assessment of high burnup MOX fuel cores in future. For this purpose, experiments and analysis of obtained data have been implemented for core physics experiments on irradiated MOX fuel and core physics experiments on full MOX cores. Analysis of isotopic composition measurements of MOX fuel irradiated in a MOX-one-third-loaded core, and isotopic composition measurements and analysis of the obtained data on MOX fuel irradiated in the full MOX BWR are planned. Results of core physics experiments on irradiated BWR MOX fuel (initial fissile Pu content: 5.5 wt%, fuel burnup: 61 GWd/t) in the REBUS program have been analyzed, and it was observed that biases of effective neutron multiplication factors keff's, (keff - 1.0), were -1.0%Δk (Nuclear data library: JENDL-3.2), -0.3%Δk (JENDL-3.3) and 0.2%Δk(JENDL-3.2) for the diffusion, transport and Monte Carlo calculations, respectively. The analysis results of each analysis method reproduced well the measured fission rate and activation rate distributions for fuel rods. The core analysis results obtained by using typical core analysis codes (CASMO-4/ SIMULATE-3 (JNES's version)) with coarse geometrical meshes and a few energy groups showed keff's of 0.986 and 0.993 for the six critical cores of the REBUS program, which showed underestimation. Core physics experiments of the FUBILA program included a full MOX core controlled by a B4C control rod, a core containing Gd2O3-UO2, fuel and UO2 fuel rods, a 10x10 MOX assembly core and a 9x9 reference core consisting of about 17 month elapsed MOX fuel. Analysis

  7. Use of Photogrammetry and Biomechanical Gait analysis to Identify Individuals

    DEFF Research Database (Denmark)

    Larsen, Peter Kastmand; Simonsen, Erik Bruun; Lynnerup, Niels

    Photogrammetry and recognition of gait patterns are valuable tools to help identify perpetrators based on surveillance recordings. We have found that stature but only few other measures have a satisfying reproducibility for use in forensics. Several gait variables with high recognition rates were...

  8. MRI-derived measurements of fibrous-cap and lipid-core thickness: the potential for identifying vulnerable carotid plaques in vivo

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, Rikin A. [Addenbrooke' s Hospital, University Department of Radiology, Cambridge (United Kingdom); Addenbrooke' s Hospital, Academic Department of Neurosurgery, Cambridge (United Kingdom); U-King-Im, Jean-Marie; Graves, Martin J. [Addenbrooke' s Hospital, University Department of Radiology, Cambridge (United Kingdom); Horsley, Jo; Goddard, Martin [Papworth Hospital, Department of Histopathology, Papworth Everard (United Kingdom); Kirkpatrick, Peter J. [Addenbrooke' s Hospital, Academic Department of Neurosurgery, Cambridge (United Kingdom); Gillard, Jonathan H. [Addenbrooke' s Hospital, University Department of Radiology, Cambridge (United Kingdom); Addenbrooke' s Hospital, Hills Road, Box 219, Cambridge (United Kingdom)

    2004-09-01

    Vulnerable plaques have thin fibrous caps overlying large necrotic lipid cores. Recent studies have shown that high-resolution MR imaging can identify these components. We set out to determine whether in vivo high-resolution MRI could quantify this aspect of the vulnerable plaque. Forty consecutive patients scheduled for carotid endarterectomy underwent pre-operative in vivo multi-sequence MR imaging of the carotid artery. Individual plaque constituents were characterised on MR images. Fibrous-cap and lipid-core thickness was measured on MRI and histology images. Bland-Altman plots were generated to determine the level of agreement between the two methods. Multi-sequence MRI identified 133 corresponding MR and histology slices. Plaque calcification or haemorrhage was seen in 47 of these slices. MR and histology derived fibrous cap-lipid-core thickness ratios showed strong agreement with a mean difference between MR and histology ratios of 0.02 ({+-}0.04). The intra-class correlation coefficient between two readers for measurements was 0.87 (95% confidence interval, 0.73 and 0.93). Multi-sequence, high-resolution MR imaging accurately quantified the relative thickness of fibrous-cap and lipid-core components of carotid atheromatous plaques. This may prove to be a useful tool to characterise vulnerable plaques in vivo. (orig.)

  9. MRI-derived measurements of fibrous-cap and lipid-core thickness: the potential for identifying vulnerable carotid plaques in vivo

    International Nuclear Information System (INIS)

    Vulnerable plaques have thin fibrous caps overlying large necrotic lipid cores. Recent studies have shown that high-resolution MR imaging can identify these components. We set out to determine whether in vivo high-resolution MRI could quantify this aspect of the vulnerable plaque. Forty consecutive patients scheduled for carotid endarterectomy underwent pre-operative in vivo multi-sequence MR imaging of the carotid artery. Individual plaque constituents were characterised on MR images. Fibrous-cap and lipid-core thickness was measured on MRI and histology images. Bland-Altman plots were generated to determine the level of agreement between the two methods. Multi-sequence MRI identified 133 corresponding MR and histology slices. Plaque calcification or haemorrhage was seen in 47 of these slices. MR and histology derived fibrous cap-lipid-core thickness ratios showed strong agreement with a mean difference between MR and histology ratios of 0.02 (±0.04). The intra-class correlation coefficient between two readers for measurements was 0.87 (95% confidence interval, 0.73 and 0.93). Multi-sequence, high-resolution MR imaging accurately quantified the relative thickness of fibrous-cap and lipid-core components of carotid atheromatous plaques. This may prove to be a useful tool to characterise vulnerable plaques in vivo. (orig.)

  10. Core networks for visual-concrete and abstract thought content: a brain electric microstate analysis.

    Science.gov (United States)

    Lehmann, Dietrich; Pascual-Marqui, Roberto D; Strik, Werner K; Koenig, Thomas

    2010-01-01

    Commonality of activation of spontaneously forming and stimulus-induced mental representations is an often made but rarely tested assumption in neuroscience. In a conjunction analysis of two earlier studies, brain electric activity during visual-concrete and abstract thoughts was studied. The conditions were: in study 1, spontaneous stimulus-independent thinking (post-hoc, visual imagery or abstract thought were identified); in study 2, reading of single nouns ranking high or low on a visual imagery scale. In both studies, subjects' tasks were similar: when prompted, they had to recall the last thought (study 1) or the last word (study 2). In both studies, subjects had no instruction to classify or to visually imagine their thoughts, and accordingly were not aware of the studies' aim. Brain electric data were analyzed into functional topographic brain images (using LORETA) of the last microstate before the prompt (study 1) and of the word-type discriminating event-related microstate after word onset (study 2). Conjunction analysis across the two studies yielded commonality of activation of core networks for abstract thought content in left anterior superior regions, and for visual-concrete thought content in right temporal-posterior inferior regions. The results suggest that two different core networks are automatedly activated when abstract or visual-concrete information, respectively, enters working memory, without a subject task or instruction about the two classes of information, and regardless of internal or external origin, and of input modality. These core machineries of working memory thus are invariant to source or modality of input when treating the two types of information. PMID:19646538

  11. Methodology for seismic analysis of FBR core assembly using variable added damping

    International Nuclear Information System (INIS)

    It is necessary to consider the fluid-solid coupling effect due to the interaction between coolant and core assembly when analyzing the seismic performance of the FBR core assembly. The added damping was treated mostly as a constant in previous researches. In fact, the effect on assemblies from the coolant depends strongly on the gap between the core assemblies, and the damping should be considered as a variable. In order to simulate the vibration of the core assembly more accurately, the methodology for the seismic analysis of FBR core assemblies using variable added damping was studied. In this paper, the seismic analysis model of one single row of core assemblies (5 assemblies) of FBR was established. By comparing the two kinds of added damping models, the constant and variable ones respectively, the results show that the seismic analysis of the core assemblies with variable added damping is feasible and effective. Meanwhile, the simulation method used in this paper can obtain more precise approximation of the vibration of the core assembly and lays the foundation for more realistically simulating the seismic response of reactor core assemblies. It also helps to reduce the conservative margin of the structural design and is meaningful in engineering application. (authors)

  12. Heterogeneous Multi core processors for improving the efficiency of Market basket analysis algorithm in data mining

    OpenAIRE

    L, Aashiha Priyadarshni.

    2014-01-01

    Heterogeneous multi core processors can offer diverse computing capabilities. The efficiency of Market Basket Analysis Algorithm can be improved with heterogeneous multi core processors. Market basket analysis algorithm utilises apriori algorithm and is one of the popular data mining algorithms which can utilise Map/Reduce framework to perform analysis. The algorithm generates association rules based on transactional data and Map/Reduce motivates to redesign and convert the existing sequentia...

  13. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  14. Rice transcriptome analysis to identify possible herbicide quinclorac detoxification genes

    OpenAIRE

    Xu, Wenying; Di, Chao; Zhou, Shaoxia; Liu, Jia; LI Li; Liu, Fengxia; Yang, Xinling; Ling, Yun; Su, Zhen

    2015-01-01

    Quinclorac is a highly selective auxin-type herbicide and is widely used in the effective control of barnyard grass in paddy rice fields, improving the world's rice yield. The herbicide mode of action of quinclorac has been proposed, and hormone interactions affecting quinclorac signaling has been identified. Because of widespread use, quinclorac may be transported outside rice fields with the drainage waters, leading to soil and water pollution and other environmental health problems. In thi...

  15. Association analysis identifies ZNF750 regulatory variants in psoriasis

    Directory of Open Access Journals (Sweden)

    Birnbaum Ramon Y

    2011-12-01

    Full Text Available Abstract Background Mutations in the ZNF750 promoter and coding regions have been previously associated with Mendelian forms of psoriasis and psoriasiform dermatitis. ZNF750 encodes a putative zinc finger transcription factor that is highly expressed in keratinocytes and represents a candidate psoriasis gene. Methods We examined whether ZNF750 variants were associated with psoriasis in a large case-control population. We sequenced the promoter and exon regions of ZNF750 in 716 Caucasian psoriasis cases and 397 Caucasian controls. Results We identified a total of 47 variants, including 38 rare variants of which 35 were novel. Association testing identified two ZNF750 haplotypes associated with psoriasis (p ZNF750 promoter and 5' UTR variants displayed a 35-55% reduction of ZNF750 promoter activity, consistent with the promoter activity reduction seen in a Mendelian psoriasis family with a ZNF750 promoter variant. However, the rare promoter and 5' UTR variants identified in this study did not strictly segregate with the psoriasis phenotype within families. Conclusions Two haplotypes of ZNF750 and rare 5' regulatory variants of ZNF750 were found to be associated with psoriasis. These rare 5' regulatory variants, though not causal, might serve as a genetic modifier of psoriasis.

  16. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated

  17. A TEM analysis of nanoparticulates in a Polar ice core

    International Nuclear Information System (INIS)

    This paper explores the prospect for analyzing nanoparticulates in age-dated ice cores representing times in antiquity to establish a historical reference for atmospheric particulate regimes. Analytical transmission electron microscope (TEM) techniques were utilized to observe representative ice-melt water drops dried down on carbon/formvar or similar coated grids. A 10,000-year-old Greenland ice core was melted, and representative water drops were transferred to coated grids in a clean room environment. Essentially, all particulates observed were aggregates and either crystalline or complex mixtures of nanocrystals. Especially notable was the observation of carbon nanotubes and related fullerene-like nanocrystal forms. These observations are similar with some aspects of contemporary airborne particulates including carbon nanotubes and complex nanocrystal aggregates

  18. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  19. Joyo MK-II core bowing analysis based on thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    A study on the inherent safety test at the Experimental Fast Reactor Joyo has been performed to demonstrate the inherent safety of fast breeder reactors. In this study, emphasis was placed on the improvement on the feedback reactivity calculation accuracy. The investigation work for core bowing calculation has been continued because it is expected to cause negative feedback reactivity that would improve the passive safety of a fast breeder reactor. The core bowing behavior in JOYO has been analyzed by the system which consists of the MK-II core management code system MAGI, the interface code TETRAS and the core bowing calculation code BEACON. As it was supposed that the coolant flow inside of the reactor vessel might effects on wrapper tube temperature, detailed coolant flow was calculated by single phase multidimensional thermal-hydraulic analysis code AQUA. (1) As a result of the AQUA calculation, it was made clear that the coolant flow effect on the coolant temperature was negligible in fuel region. (2) The coolant temperature at the outlet of reflectors adjacent to a fuel subassembly are affected by the coolant flow that comes from the outlet of reflectors in the 6th and the 7th row. It decreases the outlet temperature of the reflectors in the 5th row in AQUA calculation. (3) High temperature coolant flow exists in neighbor of the outlet of reflectors in the 8-10th row. As a result, coolant temperature calculated by AQUA are higher in 30-40degC than that calculated by TETRAS. It was made clear that the coolant flow inside of the reactor vessel had no effect on driver fuel bowing, which was the dominant factor of the core bowing reactivity. On the other hand, in reflectors region, it affects the wrapper tube temperature, which determine the irreversible swelling and creep. Essentially, in order to verify the feedback reactivity effect caused by the core bowing, it is desired to measure the mechanical behavior of the subassemblies under power operation, but it is

  20. Supermode analysis of the 18-core photonic crystal fiber laser

    Institute of Scientific and Technical Information of China (English)

    王远; 姚建铨; 郑一博; 温午麒; 陆颖; 王鹏

    2012-01-01

    The modal of 18-core photonic crystal fiber laser is discussed and calculated.And corresponding far-field distribution of the supermodes is given by Fresnel diffraction integral.For improving beam quality,the mode selection method based on the Talbot effect is introduced.The reflection coefficients are calculated,and the result shows that an in-phase supermode can be locked better at a large propagation distance.

  1. Core support block thermal mixing test analysis report

    International Nuclear Information System (INIS)

    The extent of gas mixing and pressure drop within the core support block was experimentally investigated for various geometric and height configurations. These tests were conducted by the Experimental Engineering Branch of General Atomic Company. As a result of this investigation, the core support block thermal mixing and pressure drop has been quantified. Thermal mixing and the temperature sensor accuracy can be substantially improved at the cost of higher pressure drop. A 70-degree miter angle configuration is recommended for the reference design of the HTGR core support block (CSB). The recommended CSB height will depend on further evaluation of the possible range of variations in fuel region reactor conditions to be determmined by the Systems Engineering Department. The average temperature in a rodded region (a region with control rods in the lowered position) can be measured by the temperature sensor to within a 450F accuracy, a big improvement from an early CSB design tested by the Commissariat a La Energie Atomique at Saclay, France in 1974 and 1975

  2. Application of nodal equivalence parameters to prismatic VHTR core analysis

    International Nuclear Information System (INIS)

    The generation of nodal cross sections and equivalence parameters for prismatic VHTR core components is discussed. For fuel-block cross section generation, a conventional single-block model with a reflective boundary condition is used. A one-dimensional fuel-reflector model is proposed for reflector cross section generation in order to accurately represent the significant neutron spectrum variation at the core-reflector interface. Two-dimensional multi-block models are used for obtaining control rod cross sections for rodded fuel and reflector blocks to best approximate actual spectra in the blocks. The verification of the models was performed by generating cross sections with the DRAGON and HELIOS codes, using the cross section data in 2-D and 3-D DIF3D nodal calculations, and comparing the results to MCNP4C ones. The results show that the use of discontinuity factors reduces errors in nodal solutions for the multiplication factor and power distribution. Surface-dependent discontinuity factors are found essential for improving the accuracy of the power distribution of cores with asymmetrically rodded blocks when nodal calculations are performed with one node per hexagonal block. (authors)

  3. Evaluation of energy system analysis techniques for identifying underground facilities

    Energy Technology Data Exchange (ETDEWEB)

    VanKuiken, J.C.; Kavicky, J.A.; Portante, E.C. [and others

    1996-03-01

    This report describes the results of a study to determine the feasibility and potential usefulness of applying energy system analysis techniques to help detect and characterize underground facilities that could be used for clandestine activities. Four off-the-shelf energy system modeling tools were considered: (1) ENPEP (Energy and Power Evaluation Program) - a total energy system supply/demand model, (2) ICARUS (Investigation of Costs and Reliability in Utility Systems) - an electric utility system dispatching (or production cost and reliability) model, (3) SMN (Spot Market Network) - an aggregate electric power transmission network model, and (4) PECO/LF (Philadelphia Electric Company/Load Flow) - a detailed electricity load flow model. For the purposes of most of this work, underground facilities were assumed to consume about 500 kW to 3 MW of electricity. For some of the work, facilities as large as 10-20 MW were considered. The analysis of each model was conducted in three stages: data evaluation, base-case analysis, and comparative case analysis. For ENPEP and ICARUS, open source data from Pakistan were used for the evaluations. For SMN and PECO/LF, the country data were not readily available, so data for the state of Arizona were used to test the general concept.

  4. Real-time analysis application for identifying bursty local areas related to emergency topics.

    Science.gov (United States)

    Sakai, Tatsuhiro; Tamura, Keiichi

    2015-01-01

    Since social media started getting more attention from users on the Internet, social media has been one of the most important information source in the world. Especially, with the increasing popularity of social media, data posted on social media sites are rapidly becoming collective intelligence, which is a term used to refer to new media that is displacing traditional media. In this paper, we focus on geotagged tweets on the Twitter site. These geotagged tweets are referred to as georeferenced documents because they include not only a short text message, but also the documents' posting time and location. Many researchers have been tackling the development of new data mining techniques for georeferenced documents to identify and analyze emergency topics, such as natural disasters, weather, diseases, and other incidents. In particular, the utilization of geotagged tweets to identify and analyze natural disasters has received much attention from administrative agencies recently because some case studies have achieved compelling results. In this paper, we propose a novel real-time analysis application for identifying bursty local areas related to emergency topics. The aim of our new application is to provide new platforms that can identify and analyze the localities of emergency topics. The proposed application is composed of three core computational intelligence techniques: the Naive Bayes classifier technique, the spatiotemporal clustering technique, and the burst detection technique. Moreover, we have implemented two types of application interface: a Web application interface and an android application interface. To evaluate the proposed application, we have implemented a real-time weather observation system embedded the proposed application. we used actual crawling geotagged tweets posted on the Twitter site. The weather observation system successfully detected bursty local areas related to observed emergency weather topics. PMID:25918679

  5. Performance modeling and analysis of parallel Gaussian elimination on multi-core computers

    Directory of Open Access Journals (Sweden)

    Fadi N. Sibai

    2014-01-01

    Full Text Available Gaussian elimination is used in many applications and in particular in the solution of systems of linear equations. This paper presents mathematical performance models and analysis of four parallel Gaussian Elimination methods (precisely the Original method and the new Meet in the Middle –MiM– algorithms and their variants with SIMD vectorization on multi-core systems. Analytical performance models of the four methods are formulated and presented followed by evaluations of these models with modern multi-core systems’ operation latencies. Our results reveal that the four methods generally exhibit good performance scaling with increasing matrix size and number of cores. SIMD vectorization only makes a large difference in performance for low number of cores. For a large matrix size (n ⩾ 16 K, the performance difference between the MiM and Original methods falls from 16× with four cores to 4× with 16 K cores. The efficiencies of all four methods are low with 1 K cores or more stressing a major problem of multi-core systems where the network-on-chip and memory latencies are too high in relation to basic arithmetic operations. Thus Gaussian Elimination can greatly benefit from the resources of multi-core systems, but higher performance gains can be achieved if multi-core systems can be designed with lower memory operation, synchronization, and interconnect communication latencies, requirements of utmost importance and challenge in the exascale computing age.

  6. Identifying Colluvial Slopes by Airborne LiDAR Analysis

    Science.gov (United States)

    Kasai, M.; Marutani, T.; Yoshida, H.

    2015-12-01

    Colluvial slopes are one of major sources of landslides. Identifying the locations of the slopes will help reduce the risk of disasters, by avoiding building infrastructure and properties nearby, or if they are already there, by applying appropriate counter measures before it suddenly moves. In this study, airborne LiDAR data was analyzed to find their geomorphic characteristics to use for extracting their locations. The study site was set in the suburb of Sapporo City, Hokkaido in Japan. The area is underlain by Andesite and Tuff and prone to landslides. Slope angle and surface roughness were calculated from 5 m resolution DEM. These filters were chosen because colluvial materials deposit at around the angle of repose and accumulation of loose materials was considered to form a peculiar surface texture differentiable from other slope types. Field survey conducted together suggested that colluvial slopes could be identified by the filters with a probability of 80 percent. Repeat LiDAR monitoring of the site by an unmanned helicopter indicated that those slopes detected as colluviums appeared to be moving at a slow rate. In comparison with a similar study from the crushed zone in Japan, the range of slope angle indicative of colluviums agreed with the Sapporo site, while the texture was rougher due to larger debris composing the slopes.

  7. Temperature-based Instanton Analysis: Identifying Vulnerability in Transmission Networks

    Energy Technology Data Exchange (ETDEWEB)

    Kersulis, Jonas [Univ. of Michigan, Ann Arbor, MI (United States); Hiskens, Ian [Univ. of Michigan, Ann Arbor, MI (United States); Chertkov, Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Backhaus, Scott N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bienstock, Daniel [Columbia Univ., New York, NY (United States)

    2015-04-08

    A time-coupled instanton method for characterizing transmission network vulnerability to wind generation fluctuation is presented. To extend prior instanton work to multiple-time-step analysis, line constraints are specified in terms of temperature rather than current. An optimization formulation is developed to express the minimum wind forecast deviation such that at least one line is driven to its thermal limit. Results are shown for an IEEE RTS-96 system with several wind-farms.

  8. Efficacy of fractal analysis in identifying glaucomatous damage

    Science.gov (United States)

    Kim, P. Y.; Iftekharuddin, K. M.; Gunvant, P.; Tóth, M.; Holló, G.; Essock, E. A.

    2010-02-01

    In this work, we propose a novel fractal-based technique to analyze pseudo 2D representation of 1D retinal nerve fiber layer (RNFL) thickness measurement data vector set for early detection of glaucoma. In our proposed technique, we first convert the 1D RNFL data vector sets into pseudo 2D images and then exploit 2D fractal analysis (FA) technique to obtain the representative features. These 2D fractal-based features are further processed using principal component analysis (PCA) and the final classification between normal and glaucomatous eyes is obtained using Fischer's linear discriminant analysis (LDA). An independent dataset is used for training and testing the classifier. The technique is used on randomly selected GDx variable corneal compensator (VCC) eye data from 227 study participants (116 patients with glaucoma and 111 patients with healthy eyes). We compute sensitivity, specificity and area under receiver operating curve (AUROC) for statistical performance comparison with other known techniques. Our classification performance shows that fractal-based technique is superior to the standard machine classifier Nerve Fiber Indicator (NFI).

  9. Predicting missing links and identifying spurious links via likelihood analysis.

    Science.gov (United States)

    Pan, Liming; Zhou, Tao; Lü, Linyuan; Hu, Chin-Kun

    2016-01-01

    Real network data is often incomplete and noisy, where link prediction algorithms and spurious link identification algorithms can be applied. Thus far, it lacks a general method to transform network organizing mechanisms to link prediction algorithms. Here we use an algorithmic framework where a network's probability is calculated according to a predefined structural Hamiltonian that takes into account the network organizing principles, and a non-observed link is scored by the conditional probability of adding the link to the observed network. Extensive numerical simulations show that the proposed algorithm has remarkably higher accuracy than the state-of-the-art methods in uncovering missing links and identifying spurious links in many complex biological and social networks. Such method also finds applications in exploring the underlying network evolutionary mechanisms. PMID:26961965

  10. Analysis of the critical and first full power operating cores for PARR using leu oxide fuel

    International Nuclear Information System (INIS)

    This paper explains the analysis for determining the first full power operating core for PARR using LEU oxide fuel. The core configuration selected for this first full power operation contains about 6.13 kg of U-235 distributed in 19 standard and five control fuel elements. The neutron flux level is doubled when core is shifted from 5MW to 10 MW. Total nuclear power peaking factor of the core is 2.03. The analysis shows that the core can be operated safely at 5 MW with a flow rate of 520 meter cube per hour and at 10 MW with a flow rate of 900 meter cube per hour. (A.B.). 10 figs

  11. Validation of helios for ATR core follow analysis

    International Nuclear Information System (INIS)

    This work summarizes the validation analyses for the HELIOS code to support core design and safety assurance calculations of the Advanced Test Reactor (ATR). Past and current core safety assurance is performed by the PDQ-7 diffusion code; a state of the art reactor physics simulation tool from the nuclear industry's earlier days. Over the past twenty years, improvements in computational speed have enabled the use of modern neutron transport methodologies to replace the role of diffusion theory for simulation of complex systems, such as the ATR. More exact methodologies have enabled a paradigm-shift away from highly tuned codes that force compliance with a bounding safety envelope, and towards codes regularly validated against routine measurements. To validate HELIOS, the 16 ATR operational cycles from late-2009 to present were modeled. The computed power distribution was compared against data collected by the ATR's on-line power surveillance system. It was found that the ATR's lobe-powers could be determined with ±10% accuracy. Also, the ATR's cold startup shim configuration for each of these 16 cycles was estimated and compared against the reported critical position from the reactor log-book. HELIOS successfully predicted criticality within the tolerance set by the ATR startup procedure for 13 out of the 16 cycles. This is compared to 12 times for PDQ (without empirical adjustment). These findings, as well as other insights discussed in this report, suggest that HELIOS is highly suited for replacing PDQ for core safety assurance of the ATR. Furthermore, a modern verification and validation framework has been established that allows reactor and fuel performance data to be computed with a known degree of accuracy and stated uncertainty. (author)

  12. The results of BN600 hybrid benchmark core analysis

    International Nuclear Information System (INIS)

    The present paper includes the results of phase 1 (RZ, two dimensional model) calculations of the BN-600 hybrid core benchmark problem. The methods applied consisted of: diffusion approximation with calculation of direct and adjoint problems; calculation of reactivity coefficients; ABBN-93 nuclear data library processing, (18 group calculations). Phase 2 (Hex-Z, three dimensional model) consists of diffusion approximation with calculation of direct and adjoint problems; calculation of reactivity coefficients using first order perturbation theory; nuclear data processing code for the ABBN-78 data library. Results presented include: multiplication factors, Doppler coefficients, fuel and structure density coefficients, expansion coefficients, power distribution, beta-effective values, reaction rate distributions

  13. Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

    International Nuclear Information System (INIS)

    Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

  14. Potential bacterial core species associated with digital dermatitis in cattle herds identified by molecular profiling of interdigital skin samples

    DEFF Research Database (Denmark)

    Weiss Nielsen, Martin; Strube, Mikael Lenz; Isbrand, Anastasia;

    2016-01-01

    Although treponemes are consistently identified in tissue from bovine digital dermatitis (DD) lesions, the definitive etiology of this debilitating polymicrobial disease is still unresolved. To study the microbiomes of 27 DD-infected and 10 healthy interdigital skin samples, we used a combination...

  15. TRAC-BF1 thermal-hydraulic, ANSYS stress analysis for core shroud cracking phenomena

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission sent Generic Letter 94-03 informing all licensees about the intergranular stress corrosion cracking (IGSCC) of core shrouds found in both Dresden unit I and Quad Cities unit 1. The letter directed all licensees to perform safety analysis of their boiling water reactor (BWR) units. Two transients of special concern for the core shroud safety analysis include the main steam line break (MSLB) and recirculation line break transient

  16. Preliminary Verification Calculation of DeCART/CAPP System by HTTR Core Analysis

    International Nuclear Information System (INIS)

    In this study, the DeCART/CAPP system verification calculations have been performed against the Japan's HTTR (High Temperature Engineering Test Reactor) configurations. The calculations are carried out for single cell and single block models. The reference calculations are performed by the McCARD code. The two step core analysis system HELIOS/CAPP or DeCART/CAPP has been developed for VHTR core analysis by KAERI. In the system, first the HELIOS or DeCART code is used for homogenized cross-section generation, and second the CAPP is used to calculate the core physics parameters

  17. Preliminary Verification Calculation of DeCART/CAPP System by HTTR Core Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Han, Tae Young; Lee, Hyun Chul; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In this study, the DeCART/CAPP system verification calculations have been performed against the Japan's HTTR (High Temperature Engineering Test Reactor) configurations. The calculations are carried out for single cell and single block models. The reference calculations are performed by the McCARD code. The two step core analysis system HELIOS/CAPP or DeCART/CAPP has been developed for VHTR core analysis by KAERI. In the system, first the HELIOS or DeCART code is used for homogenized cross-section generation, and second the CAPP is used to calculate the core physics parameters.

  18. Fast reactor core monitoring by analysis of temperature noise

    International Nuclear Information System (INIS)

    The study shows, with the results obtained, how it is possible to approach the problem of diagnosis with a technique based on the use of algorithms for statistical pattern recognition was justifiable. The results presented here, with a view to their use for fast breeder reactor core surveillance, are very encouraging, the most important point being the data representation. For this study, it was difficult to find the most suitable parameters for characterizing the various simulated core states, however, despite this handicap, the classification algorithm provided quite acceptable results. The second point concerns the characterization of a system's evolution. The criterion defined was chosen for adaptation to our algorithm. One acertained that it was possible to characterize evolution on the basis of this criterion as long as the rejected points were not too far from the known learning sets. Under these circumstances, the advantage in characterizing evolution in that the changes in evolution occur when the rejected points have a tendency to agglomerate in a small area of space could be seen. This phenomenon thus makes it possible to forsee whether the creation of a new class is possible. Where the rejected points are far away from the known learning sets, the criterion used proved to be too sensitive and the characterization of evolution was less satisfactory

  19. Network stratification analysis for identifying function-specific network layers.

    Science.gov (United States)

    Zhang, Chuanchao; Wang, Jiguang; Zhang, Chao; Liu, Juan; Xu, Dong; Chen, Luonan

    2016-04-22

    A major challenge of systems biology is to capture the rewiring of biological functions (e.g. signaling pathways) in a molecular network. To address this problem, we proposed a novel computational framework, namely network stratification analysis (NetSA), to stratify the whole biological network into various function-specific network layers corresponding to particular functions (e.g. KEGG pathways), which transform the network analysis from the gene level to the functional level by integrating expression data, the gene/protein network and gene ontology information altogether. The application of NetSA in yeast and its comparison with a traditional network-partition both suggest that NetSA can more effectively reveal functional implications of network rewiring and extract significant phenotype-related biological processes. Furthermore, for time-series or stage-wise data, the function-specific network layer obtained by NetSA is also shown to be able to characterize the disease progression in a dynamic manner. In particular, when applying NetSA to hepatocellular carcinoma and type 1 diabetes, we can derive functional spectra regarding the progression of the disease, and capture active biological functions (i.e. active pathways) in different disease stages. The additional comparison between NetSA and SPIA illustrates again that NetSA could discover more complete biological functions during disease progression. Overall, NetSA provides a general framework to stratify a network into various layers of function-specific sub-networks, which can not only analyze a biological network on the functional level but also investigate gene rewiring patterns in biological processes. PMID:26879865

  20. Potential of isotope analysis (C, Cl) to identify dechlorination mechanisms

    Science.gov (United States)

    Cretnik, Stefan; Thoreson, Kristen; Bernstein, Anat; Ebert, Karin; Buchner, Daniel; Laskov, Christine; Haderlein, Stefan; Shouakar-Stash, Orfan; Kliegman, Sarah; McNeill, Kristopher; Elsner, Martin

    2013-04-01

    Chloroethenes are commonly used in industrial applications, and detected as carcinogenic contaminants in the environment. Their dehalogenation is of environmental importance in remediation processes. However, a detailed understanding frequently accounted problem is the accumulation of toxic degradation products such as cis-dichloroethylene (cis-DCE) at contaminated sites. Several studies have addressed the reductive dehalogenation reactions using biotic and abiotic model systems, but a crucial question in this context has remained open: Do environmental transformations occur by the same mechanism as in their corresponding in vitro model systems? The presented study shows the potential to close this research gap using the latest developments in compound specific chlorine isotope analysis, which make it possible to routinely measure chlorine isotope fractionation of chloroethenes in environmental samples and complex reaction mixtures.1,2 In particular, such chlorine isotope analysis enables the measurement of isotope fractionation for two elements (i.e., C and Cl) in chloroethenes. When isotope values of both elements are plotted against each other, different slopes reflect different underlying mechanisms and are remarkably insensitive towards masking. Our results suggest that different microbial strains (G. lovleyi strain SZ, D. hafniense Y51) and the isolated cofactor cobalamin employ similar mechanisms of reductive dechlorination of TCE. In contrast, evidence for a different mechanism was obtained with cobaloxime cautioning its use as a model for biodegradation. The study shows the potential of the dual isotope approach as a tool to directly compare transformation mechanisms of environmental scenarios, biotic transformations, and their putative chemical lab scale systems. Furthermore, it serves as an essential reference when using the dual isotope approach to assess the fate of chlorinated compounds in the environment.

  1. [Research of Identify Spatial Object Using Spectrum Analysis Technique].

    Science.gov (United States)

    Song, Wei; Feng, Shi-qi; Shi, Jing; Xu, Rong; Wang, Gong-chang; Li, Bin-yu; Liu, Yu; Li, Shuang; Cao Rui; Cai, Hong-xing; Zhang, Xi-he; Tan, Yong

    2015-06-01

    The high precision scattering spectrum of spatial fragment with the minimum brightness of 4.2 and the resolution of 0.5 nm has been observed using spectrum detection technology on the ground. The obvious differences for different types of objects are obtained by the normalizing and discrete rate analysis of the spectral data. Each of normalized multi-frame scattering spectral line shape for rocket debris is identical. However, that is different for lapsed satellites. The discrete rate of the single frame spectrum of normalized space debris for rocket debris ranges from 0.978% to 3.067%, and the difference of oscillation and average value is small. The discrete rate for lapsed satellites ranges from 3.118 4% to 19.472 7%, and the difference of oscillation and average value relatively large. The reason is that the composition of rocket debris is single, while that of the lapsed satellites is complex. Therefore, the spectrum detection technology on the ground can be used to the classification of the spatial fragment. PMID:26601348

  2. Identifying a preservation zone using multi–criteria decision analysis

    Directory of Open Access Journals (Sweden)

    Farashi, A.

    2016-03-01

    Full Text Available Zoning of a protected area is an approach to partition landscape into various land use units. The management of these landscape units can reduce conflicts caused by human activities. Tandoreh National Park is one of the most biologically diverse, protected areas in Iran. Although the area is generally designed to protect biodiversity, there are many conflicts between biodiversity conservation and human activities. For instance, the area is highly controversial and has been considered as an impediment to local economic development, such as tourism, grazing, road construction, and cultivation. In order to reduce human conflicts with biodiversity conservation in Tandoreh National Park, safe zones need to be established and human activities need to be moved out of the zones. In this study we used a systematic methodology to integrate a participatory process with Geographic Information Systems (GIS using a multi–criteria decision analysis (MCDA technique to guide a zoning scheme for the Tandoreh National Park, Iran. Our results show that the northern and eastern parts of the Tandoreh National Park that were close to rural areas and farmlands returned less desirability for selection as a preservation area. Rocky Mountains were the most important and most destructed areas and abandoned plains were the least important criteria for preservation in the area. Furthermore, the results reveal that the land properties were considered to be important for protection based on the obtaine

  3. Reconstruction and analysis of temperature and density spatial profiles inertial confinement fusion implosion cores

    International Nuclear Information System (INIS)

    We discuss several methods for the extraction of temperature and density spatial profiles in inertial confinement fusion implosion cores based on the analysis of the x-ray emission from spectroscopic tracers added to the deuterium fuel. The ideas rely on (1) detailed spectral models that take into account collisional-radiative atomic kinetics, Stark broadened line shapes, and radiation transport calculations, (2) the availability of narrow-band, gated pinhole and slit x-ray images, and space-resolved line spectra of the core, and (3) several data analysis and reconstruction methods that include a multi-objective search and optimization technique based on a novel application of Pareto genetic algorithms to plasma spectroscopy. The spectroscopic analysis yields the spatial profiles of temperature and density in the core at the collapse of the implosion, and also the extent of shell material mixing into the core. Results are illustrated with data recorded in implosion experiments driven by the OMEGA and Z facilities

  4. Meconium microbiome analysis identifies bacteria correlated with premature birth.

    Directory of Open Access Journals (Sweden)

    Alexandria N Ardissone

    Full Text Available Preterm birth is the second leading cause of death in children under the age of five years worldwide, but the etiology of many cases remains enigmatic. The dogma that the fetus resides in a sterile environment is being challenged by recent findings and the question has arisen whether microbes that colonize the fetus may be related to preterm birth. It has been posited that meconium reflects the in-utero microbial environment. In this study, correlations between fetal intestinal bacteria from meconium and gestational age were examined in order to suggest underlying mechanisms that may contribute to preterm birth.Meconium from 52 infants ranging in gestational age from 23 to 41 weeks was collected, the DNA extracted, and 16S rRNA analysis performed. Resulting taxa of microbes were correlated to clinical variables and also compared to previous studies of amniotic fluid and other human microbiome niches.Increased detection of bacterial 16S rRNA in meconium of infants of <33 weeks gestational age was observed. Approximately 61·1% of reads sequenced were classified to genera that have been reported in amniotic fluid. Gestational age had the largest influence on microbial community structure (R = 0·161; p = 0·029, while mode of delivery (C-section versus vaginal delivery had an effect as well (R = 0·100; p = 0·044. Enterobacter, Enterococcus, Lactobacillus, Photorhabdus, and Tannerella, were negatively correlated with gestational age and have been reported to incite inflammatory responses, suggesting a causative role in premature birth.This provides the first evidence to support the hypothesis that the fetal intestinal microbiome derived from swallowed amniotic fluid may be involved in the inflammatory response that leads to premature birth.

  5. Geochemical analysis of core from a geothermal anomaly

    International Nuclear Information System (INIS)

    A mild geothermal area in western Montana, USA, has been studied, as a natural analog, to learn about the effects that long-term heat generated by a repository containing spent nuclear fuel might have on the surrounding rock mass. The results of previous geological, geophysical and hydrogeological studies are briefly summarized. Extensive petrological studies have been undertaken on core samples obtained from a 2 km deep borehole drilled into the Empire Creek Stock. These include a detailed petrographic study, x-ray diffraction analyses, scanning electron microscope and electron microprobe analyses, porosity and permeability measurements, oxygen isotope analyses, uranium disequilibrium analyses and K-Ar age determinations. The implications to deep burial of nuclear wastes are discussed. 40 refs

  6. Using borehole core analysis to reveal Late Quaternary paleoearthquakes along the Nankou-Sunhe Fault,Beijing

    Institute of Scientific and Technical Information of China (English)

    ZHANG ShiMin; WANG DanDan; LIU XuDong; ZHANG GuoHong; ZHAO JunXiang; LUO MingHui; REN JunJie; WANG Rui; ZHANG YingLi

    2008-01-01

    The Nankou-Sunhe Fault is a buried active normal fault that traverses the urban area of Beijing. Its seismic risks have caused considerable concerns. This paper studies paleoearthquakes along this fault by analyzing and correlating bore-hole cores obtained from triple-tube coring, incorporating experience acquired from trenching. As a result, a model for identifying earthquake-derived colluvium by sediment-core analysis is proposed. Triple-tube coring technique is useful to collect continuous undisturbed soil core near the Nankou-Sunhe Fault. By identifying fault-scarp colluviums, determining cumulative displacement, and analysing stratum thickening on the hanging wall, we are able to establish a preliminary paleoearthquake sequence consisting of 13 surface-rupturing events since 60 ka. The seismic history can be divided into three periods based on different recurrence intervals. Between 60 and 40 ka, three earthquakes occurred with recurrence interval of ~10 ka. From 40 to 25 ka, there were six earthquakes with the recurrence interval of about 2.5 ka. In the last 25 ka, four earthquakes have taken place with the recurrence interval varying considerably, The recurrence interval between the last three events is ~5 ka. Smaller recurrence intervals correspond to stages of faster fault slip. The coseismic displacement of a single event is 0.8 to 2.2 m, average 1.4 m, largely equivalent to moment magnitudes 6.7-7.1. This study demonstrates the feasibility of bore-hole drilling in investigating paleoearthquakes along normal faults. It also suggests that closely spaced boreholes with continuous undisturbed cores are essential for reconstructing the complete paleoearthquake sequence.

  7. The effects of core zoning on optimization of design analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Highlights: • 1/8 of core is simulated by MCNP and thermal-hydraulic code simultaneously. • Effects of core zoning are studied by dividing the core into two regions. • Both the neutronics and thermal-hydraulic behavior are investigated. • The flat flux distribution is achieved in the optimization analysis. • The flat flux can lead to worse thermal-hydraulic behavior occasionally. - Abstract: The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly

  8. Inspection of spar-core bond in helicopter rotor blades using finite element analysis

    Science.gov (United States)

    Chakrapani, Sunil Kishore; Barnard, Daniel J.; Dayal, Vinay

    2015-03-01

    This work focuses on inspection of spar-core bond of a helicopter rotor blade using finite element analysis. Structures which have high density, high stiffness cores can be very difficult to inspect due to various mode conversions. FEM was used to capture these mode conversions effectively. The structure consists of a thin spar section followed by a spar-core half space and another thin spar section. A Lamb wave excited in the spar section can mode convert into a Rayleigh wave in the spar-core section due to the coupling of the core material. This in turn mode converts back into a Lamb wave upon interacting with the next spar section. This work focuses solely on capturing the mode conversions between Rayleigh and Lamb modes at different discontinuities in the geometry.

  9. Analysis of impurity effect on Silicide fuels of the RSG-GAS core

    International Nuclear Information System (INIS)

    Simulation of impurity effect on silicide fuel of the RSG-GAS core has been done. The aim of this research is to know impurity effect of the U-234 and U-236 isotopes in the silicide fuels on the core criticality. The silicide fuels of 250 g U loading and 19.75 of enrichment is used in this simulation. Cross section constant of fuels and non-structure material of core are generated by WIMSD/4 computer code, meanwhile impurity concentration was arranged from 0.01% to 2%. From the result of analysis can be concluded that the isotopes impurity in the fuels could make trouble in the core and the core can not be operated at critical after a half of its cycle length (350 MW D)

  10. New time-line technique for station blackout core-melt analysis

    International Nuclear Information System (INIS)

    Florida Power Corporation (FPC) has developed a new method for analyzing station blackout (SBO) core-melt accidents. This method, created during the recent probabilistic risk assessment (PRA) of Crystal River Unit 3 (CR-3), originated from the need to analyze the interactions among the two-train emergency feedwater (EFW) system, station batteries, and diesel generators (DGs) following a loss of off-site power (LOSP) event. SBO core-melt sequences for CR-3 are unique since the time core-melt commences depends on which DG fails last. The purpose of this paper is to outline the new method of analysis of SBO core-melt accidents at CR-3. The significance of SBO core-melt accidents to total plant risk, along with the efficacy of various methods to reduce SBO risk, are also discussed

  11. Reactor analysis methods. 7. MCNP4B Analysis of the HTR-10 Startup Core

    International Nuclear Information System (INIS)

    A benchmark criticality analysis was performed of the HTR-10 pebble bed reactor recently constructed at the Institute of Nuclear Energy Technology, Beijing, using the MCNP4B continuous-energy Monte Carlo code. This analysis was part of the U.S. contribution to the International Atomic Energy Agency Coordinated Research Program (CRP-5) on the evaluation of high-temperature gas-cooled reactor performance. The HTR-10 is a 10-MW(thermal) pebble bed reactor that uses graphite spheres that are 6 cm in diameter and contain embedded coated fuel particles (CFPs) with 17% enriched UO2. The uranium loading per fuel sphere is 5.0 g. The full core consists of ∼27 000 spheres randomly packed in a cylindrical cavity with a mean height of 1.97 m, a diameter of 1.8 m, and a volume of 5.0m3. The core is surrounded by a structure consisting of a graphite reflector and a borated carbon shield. The radial reflector, which is 1 m thick, contains channels for the control and shutdown systems, irradiation sites, and helium coolant. The initial approach to critical was achieved by filling the discharge tube and cone at the bottom of the core with moderator spheres, then adding a random mixture of fuel and moderator spheres until the critical mass was achieved. The total number of spheres needed to reach criticality was 16 890, with a fuel-to-moderator sphere ratio (F/M) of 57 to 43%. Although the physics benchmark problem consists of three parts, only the first part is considered here. Problem B1 calls for the prediction of the initial, cold, critical core loading with the control and shutdown absorbers completely withdrawn at 20 deg. C and a helium pressure of 3.0 MPa. The detailed MCNP4B model of the reactor included the double-heterogeneity of the CFPs and the graphite spheres, and an explicit representation of the graphite reflector. A body-centered cubic lattice was used to approximate the packing of spheres in the core, with the size of the moderator sphere reduced in a manner that

  12. Monju core physics test analysis with various nuclear data libraries

    International Nuclear Information System (INIS)

    JAEA has been re-analyzing Monju core physics tests to validate the JAEA's neutronics calculation system to be used in the next Monju core physics tests. Precedent results presented in PHYSOR2008 have demonstrated the validity of the system based on the basic physical parameters, such as criticality, control rod worth, isothermal temperature coefficient, and power coefficient. This paper is a continuation of the validation study focusing on the other parameters, such as fixed absorber reactivity worth, fuel sub-assembly reactivity worth, coolant reactivity worth, burnup coefficient, and reaction rate. The fixed absorber reactivity worth is a reactivity induced by the replacement of a blanket sub-assembly to a fixed absorber sub-assembly. The fuel sub-assembly reactivity worth is a reactivity induced by the replacement of a fuel sub-assembly to a non-fissile dummy sub-assembly. The coolant reactivity worth is a reactivity induced by the replacement of a non- fissile dummy sub-assembly containing sodium to that containing helium. The reaction rate data include the reaction rate ratio of 238U capture to 239Pu fission. Each of the data is useful to check the calculation system in a particular aspect. For example, the first two data are suitable to check the calculation accuracy of a blanket region and a fuel sub-assembly, respectively. The parameters are simulated using the JAEA's neutronics calculation system with various nuclear date libraries, JENDL-3.2, JENDL-3.3, JENDL/AC-2008, JEFF-3.1, and ENDF/B-VII. A continuous energy Monte Carlo calculation code, MVP, is employed to check calculation methods. Figure 1 shows an example of the C/E (Calculation over Experiment) values. The C/E values are within experimental errors for the fixed absorber reactivity worth and the fuel sub- assembly reactivity worth. Those for the burnup reactivity coefficient are around the experimental error and show a tendency of overestimation. About the comparison with the Monte Carlo

  13. Development of pin-by-pin core analysis method using three-dimensional direct response matrix

    International Nuclear Information System (INIS)

    A three-dimensional direct response matrix method using a Monte Carlo calculation has been developed. The direct response matrix is formalized by four subresponse matrices in order to respond to a core eigenvalue k and thus can be recomposed at each outer iteration in core analysis. The subresponse matrices can be evaluated by ordinary single fuel assembly calculations with the Monte Carlo method in three dimensions. Since these subresponse matrices are calculated for the actual geometry of the fuel assembly, the effects of intra- and inter-assembly heterogeneities can be reflected on global partial neutron current balance calculations in core analysis. To verify this method, calculations for heterogeneous systems were performed. The results obtained using this method agreed well with those obtained using direct calculations with a Monte Carlo method. This means that this method accurately reflects the effects of intra- and inter-assembly heterogeneities and can be used for core analysis. A core analysis method, in which neutronic calculations using this direct response matrix method are coupled with thermal-hydraulic calculations, has also been developed. As it requires neither diffusion approximation nor a homogenization process of lattice constants, a precise representation of the effects of neutronic heterogeneities is possible. Moreover, the fuel rod power distribution can be directly evaluated, which enables accurate evaluations of core thermal margins. A method of reconstructing the response matrices according to the condition of each node in the core has been developed. The test revealed that the neutron multiplication factors and the fuel rod neutron production rates could be reproduced by interpolating the elements of the response matrix. A coupled analysis of neutronic calculations using the direct response matrix method and thermal-hydraulic calculations for an ABWR quarter core was performed, and it was found that the thermal power and coolant

  14. Analysis of Fusarium avenaceum Metabolites Produced during Wet Apple Core Rot

    DEFF Research Database (Denmark)

    Sørensen, Jens Laurids; Phipps, Richard Kerry; Nielsen, Kristian Fog;

    2009-01-01

    Wet apple core rot (wACR) is a well-known disease of susceptible apple cultivars such as Gloster, Jona Gold, and Fuji. Investigations in apple orchards in Slovenia identified Fusarium avenaceum, a known producer of several mycotoxins, as the predominant causal agent of this disease. A LC-MS/MS me......Wet apple core rot (wACR) is a well-known disease of susceptible apple cultivars such as Gloster, Jona Gold, and Fuji. Investigations in apple orchards in Slovenia identified Fusarium avenaceum, a known producer of several mycotoxins, as the predominant causal agent of this disease. A LC...

  15. Criticality Benchmark Analysis of the HTTR Annular Startup Core Configurations

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2009-11-01

    One of the high priority benchmarking activities for corroborating the Next Generation Nuclear Plant (NGNP) Project and Very High Temperature Reactor (VHTR) Program is evaluation of Japan's existing High Temperature Engineering Test Reactor (HTTR). The HTTR is a 30 MWt engineering test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. A large amount of critical reactor physics data is available for validation efforts of High Temperature Gas-cooled Reactors (HTGRs). Previous international reactor physics benchmarking activities provided a collation of mixed results that inaccurately predicted actual experimental performance.1 Reevaluations were performed by the Japanese to reduce the discrepancy between actual and computationally-determined critical configurations.2-3 Current efforts at the Idaho National Laboratory (INL) involve development of reactor physics benchmark models in conjunction with the International Reactor Physics Experiment Evaluation Project (IRPhEP) for use with verification and validation methods in the VHTR Program. Annular cores demonstrate inherent safety characteristics that are of interest in developing future HTGRs.

  16. Criticality Benchmark Analysis of the HTTR Annular Startup Core Configurations

    International Nuclear Information System (INIS)

    One of the high priority benchmarking activities for corroborating the Next Generation Nuclear Plant (NGNP) Project and Very High Temperature Reactor (VHTR) Program is evaluation of Japan's existing High Temperature Engineering Test Reactor (HTTR). The HTTR is a 30 MWt engineering test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. A large amount of critical reactor physics data is available for validation efforts of High Temperature Gas-cooled Reactors (HTGRs). Previous international reactor physics benchmarking activities provided a collation of mixed results that inaccurately predicted actual experimental performance.1 Reevaluations were performed by the Japanese to reduce the discrepancy between actual and computationally-determined critical configurations.2-3 Current efforts at the Idaho National Laboratory (INL) involve development of reactor physics benchmark models in conjunction with the International Reactor Physics Experiment Evaluation Project (IRPhEP) for use with verification and validation methods in the VHTR Program. Annular cores demonstrate inherent safety characteristics that are of interest in developing future HTGRs.

  17. Thermal buckling analysis of truss-core sandwich plates

    Institute of Scientific and Technical Information of China (English)

    陈继伟; 刘咏泉; 刘伟; 苏先樾

    2013-01-01

    Truss-core sandwich plates have received much attention in virtue of the high values of strength-to-weight and stiffness-to-weight as well as the great ability of impulse-resistance recently. It is necessary to study the stability of sandwich panels under the influence of the thermal load. However, the sandwich plates are such complex three-dimensional (3D) systems that direct analytical solutions do not exist, and the finite element method (FEM) cannot represent the relationship between structural parameters and mechanical properties well. In this paper, an equivalent homogeneous continuous plate is idealized by obtaining the effective bending and transverse shear stiffness based on the characteristics of periodically distributed unit cells. The first order shear deformation theory for plates is used to derive the stability equation. The buckling temperature of a simply supported sandwich plate is given and verified by the FEM. The effect of related parameters on mechanical properties is investigated. The geometric parameters of the unit cell are optimized to attain the maximum buckling temperature. It is shown that the optimized sandwich plate can improve the resistance to thermal buckling significantly.

  18. Gas-core reactor power transient analysis. Final report

    International Nuclear Information System (INIS)

    The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)

  19. Development of core thermal hydraulic analysis methodology using multichannel code system

    International Nuclear Information System (INIS)

    A multi-channel core analysis model using a subchannel code TORC is developed to improve the thermal margin, and is assessed and compared with the existing single-channel analysis model. To apply the TORC code to the w-type reactor core, a hot subchannel DNBR analysis model is developed using the lumping technology. In addition, the sensitivity of TORC to various models and input parameters are carried out to appreciate the code characteristics. The developed core analysis model is applied to the evaluation of the thermal margin for 17 x 17 KOFA loaded core. For this calculation, the KRB1 CHF correlation is developed on the basis of w and Siemens bundle CHF data, and the DNB design limit is established using the STDP method. From the result of the steady-state and transient analysis of the 17 x 17 KOFA loaded core, it is found that the extra 10% DNBR margin can be obtained compared with the existing single-channel analysis methodology. (Author) 65 figs., 12 tabs

  20. Extracting the core indicators of pulverized coal for blast furnace injection based on principal component analysis

    Science.gov (United States)

    Guo, Hong-wei; Su, Bu-xin; Zhang, Jian-liang; Zhu, Meng-yi; Chang, Jian

    2013-03-01

    An updated approach to refining the core indicators of pulverized coal used for blast furnace injection based on principal component analysis is proposed in view of the disadvantages of the existing performance indicator system of pulverized coal used in blast furnaces. This presented method takes into account all the performance indicators of pulverized coal injection, including calorific value, igniting point, combustibility, reactivity, flowability, grindability, etc. Four core indicators of pulverized coal injection are selected and studied by using principal component analysis, namely, comprehensive combustibility, comprehensive reactivity, comprehensive flowability, and comprehensive grindability. The newly established core index system is not only beneficial to narrowing down current evaluation indices but also effective to avoid previous overlapping problems among indicators by mutually independent index design. Furthermore, a comprehensive property indicator is introduced on the basis of the four core indicators, and the injection properties of pulverized coal can be overall evaluated.

  1. A SAS2H/KENO-V Methodology for 3D Full Core depletion analysis

    International Nuclear Information System (INIS)

    This paper describes the use of a SAS2H/KENO-V methodology for 3D full core depletion analysis and illustrates its capabilities by applying it to burnup analysis of the IRIS core benchmarks. This new SAS2H/KENO-V sequence combines a 3D Monte Carlo full core calculation of node power distribution and a 1D Wigner-Seitz equivalent cell transport method for independent depletion calculation of each of the nodes. This approach reduces by more than an order of magnitude the time required for getting comparable results using the MOCUP code system. The SAS2H/KENO-V results for the asymmetric IRIS core benchmark are in good agreement with the results of the ALPHA/PHOENIX/ANC code system. (author)

  2. Development of Computer Program for Whole Core Thermal-Hydraulic Analysis of Fast Reactors

    International Nuclear Information System (INIS)

    A whole core thermal-hydraulic analysis program ACT was developed for the purpose of evaluating detailed in-core thermal-hydraulic phenomena of sodium cooled fast reactors under various reactor operation conditions. ACT consists of four kinds of calculation modules, i.e., fuel-assembly, inter-wrapper gap (core barrel), upper plenum and heat transport system modules. The latter two modules give proper boundary conditions for the reactor core thermal-hydraulic analysis. These four modules are coupled with each other by using MPI and calculate simultaneously on a cluster workstation. ACT was applied to analyzing a sodium experiment performed at JNC, which simulated the natural circulation decay heat removal under PRACS and DRACS operation condition. In the experiment, not only inter-wrapper flows but also reverses flows in the fuel assemblies were observed. ACT succeeded in simulating such complicated phenomena. (authors)

  3. Steady state analysis of SFR cores using DYN3D-Serpent codes sequence

    International Nuclear Information System (INIS)

    A few-group cross section generation methodology for the deterministic analysis of SFR cores with DYN3D code has been proposed. The full core DYN3D results obtained using the few-group constants produced by Serpent agreed very well with that of the reference full core MC simulations. Such an agreement demonstrates the feasibility of the proposed few-group cross section generation procedure. In summary, this study showed that the Serpent-DYN3D code sequence can be successfully used for modeling fast spectrum reactor systems. (orig.)

  4. European contribution to Phase 3 of the benchmark core analysis for the BN-600 hybrid reactor

    International Nuclear Information System (INIS)

    This European participation in Phase 3 of the benchmark (BN-600) analysis consist of a joint contribution from France and the UK. Calculations were performed by ERANOS code and data system which has been developed in the framework of European cooperation on fast reactors. Results are presented for all the core neutronic parameters, both for homogeneous and heterogeneous core models and both for beginning and end of fuel cycle

  5. Analysis of Maize Crop Leaf using Multivariate Image Analysis for Identifying Soil Deficiency

    Directory of Open Access Journals (Sweden)

    S. Sridevy

    2014-11-01

    Full Text Available Image processing analysis for the soil deficiency identification has become an active area of research in this study. The changes in the color of the leaves are used to analyze and identify the deficiency of soil nutrients such as Nitrogen (N, Phosphorus (P and potassium (K by digital color image analysis. This research study focuses on the image analysis of the maize crop leaf using multivariate image analysis. In this proposed novel approach, initially, a color transformation for the input RGB image is formed and this RGB is converted to HSV because RGB is ideal for color generation but HSV is very suitable for color perception. Then green pixels are masked and removed using specific threshold value by applying histogram equalization. This masking approach is done through specific customized filtering approach which exclusively filters the green color of the leaf. After the filtering step, only the deficiency part of the leaf is taken for consideration. Then, a histogram generation is carried out for the deficiency part of the leaf. Then, Multivariate Image Analysis approach using Independent Component Analysis (ICA is carried out to extract a reference eigenspace from a matrix built by unfolding color data from the deficiency part. Test images are also unfolded and projected onto the reference eigenspace and the result is a score matrix which is used to compute nutrient deficiency based on the T2 statistic. In addition, a multi-resolution scheme by scaling down process is carried out to speed up the process. Finally, based on the training samples, the soil deficiency is identified based on the color of the maize crop leaf.

  6. Steady state thermal hydraulic analysis of LMR core using COBRA-K code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Kim, Young Gyun; Kim Young In; Kim Young Cheol

    1997-02-01

    A thermal hydraulics analysis code COBRA-K is being developed by the KAERI LMR core design technology development team. COBRA-K is a part of the integrated computation system for LMR core design and analysis, the K-CORE system. COBRA-K is supposed to predict the flow and temperature distributions in LMR core. COBRA-K is an extension of the previously published COBRA-IV-I code with several functional improvements. Specially COBRA-K has been improved to analyze single and multi-assembly, and whole-core in the transient condition. This report describes the overall features of COBRA-K and gives general input descriptions. The 19 pin assembly experimental data of ORNL were used to verify the accuracy of this code for the steady state analysis. The comparative results show good agreements between the calculated and the measured data. And COBRA-K can be used to predict flow and temperature distributions for the LMR core design. (author). 7 refs., 6 tabs., 13 figs.

  7. Development of flow network analysis code for block type VHTR core by linear theory method

    International Nuclear Information System (INIS)

    VHTR (Very High Temperature Reactor) is high-efficiency nuclear reactor which is capable of generating hydrogen with high temperature of coolant. PMR (Prismatic Modular Reactor) type reactor consists of hexagonal prismatic fuel blocks and reflector blocks. The flow paths in the prismatic VHTR core consist of coolant holes, bypass gaps and cross gaps. Complicated flow paths are formed in the core since the coolant holes and bypass gap are connected by the cross gap. Distributed coolant was mixed in the core through the cross gap so that the flow characteristics could not be modeled as a simple parallel pipe system. It requires lot of effort and takes very long time to analyze the core flow with CFD analysis. Hence, it is important to develop the code for VHTR core flow which can predict the core flow distribution fast and accurate. In this study, steady state flow network analysis code is developed using flow network algorithm. Developed flow network analysis code was named as FLASH code and it was validated with the experimental data and CFD simulation results. (authors)

  8. Decay heat analysis of a VHTR core using the HELIOS and origen-2 codes

    International Nuclear Information System (INIS)

    This paper describes the procedure and results of a decay heat analysis in a relatively short time after a shutdown for the safety analysis of a VHTR core. In this analysis, HELIOS provides the one-group actinide cross sections to ORIGEN-2 through the 190 group lattice calculation for a single fuel block. Then, ORIGEN-2 performs the depletion and decay heat calculations using these actinide cross sections. After benchmarking this procedure against a PWR core, it was applied to a 200 MWth prismatic VHTR core. The results showed that the decay heat per unit operating power is very comparable to that for a typical large power PWR core, although the decay heat per unit heavy metal mass is three times higher than that in the PWR core. It was also found from the results that the decay heat fraction to operating power decreases very slowly with the core burnup after it reaches a maximum value of 6.1 percents at 5 GWd/tHM. (authors)

  9. Transient and stability analysis of a BWR core with thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779 Col. Narvarte, 03020 Mexico, DF (Mexico); Espinosa-Paredes, Gilberto [Division de Ciencias Basicas e Ingenieria, Universidad Autonoma Metropolitana, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, DF (Mexico)], E-mail: gepe@xanum.uam.mx; Francois, Juan-Luis [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec Mor. (Mexico)

    2008-08-15

    The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket-seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal-hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO{sub 2} core, even during transient conditions. The stability and transient analysis show that the thorium-uranium fuel can be operated safely in current BWRs.

  10. HORECA. Hoger onderwijs reactor elementary core analysis system. User's manual

    International Nuclear Information System (INIS)

    HORECA is developed at IRI Delft for quick analysis of power distribution, burnup and safety for the HOR. It can be used for the manual search of a better loading of the reactor. HORECA is based on the Penn State Fuel Management Package and uses the MCRAC code included in this package as a calculation engine. (orig./HP)

  11. SMAD-2A mass spectrometer core isotopic analysis of water

    International Nuclear Information System (INIS)

    SMAT-2A is a small mass spectrometer for water isotopes analysis, double collector method (HD/H2 ratio). Improvements in maintenance automation data acquisition and processing are presented. Analyses of liquid and gaseous samples were made and internal reproducibility and precision are calculated. (Author)

  12. Analysis of void coefficient in fast spectrum BWR core with Monte Carlo code 'MVP'

    International Nuclear Information System (INIS)

    An innovative large BWR core concept has been proposed for aiming at fuel breeding as well as negative void reactivity coefficient. The core consists of two types of MOX fuel assemblies. One is a triangular tight lattice bundle 1.6 m in active core height and the other is the same bundle 0.8 m. The ratio of flow area to fuel area of the bundle is set at about 0.5 in order to increase breeding ratio. A neutron-streaming channel that consists of a cavity-can containing helium gas and a flow gap between the cavity-can and the channel box is located above each short bundle. It will decrease void reactivity coefficient by enhancing neutron leakage from the core when the void fraction is increased in the flow gap. A core composed of tight lattice bundles provides a much harder neutron spectrum than that of conventional BWRs but a slightly softer one than that of typical FBRs. The cavity-can and the flow gap will cause a steep gradient of neutron flux. The neutronics for such a complicated core structure could not be properly analyzed by conventional analysis methods. In particular, the analysis of void reactivity coefficient requires a sophisticated method because it deals with a small change in core composition. In the analysis of the void reactivity coefficient, we adopted a three-dimensional Monte Carlo code 'MVP', which has been developed by JAERI and has many advantages such as an easy input form for lattice structures, a short run time and a continuous neutron energy method. The continuous neutron energy method is important for the analysis of this core because fission reactions occur mainly in the resonance energy region, where the evaluation of accurate cross sections is difficult with conventional methods. The library used is JENDL-3.2. The multi-layer structure of lattices is also essential for the analysis because its hard spectrum and relatively long neutron mean free path require a modeling for the full core with a lot of bundles. The analysis indicates that

  13. Analysis of ultra-deep pyrosequencing and cloning based sequencing of the basic core promoter/precore/core region of hepatitis B virus using newly developed bioinformatics tools.

    Directory of Open Access Journals (Sweden)

    Mukhlid Yousif

    Full Text Available AIMS: The aims of this study were to develop bioinformatics tools to explore ultra-deep pyrosequencing (UDPS data, to test these tools, and to use them to determine the optimum error threshold, and to compare results from UDPS and cloning based sequencing (CBS. METHODS: Four serum samples, infected with either genotype D or E, from HBeAg-positive and HBeAg-negative patients were randomly selected. UDPS and CBS were used to sequence the basic core promoter/precore region of HBV. Two online bioinformatics tools, the "Deep Threshold Tool" and the "Rosetta Tool" (http://hvdr.bioinf.wits.ac.za/tools/, were built to test and analyze the generated data. RESULTS: A total of 10952 reads were generated by UDPS on the 454 GS Junior platform. In the four samples, substitutions, detected at 0.5% threshold or above, were identified at 39 unique positions, 25 of which were non-synonymous mutations. Sample #2 (HBeAg-negative, genotype D had substitutions in 26 positions, followed by sample #1 (HBeAg-negative, genotype E in 12 positions, sample #3 (HBeAg-positive, genotype D in 7 positions and sample #4 (HBeAg-positive, genotype E in only four positions. The ratio of nucleotide substitutions between isolates from HBeAg-negative and HBeAg-positive patients was 3.5 ∶ 1. Compared to genotype E isolates, genotype D isolates showed greater variation in the X, basic core promoter/precore and core regions. Only 18 of the 39 positions identified by UDPS were detected by CBS, which detected 14 of the 25 non-synonymous mutations detected by UDPS. CONCLUSION: UDPS data should be approached with caution. Appropriate curation of read data is required prior to analysis, in order to clean the data and eliminate artefacts. CBS detected fewer than 50% of the substitutions detected by UDPS. Furthermore it is important that the appropriate consensus (reference sequence is used in order to identify variants correctly.

  14. Analysis of Ultra-Deep Pyrosequencing and Cloning Based Sequencing of the Basic Core Promoter/Precore/Core Region of Hepatitis B Virus Using Newly Developed Bioinformatics Tools

    Science.gov (United States)

    Yousif, Mukhlid; Bell, Trevor G.; Mudawi, Hatim; Glebe, Dieter; Kramvis, Anna

    2014-01-01

    Aims The aims of this study were to develop bioinformatics tools to explore ultra-deep pyrosequencing (UDPS) data, to test these tools, and to use them to determine the optimum error threshold, and to compare results from UDPS and cloning based sequencing (CBS). Methods Four serum samples, infected with either genotype D or E, from HBeAg-positive and HBeAg-negative patients were randomly selected. UDPS and CBS were used to sequence the basic core promoter/precore region of HBV. Two online bioinformatics tools, the “Deep Threshold Tool” and the “Rosetta Tool” (http://hvdr.bioinf.wits.ac.za/tools/), were built to test and analyze the generated data. Results A total of 10952 reads were generated by UDPS on the 454 GS Junior platform. In the four samples, substitutions, detected at 0.5% threshold or above, were identified at 39 unique positions, 25 of which were non-synonymous mutations. Sample #2 (HBeAg-negative, genotype D) had substitutions in 26 positions, followed by sample #1 (HBeAg-negative, genotype E) in 12 positions, sample #3 (HBeAg-positive, genotype D) in 7 positions and sample #4 (HBeAg-positive, genotype E) in only four positions. The ratio of nucleotide substitutions between isolates from HBeAg-negative and HBeAg-positive patients was 3.5∶1. Compared to genotype E isolates, genotype D isolates showed greater variation in the X, basic core promoter/precore and core regions. Only 18 of the 39 positions identified by UDPS were detected by CBS, which detected 14 of the 25 non-synonymous mutations detected by UDPS. Conclusion UDPS data should be approached with caution. Appropriate curation of read data is required prior to analysis, in order to clean the data and eliminate artefacts. CBS detected fewer than 50% of the substitutions detected by UDPS. Furthermore it is important that the appropriate consensus (reference) sequence is used in order to identify variants correctly. PMID:24740330

  15. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

    2014-10-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

  16. Sensitivity analysis of core parameters to fuel manufacturing uncertainties

    International Nuclear Information System (INIS)

    Heterogeneities cannot be avoided at the manufacturing stage of nuclear fuel. In the project, control parameters are always best estimate computed, so that heterogeneities in the fuel affect them as an uncertainty. These heterogeneities cannot be only represented by a mean value and a standard deviation: their joint distribution must be evaluated according to post-manufacture statistical analysis. In order to improve the nuclear fuel performances and respect the safety rules, FRAMATOME has developed an original analysis method based on stochastic approach and the generalized perturbation theory. The joint low of the pointwise power variation is given by the sensitivity matrix, determined by the generalized perturbation theory, using the first order Taylor stochastic development. (authors). 8 refs., 5 figs

  17. Building waste management core indicators through Spatial Material Flow Analysis: Net recovery and transport intensity indexes

    International Nuclear Information System (INIS)

    Highlights: ► Sustainability and proximity principles have a key role in waste management. ► Core indicators are needed in order to quantify and evaluate them. ► A systematic, step-by-step approach is developed in this study for their development. ► Transport may play a significant role in terms of environmental and economic costs. ► Policy action is required in order to advance in the consecution of these principles. - Abstract: In this paper, the material and spatial characterization of the flows within a municipal solid waste (MSW) management system are combined through a Network-Based Spatial Material Flow Analysis. Using this information, two core indicators are developed for the bio-waste fraction, the Net Recovery Index (NRI) and the Transport Intensity Index (TII), which are aimed at assessing progress towards policy-related sustainable MSW management strategies and objectives. The NRI approaches the capacity of a MSW management system for converting waste into resources through a systematic metabolic approach, whereas the TII addresses efficiency in terms of the transport requirements to manage a specific waste flow throughout the entire MSW management life cycle. Therefore, both indicators could be useful in assessing key MSW management policy strategies, such as the consecution of higher recycling levels (sustainability principle) or the minimization of transport by locating treatment facilities closer to generation sources (proximity principle). To apply this methodological approach, the bio-waste management system of the region of Catalonia (Spain) has been chosen as a case study. Results show the adequacy of both indicators for identifying those points within the system with higher capacity to compromise its environmental, economic and social performance and therefore establishing clear targets for policy prioritization. Moreover, this methodological approach permits scenario building, which could be useful in assessing the outcomes of

  18. Analysis of core samples from the BPXA-DOE-USGS Mount Elbert gas hydrate stratigraphic test well: Insights into core disturbance and handling

    Energy Technology Data Exchange (ETDEWEB)

    Kneafsey, Timothy J.; Lu, Hailong; Winters, William; Boswell, Ray; Hunter, Robert; Collett, Timothy S.

    2009-09-01

    Collecting and preserving undamaged core samples containing gas hydrates from depth is difficult because of the pressure and temperature changes encountered upon retrieval. Hydrate-bearing core samples were collected at the BPXA-DOE-USGS Mount Elbert Gas Hydrate Stratigraphic Test Well in February 2007. Coring was performed while using a custom oil-based drilling mud, and the cores were retrieved by a wireline. The samples were characterized and subsampled at the surface under ambient winter arctic conditions. Samples thought to be hydrate bearing were preserved either by immersion in liquid nitrogen (LN), or by storage under methane pressure at ambient arctic conditions, and later depressurized and immersed in LN. Eleven core samples from hydrate-bearing zones were scanned using x-ray computed tomography to examine core structure and homogeneity. Features observed include radial fractures, spalling-type fractures, and reduced density near the periphery. These features were induced during sample collection, handling, and preservation. Isotopic analysis of the methane from hydrate in an initially LN-preserved core and a pressure-preserved core indicate that secondary hydrate formation occurred throughout the pressurized core, whereas none occurred in the LN-preserved core, however no hydrate was found near the periphery of the LN-preserved core. To replicate some aspects of the preservation methods, natural and laboratory-made saturated porous media samples were frozen in a variety of ways, with radial fractures observed in some LN-frozen sands, and needle-like ice crystals forming in slowly frozen clay-rich sediments. Suggestions for hydrate-bearing core preservation are presented.

  19. RAVE code system for 3-D core non-LOCA accident analysis

    International Nuclear Information System (INIS)

    Full text of publication follows: This paper provides an overview of the application of the Westinghouse updated RAVE three dimensional (3-D) core transient analysis code system for PWR non-LOCA accident analysis. The RAVE code system consists of a linkage of the following USNRC-approved codes: the EPRI RETRAN-02 (RETRAN) system transient analysis code, the Westinghouse SPNOVA (also referred to as ANC-K) reactor core neutron kinetic nodal code, and the EPRI VIPRE-01 (VIPRE) reactor core thermal-hydraulic (T/H) code. The RETRAN code is used for calculating transient conditions in the reactor coolant system (RCS), including reactor vessel, RCS loops, pressurizer and steam generators. RETRAN also models reactor trips, engineering safety feature (ESF) functions, and the control systems. The SPNOVA code is used to perform 3-D core neutronic calculations for core average power and power distributions in the core. Its reactivity feedback calculation is based on transient fluid conditions and fuel temperatures obtained from the VIPRE code. Based on core inlet temperature, inlet flow and core exit pressure from RETRAN, and the nodal nuclear power from SPNOVA, VIPRE provides back to RETRAN transient nodal heat flux in the reactor core region. An effective 3-D analysis requires RETRAN, SPNOVA and VIPRE calculations to be closely linked for the entire reactor core. The linking architecture uses a standard external communication interface protocol for communication among the running programs on the same or different computers. The RAVE code system currently uses the Parallel Virtual Machine (PVM) software for the data transfer. Besides the necessary changes for data transfer, no other changes were made to RETRAN, SPNOVA or VIPRE fundamental code algorithms or solution methods. The RETRAN model in the RAVE system uses the same detailed reactor vessel, RCS loops, pressurizer, and steam generator, and control and protection models as has been licensed for current plant Safety

  20. Research-Based Writing Practices and the Common Core: Meta-Analysis and Meta-Synthesis

    Science.gov (United States)

    Graham, Steve; Harris, Karen R.; Santangelo, Tanya

    2015-01-01

    In order to meet writing objectives specified in the Common Core State Standards (CCSS), many teachers need to make significant changes in how writing is taught. While CCSS identified what students need to master, it did not provide guidance on how teachers are to meet these writing benchmarks. The current article presents research-supported…

  1. Verification of JUPITER standard analysis method for upgrading Joyo MK-III core design and management

    International Nuclear Information System (INIS)

    In the experimental fast reactor Joyo, loading of irradiation test rigs causes a decrease in excess reactivity because the rigs contain less fissile materials than the driver fuel. In order to carry out duty operation cycles using as many irradiation rigs as possible, it is necessary to upgrade the core performance to increase its excess reactivity and irradiation capacity. Core modification plans have been considered, such as the installation of advanced radial reflectors and reduction of the number of control rods. To implement such core modifications, it is first necessary to improve the prediction accuracy in core design and to optimize safety margins. In the present study, verification of the JUPITER fast reactor standard analysis method was conducted through a comparison between the calculated and the measured Joyo MK-III core characteristics, and it was concluded that the accuracy for a small sodium-cooled fast reactor with a hard neutron spectrum was within 5% of unity. It was shown that, the performance of the irradiation bed core could be upgraded by the improvement of the prediction accuracy of the core characteristics and optimization of safety margins. (author)

  2. Preliminary analysis of a large 1600 MWe PWR core loaded with 30% MOX fuel

    International Nuclear Information System (INIS)

    The paper presents a full-core 3-D analysis of the performances of a large 1600 MWe PWR core, loaded with 30% MOX fuel, in accordance with the European Utility Requirements (EUR). These requirements state that the European next generation power plants have to be designed capable to use MOX (UO2 - PuO2) fuel assemblies up to 50% of the core, together with UO2 fuel assemblies. The use of MOX assemblies has a significant impact on key physic parameters and on safety. A lot of studies have been carried out in the past to explore the feasibility of plutonium recycling strategies by loading LWR reactors with MOX fuel. Many of these works were based on lattice codes, in order to perform detailed analyses of the neutronic characteristics of MOX assemblies. With the aim to take into account their interaction with surrounding UO2 fuel elements, and the global effects on the core at operational conditions, an integrated approach making use of a 3-D core simulation is required. In this light, the present study adopts the state-of-art numerical models CASMO-5 and SIMULATE-3 to analyze the behavior of the core fueled with 30% MOX and to compare it with that of a large PWR reference core, fueled with UO2. (author)

  3. Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

    International Nuclear Information System (INIS)

    The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

  4. Analysis of ringing due to magnetic core materials used in pulsed nuclear magnetic resonance applications

    Science.gov (United States)

    Prabhu Gaunkar, Neelam; Nlebedim, Cajetan; Hadimani, Ravi; Bulu, Irfan; Song, Yi-Qiao; Mina, Mani; Jiles, David

    Oil-field well logging instruments employ pulsed nuclear magnetic resonance (NMR) techniques and use inductive sensors to detect and evaluate the presence of particular fluids in geological formations. Acting as both signal transmitters and receivers most inductive sensors employ magnetic cores to enhance the quality and amplitude of signals recorded during field measurements. It is observed that the magnetic core also responds to the applied input signal thereby generating a signal (`ringing') that interferes with the measurement of the signals from the target formations. This causes significant noise and receiver dead time and it is beneficial to eliminate/suppress the signals received from the magnetic core. In this work a detailed analysis of the magnetic core response and in particular loading of the sensor due to the presence of the magnetic core is presented. Pulsed NMR measurements over a frequency band of 100 kHz to 1MHz are used to determine the amplitude and linewidth of the signals acquired from different magnetic core materials. A lower signal amplitude and a higher linewidth are vital since these would correspond to minimal contributions from the magnetic core to the inductive sensor response and thus leading to minimized receiver dead time.

  5. LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Shengcheng; Wu, Hongchun; Cao, Liangzhi; Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn; Huang, Kai; He, Mingtao; Li, Xunzhao

    2014-10-15

    Highlights: • A new code system for design studies of accelerator driven subcritical reactors (ADSRs) is developed. • S{sub N} transport solver in triangular-z meshes, fine deletion analysis and multi-channel thermal-hydraulics analysis are coupled in the code. • Numerical results indicate that the code is reliable and efficient for design studies of ADSRs. - Abstract: Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional S{sub N} nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

  6. A probabilistic SSYST-3 analysis for a PWR-core during a large break LOCA

    International Nuclear Information System (INIS)

    This report demonstrates the SSYST-3 analysis and application for a German PWR of 1300 MW. The report is concerned with the probabilistic analysis of a PWR core during a loss-of-coolant accident due to a large break. With the probabilistic analysis, the distribution functions of the maximum temperatures and cladding elongations occuring in the core can be calculated. Parameters like rod power, the thermohydraulic boundary conditions, stored energy in the fuel rods and the heat transfer coefficient were found to be the most important. The expected value of core damage was determined to be 2.9% on the base of response surfaces for cladding temperature and strain deduced from SSYST-3 single rod results. (orig./HP)

  7. Safety analysis for operating the Annular Core Research Reactor with the central cavity liner removed

    International Nuclear Information System (INIS)

    Isotope production in the Annular Core Research Reactor requires highly enriched uranium targets to be irradiated in the high flux central region of the core. In order to accomplish this goal, the central cavity liner has been removed to allow for the eventual placement of targets in that region. This safety evaluation presents the analysis associated with operating the reactor in the steady state mode with the central cavity liner removed and the central region of the core filled with water and aluminum void targets. The reactor operation with enriched, uranium loaded targets will be analyzed in a future analysis document. This analysis describes only the operation of the reactor in the steady state mode; consideration of pulse mode operations with the liner removed is not presented

  8. Neutronic Analysis of HTTR Core Using DeCART Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Lee, Hyun Chul; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The neutronic analysis for the High Temperature Engineering Test Reactor (HTTR) has been performed. The HTTR is a graphite-moderated and helium gas cooled reactor with an outlet temperature of 950 .deg. C and thermal output of 30 MW. It has been studied as one of the Generation-IV (Gen-IV) reactor. In this study, the DECART code is used with a 190-group KARMA library. The calculation results are compared with those of the McCARD with ENDFB/VII.0 library.

  9. Neutronic Analysis of HTTR Core Using DeCART Code

    International Nuclear Information System (INIS)

    The neutronic analysis for the High Temperature Engineering Test Reactor (HTTR) has been performed. The HTTR is a graphite-moderated and helium gas cooled reactor with an outlet temperature of 950 .deg. C and thermal output of 30 MW. It has been studied as one of the Generation-IV (Gen-IV) reactor. In this study, the DECART code is used with a 190-group KARMA library. The calculation results are compared with those of the McCARD with ENDFB/VII.0 library

  10. Prompt Gamma Activation Analysis of the Nyírlugos obsidian core depot find

    Directory of Open Access Journals (Sweden)

    Zsolt Kasztovszky

    2014-03-01

    Full Text Available The Nyírlugos obsidian core depot find is one of the most important lithic assemblages in the collection of the Hungarian National Museum (HNM. The original set comprised 12 giant obsidian cores, of which 11 are currently on the permanent archaeological exhibition of the HNM. One of the cores is known to be inDebrecen. The first publication attributed the hoard, on the strength of giant (flint blades known from the Early and Middle Copper Age Tiszapolgár and Bodrogkeresztúr cultures, to the Copper Age. In the light of recent finds it is more likely to belong to the Middle Neolithic period. The source area was defined as Tokaj Mts., about100 kmto the NW from Nyírlugos. The size and beauty of the exceptional pieces exclude any invasive analysis. Using Prompt Gamma Activation Analysis (PGAA, we can measure major chemical components and some key trace elements of stone artefacts with adequate accuracy to successfully determine provenance of obsidian. Recent methodological development also facilitated the study of relatively large objects like the Nyírlugos cores. The cores were individually measured by PGAA. The results show that the cores originate from the Carpathian 1 sources, most probably the Viničky variety (C1b. The study of the hoard as a batch is an important contribution to the assessment of prehistoric trade and allows us to reconsider the so-called Carpathian, especially Carpathian 1 (Slovakian sources.

  11. A model of two-stage core calculation method coupled with subchannel analysis for boiling water reactors

    International Nuclear Information System (INIS)

    The two-stage core analysis method is widely used for BWR core analysis. The purpose of this study is to develop the core analysis model coupled with subchannel analysis within the two-stage calculation scheme using an assembly-based thermal-hydraulics calculation in the core analysis. This model does not change the thermal-hydraulics scheme of the core analysis. Rather, it appends the subchannel void distribution to the previous uniform analysis in lattice physics, and couples that with the subchannel analysis which axially calculates full assembly and uses the flow condition that produces the maximum void fraction in the operation core. The subchannel void distribution of one node from the subchannel analysis is only normalized and used for the lattice physics. The developed model was evaluated for the heterogeneous problem with multiple enrichments. The developed model could decrease the eigenvalue differences by more than half of that of the uniform case and made the differences of assembly power the same as the uniform case. Furthermore, it could reduce the root mean square differences to more than half of those of the uniform case in the low and high enrichment fuels. The computation times of the lattice physics become 2.3 times longer. The extended computing time does not prevent core analysis because the nuclear data are prepared in advance of the core analysis. As the result of the evaluation, the model can incorporate the subchannel effect to the core analysis. (author)

  12. Dynamical analysis of machining tool body with reinforced inner core of circular shape

    Directory of Open Access Journals (Sweden)

    Naď M.

    2009-06-01

    Full Text Available The vibration analysis of a clamped beam structure representing vibrating machining body tool is solved in this paper. The required modal properties of beam are obtained by application the reinforcing core with circular cross section. The perfect adhesion between core material and basic beam material is considered. The different material properties of beam and core are considered. The fundamental mathematical formulations describing the bending vibrations of this composite beam structure are presented. The effect of material properties and geometrical parameters of reinforcing core on natural frequencies of cantilever composite beam structure with circular and rectangular cross section is presented. This form of composite beam structure provides effective tool to modification of its dynamical properties.

  13. Computed tomography as a core analysis tool: Applications, instrument evaluation, and image improvement techniques

    International Nuclear Information System (INIS)

    In recent years, the use of computerized tomography (CT) to characterize two-phase fluid flow through porous media has become increasingly popular. This paper describes a different application of CT: it use as a core analysis tool. The advantages and disadvantages of the different technological generations of commercial medical CT scanners available as core analysis instruments are also discussed. Additionally, methods are presented for improving images and reducing CT-number errors inherent in the scanning of high-density rock samples on instruments whose software was designed for the scanning of the low-density human patients

  14. Assessing Reliability of Cellulose Hydrolysis Models to Support Biofuel Process Design – Identifiability and Uncertainty Analysis

    DEFF Research Database (Denmark)

    Sin, Gürkan; Meyer, Anne S.; Gernaey, Krist

    2010-01-01

    The reliability of cellulose hydrolysis models is studied using the NREL model. An identifiability analysis revealed that only 6 out of 26 parameters are identifiable from the available data (typical hydrolysis experiments). Attempting to identify a higher number of parameters (as done in the...

  15. Analysis on Roles for Components of Passive Emergency Core Cooling System

    International Nuclear Information System (INIS)

    International nuclear industry has been adopting a passive safety system to enhance safety and reliability of nuclear power plant with an advanced technology. Also, domestic nuclear industry issued the necessity for the development of key technologies for passive safety system design. It is necessary to develop the original technology for the improved technology, economics, and safety features. For this purpose, a Passive Emergency Core Cooling System (PECCS) is to be adopted as an improved safety design feature of APR+. When unfavorable accidents such as Station Black Out(SBO) happen, the PECCS should be able to make up the core and then cool down the core. This study discusses the applicability of PECCS and the proper design combinations especially during SBO. In this study, the applicability of PECCS and analysis on roles of components during SBO were assessed. RELAP5 calculations show that PECCS can make up the core and then prevent the core from being damaged during SBO with PAFS unavailable. Resultant analysis shows the role of the ADV for RCS depressurization, and SITs for RCS making up. When PAFS is available, ADVs is not required. Further study is required to sensitivity analysis such as actuation signal and setpoint

  16. Heterogeneous Multi Core Processors for Improving the Efficiency of Market Basket Analysis Algorithm in Data Mining

    Directory of Open Access Journals (Sweden)

    Aashiha Priyadarshni .L

    2014-09-01

    Full Text Available Heterogeneous multi core processors can offer diverse computing capabilities. The efficiency of Market Basket Analysis Algorithm can be improved with heterogeneous multi core processors. Market basket analysis algorithm utilises apriori algorithm and is one of the popular data mining algorithms which can utilise Map/Reduce framework to perform analysis. The algorithm generates association rules based on transactional data and Map/Reduce motivates to redesign and convert the existing sequential algorithms for efficiency. Hadoop is the parallel programming platform built on Hadoop Distributed File Systems(HDFS for Map/Reduce computation that process data as (key, value pairs. In Hadoop map/reduce, the sequential jobs are parallelised and the Job Tracker assigns parallel tasks to the Task Tracker. Based on single threaded or multithreaded parallel tasks in the task tracker, execution is carried out in the appropriate cores. For this, a new scheduler called MB Scheduler can be developed. Switching between the cores can be made static or dynamic. The use of heterogeneous multi core processors optimizes processing capabilities and power requirements for a processor and improves the performance of the system.

  17. Analysis on Roles for Components of Passive Emergency Core Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Soon Il; Hong, Soon Joon [FNC Tech, Yongin (Korea, Republic of); Kang, Sang Hee; Kim, Han Gon [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    International nuclear industry has been adopting a passive safety system to enhance safety and reliability of nuclear power plant with an advanced technology. Also, domestic nuclear industry issued the necessity for the development of key technologies for passive safety system design. It is necessary to develop the original technology for the improved technology, economics, and safety features. For this purpose, a Passive Emergency Core Cooling System (PECCS) is to be adopted as an improved safety design feature of APR+. When unfavorable accidents such as Station Black Out(SBO) happen, the PECCS should be able to make up the core and then cool down the core. This study discusses the applicability of PECCS and the proper design combinations especially during SBO. In this study, the applicability of PECCS and analysis on roles of components during SBO were assessed. RELAP5 calculations show that PECCS can make up the core and then prevent the core from being damaged during SBO with PAFS unavailable. Resultant analysis shows the role of the ADV for RCS depressurization, and SITs for RCS making up. When PAFS is available, ADVs is not required. Further study is required to sensitivity analysis such as actuation signal and setpoint.

  18. Analysis of core physics test data and sodium void reactivity worth calculation for MONJU core with ARCADIAN-FBR computer code system

    International Nuclear Information System (INIS)

    In order to evaluate core characteristics of fast reactors, a computer code system ARCADIAN-FBR has been developed by utilizing the existing analysis codes and the latest nuclear data library JENDL-3.3. The validity of ARCADIAN-FBR was verified by using the experimental data obtained in the MONJU core physics tests. The results of analyses are in good agreement with the experimental data and the applicability of ARCADIAN-FBR for fast reactor core analysis is confirmed. Using ARCADIAN-FBR, the sodium void reactivity worth, which is an important parameter in the safety analysis of fast reactors, was analyzed for MONJU core. 241Pu in the core fuel is transmuted to 241Am due to disintegrations. Therefore, the effect of 241Am accumulation on the sodium void reactivity worth was evaluated for MONJU core. As a result of calculation, it was confirmed that the accumulation of 241Am significantly influences on the sodium void reactivity worth and hence on the safety analysis of sodium-cooled fast reactors. (author)

  19. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    International Nuclear Information System (INIS)

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public

  20. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    Energy Technology Data Exchange (ETDEWEB)

    PARMA JR.,EDWARD J.

    2000-01-01

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

  1. Asteroseismic analysis of solar-like star KIC 6225718: constraints on stellar parameters and core overshooting

    Science.gov (United States)

    Tian, Z. J.; Bi, S. L.; Yang, W. M.; Chen, Y. Q.; Liu, Z. E.; Liu, K.; Li, T. D.; Ge, Z. S.; Yu, J.

    2014-12-01

    We analyse five seasons of short-cadence data of a solar-type star of spectral type F: KIC 6225718 observed by Kepler. We obtain the power spectrum of this star by applying the Lomb-Scargle periodogram to the smoothed time series. By applying the autocorrelation technique to the power spectrum, we derive the large-frequency separation Δν = 105.78 ± 0.65 μHz and the frequency of maximum power νmax = 2301 ± 21 μHz. We identify 33 p modes with angular degrees of l = 0-2 in the frequency range 1600-2800 μHz of the power spectrum with Bayesian Markov Chain Monte Carlo algorithms. In order to determine the parameters of the star accurately, we construct a grid of stellar models with core overshooting using the Yale stellar evolution code and then perform preliminary seismological analysis. With both asteroseismic and non-asteroseismic constraints, the following range of stellar parameters is estimated: mass M=1.10^{+0.04}_{-0.03} M_{{{odot }}}, radius R = 1.22^{+0.01}_{-0.01} R_{{{odot }}} and age t=3.35^{+0.36}_{-0.75} Gyr for this star. In addition, we analyse the effects of overshooting on stellar interiors and find that the upper limit of the overshooting parameter αov is approximately 0.2 for this star.

  2. Tank 241-AZ-102 Privatization Push Mode Core Sampling and Analysis Plan

    International Nuclear Information System (INIS)

    This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AZ-102. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AZ-102 required to satisfy the Data Quality Objectives For TWRS Privatization Phase I : Confirm Tank TIS An Appropriate Feed Source For High-Level Waste Feed Batch X(HLW DQO) (Nguyen 1999a), Data Quality Objectives For TWRS Privatization Phase 1: Confirm Tank TIS An Appropriate Feed Source For Low-Activity Waste Feed Batch X (LAW DQO) (Nguyen 1999b), Low Activity Waste and High Level Waste Feed Data Quality Objectives (L and H DQO) (Patello et al. 1999) and Characterization Data Needs for Development, Design, and Operation of Retrieval Equipment Developed through the Data Quality Objective Process (Equipment DQO) (Bloom 1996). The Tank Characterization Technical Sampling Basis document (Brown et al. 1998) indicates that these issues, except the Equipment DQO apply to tank 241-AZ-102 for this sampling event. The Equipment DQO is applied for shear strength measurements of the solids segments only. Poppiti (1999) requires additional americium-241 analyses of the sludge segments. Brown et al. (1998) also identify safety screening, regulatory issues and provision of samples to the Privatization Contractor(s) as applicable issues for this tank. However, these issues will not be addressed via this sampling event. Reynolds et al. (1999) concluded that information from previous sampling events was sufficient to satisfy the safety screening requirements for tank 241-AZ-102. Push mode core samples will be obtained from risers 15C and 24A to provide sufficient material for the chemical analyses and tests required to satisfy these data quality objectives. The 222-S Laboratory will extrude core samples, composite the liquids and solids, perform chemical analyses

  3. A method for analysis of vanillic acid in polar ice cores

    Science.gov (United States)

    Grieman, M. M.; Greaves, J.; Saltzman, E. S.

    2015-02-01

    Biomass burning generates a wide range of organic compounds that are transported via aerosols to the polar ice sheets. Vanillic acid is a product of conifer lignin combustion, which has previously been observed in laboratory and ambient biomass burning aerosols. In this study a method was developed for analysis of vanillic acid in melted polar ice core samples. Vanillic acid was chromatographically separated using reversed-phase liquid chromatography (HPLC) and detected using electrospray ionization-triple quadrupole mass spectrometry (ESI-MS/MS). Using a 100 μL injection loop and analysis time of 4 min, we obtained a detection limit of 77 ppt (parts per trillion by mass) and an analytical precision of ±10%. Measurements of vanillic acid in Arctic ice core samples from the Siberian Akademii Nauk core are shown as an example application of the method.

  4. DNBR calculation in digital core protection system by a subchannel analysis code

    International Nuclear Information System (INIS)

    The DNBR calculation uncertainty and DNBR margin were evaluated in digital core protection system by a thermal-hydrualic subchannel analysis code MATRA. A simplified thermal-hydraulic code CETOP is used to calculate on-line DNBR in core protection system at a digital PWR. The DNBR tuning process against a best-estimate subchannel analysis code is required for CETOP to ensure accurate and conservative DNBR calculation but not necessary for MATRA. The DNBR calculations by MATRA and CETOP were performed for a large number of operating condition in Yonggwang nulcear units 3-4 where the digitial core protection system is initially implemented in Korea. MATRA resulted in a less negative mean value (i.e., reduce the overconservatism) and a somewhat larger standard deviation of the DNBR error. The uncertainty corrected minimum DNBR by MATRA was shown to be higher by 1.8% -9.9% that the CETOP DNBR

  5. Analysis of Stainless Steel Sandwich Panels with a Metal Foam Core for Lightweight Fan Blade Design

    Science.gov (United States)

    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.

  6. Nonlinear dynamic analysis of prismatic elements for high-temperature gas-cooled reactor cores

    International Nuclear Information System (INIS)

    The high-temperature gas-cooled reactor (HTGR) core consists of several thousand prismatic graphite fuel elements arranged in columns within a prestressed concrete vessel. A major research and development effort was initiated in 1970 at General Atomic Company to study the dynamic response of the HTGR core arrangement to seismic excitation. A discussion is pesented of the history and some of the results of this effort with respect to the advances made in the development of analytical methods. The computer programs developed to perform the analysis are described, along with certain techniques and the modeling required to utilize them. The nonlinear dynamic analysis techniques employed to analyze the HTGR core are described

  7. CoreFlow: A computational platform for integration, analysis and modeling of complex biological data

    DEFF Research Database (Denmark)

    Pasculescu, Adrian; Schoof, Erwin; Creixell, Pau;

    2014-01-01

    the relationships between the data, the assumptions that have been made, and the manipulations that have been performed. Since the scripts use commonly available programming languages, they can easily be transferred to and from other computational environments for debugging or faster processing. This focus on ‘on...... provides programmers with a framework to manage data in real-time. It allows users to upload data into a relational database (MySQL), and to create custom scripts in high-level languages such as R, Python, or Perl for processing, correcting and modeling this data. CoreFlow organizes these scripts...... the fly’ analysis sets CoreFlow apart from other workflow applications that require wrapping of scripts into particular formats and development of specific user interfaces. Importantly, current and future releases of data analysis scripts in CoreFlow format will be of widespread benefit to the proteomics...

  8. Nuclear analysis and performance of the Light Water Breeder Reactor (LWBR) core power operation at Shippingport

    International Nuclear Information System (INIS)

    This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U233-Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred

  9. Thermal-hydraulic analysis of the MIT research reactor low enrichment uranium (LEU) Core

    International Nuclear Information System (INIS)

    The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the MIT reactor is evolving. The in-house multi-channel thermal-hydraulics code, MULCH, was developed specifically for the MITR. This code has been benchmarked against PLTEMP for steady-state analysis, and RELAP5 and temperature measurements for the loss of primary flow transient. In this paper, thermal hydraulic analyses using MULCH and RELAP5 in support of the MITR conversion tasks are described. Various fuel configurations are evaluated in order to support the LEU core design optimization study. The results show that a preferable LEU core design employs a fuel meat thickness of 20 mils with 18 plates per element with a hot channel factor less than 1.76. Simulation results also show that the LEU-fueled MITR can potentially operate at a higher power level, about 30 % higher than the current core. (authors)

  10. Analysis of heterogeneous boron dilution transients during outages with APROS 3D nodal core model

    International Nuclear Information System (INIS)

    A diluted water plug can form inside the primary coolant circuit if the coolant flow has stopped at least temporarily. The source of the clean water can be external or the fresh water can build up internally during boiling/condensing heat transfer mode, which can occur if the primary coolant inventory has decreased enough during an accident. If the flow restarts in the stagnant primary loop, the diluted water plug can enter the reactor core. During outages after the fresh fuel has been loaded and the temperature of the coolant is low, the dilution potential is the highest because the critical boron concentration is at the maximum. This paper examines the behaviour of the core as clean or diluted water plugs of different sizes enter the core during outages. The analysis were performed with the APROS 3D nodal core model of Loviisa VVER-440, which contains an own flow channel and 10 axial nodes for each fuel assembly. The widerange cross section data was calculated with CASMO-4E. According to the results, the core can withstand even large pure water plugs without fuel failures on natural circulation. The analyses emphasize the importance of the simulation of the backflows inside the core when the reactor is on natural circulation.

  11. Intestinal microbiota in healthy adults: temporal analysis reveals individual and common core and relation to intestinal symptoms.

    Directory of Open Access Journals (Sweden)

    Jonna Jalanka-Tuovinen

    Full Text Available BACKGROUND: While our knowledge of the intestinal microbiota during disease is accumulating, basic information of the microbiota in healthy subjects is still scarce. The aim of this study was to characterize the intestinal microbiota of healthy adults and specifically address its temporal stability, core microbiota and relation with intestinal symptoms. We carried out a longitudinal study by following a set of 15 healthy Finnish subjects for seven weeks and regularly assessed their intestinal bacteria and archaea with the Human Intestinal Tract (HIT Chip, a phylogenetic microarray, in conjunction with qPCR analyses. The health perception and occurrence of intestinal symptoms was recorded by questionnaire at each sampling point. PRINCIPAL FINDINGS: A high overall temporal stability of the microbiota was observed. Five subjects showed transient microbiota destabilization, which correlated not only with the intake of antibiotics but also with overseas travelling and temporary illness, expanding the hitherto known factors affecting the intestinal microbiota. We identified significant correlations between the microbiota and common intestinal symptoms, including abdominal pain and bloating. The most striking finding was the inverse correlation between Bifidobacteria and abdominal pain: subjects who experienced pain had over five-fold less Bifidobacteria compared to those without pain. Finally, a novel computational approach was used to define the common core microbiota, highlighting the role of the analysis depth in finding the phylogenetic core and estimating its size. The in-depth analysis suggested that we share a substantial number of our intestinal phylotypes but as they represent highly variable proportions of the total community, many of them often remain undetected. CONCLUSIONS/SIGNIFICANCE: A global and high-resolution microbiota analysis was carried out to determine the temporal stability, the associations with intestinal symptoms, and the

  12. Core competence and dominant logic: contributions to the analysis of merger and acquisition process

    Directory of Open Access Journals (Sweden)

    Marcelo Pereira Binder

    2010-01-01

    Full Text Available Mergers and acquisitions are one of the most important strategic decisions a company can take. In the 1980s and 1990s, mergers and acquisitions have occurred in large numbers of companies and several theories have been developed to explain the phenomenon. However, most of these theories are related to the financial area. But, non-quantifiable aspects, such as core competencies and dominant logic, have been relegated to the background. Identified this gap, this paper proposes the inclusion of the concept of core competence and dominant logic as an analytical tool to validate a merger process. To do so, this article has rescued the discussion of these concepts in the business strategy field and a case that did not achieve the proposed goals with the merger was examined from this perspective. The result evidence that the adoption of the concepts of core competence and dominant logic helps explain a new insights with the merger process.

  13. Core analysis of the first cycle of Chashma nuclear power plant

    International Nuclear Information System (INIS)

    The up coming 300 MWe CHASHMA NPP will provide the opportunity to study the burn-up behavior of the fuel. Our experience is limited to the in-core fuel management studies when fuel burn-up remains within the design limits. The initial core is loaded in three regions with fuel of three different enrichments 2.4 w/o, 2.67 w/o and 3.0 w/o. It is intended to study the enhanced fuel burn-up vis-a-vis the expected cost benefit in due course of time. The core of the Chashma nuclear power plant is that of a typical PWR NPP of 300 MWe capacity. It has 121 fuel assemblies and all of them have identical external dimensions and hydraulic characteristics. The core height is 290 cm and equivalent diameter is 248·6 cm. The core is cooled and moderated by H2O and surrounded by a stainless steel baffle. Each fuel assembly consists of 15x15 rod array and the assembly pitch is 20·03 cm The average discharge burn-up is 30,000 MWd/MTU. Core analysis was carried out for the first cycle at hot full power (HFP). Two dimensional calculations were performed for burn-up analysis including core multiplication, flux distribution, burn-up length, isotopic inventory, peaking factor and critical boron concentration to achieve the economical fuel management within the constraints imposed by safe reactor operation. Calculations indicate that expected burn-up of the first cycle is 13479 MWd/MTU equivalent to 485 EFPD, with 25 ppm of boron is still in the system, which is very near to the design value. Similarly assembly power distribution, pin by pin power distribution and reactivity coefficients, calculated are within the acceptable limits. Efforts are on to improve further these calculations. (author)

  14. Computational Design and Analysis of Core Material of Single-Phase Capacitor Run Induction Motor

    Directory of Open Access Journals (Sweden)

    Gurmeet Singh

    2014-07-01

    Full Text Available A Single-phase induction motor (SPIM has very crucial role in industrial, domestic and commercial sectors. So, the efficient SPIM is a foremost requirement of today's market. For efficient motors, many research methodologies and propositions have been given by researchers in past. Various parameters like as stator/rotor slot variation, size and shape of stator/rotor slots, stator/rotor winding configuration, choice of core material etc. have momentous impact on machine design. Core material influences the motor performance to a degree. Magnetic flux linkage and leakage preliminary depends upon the magnetic properties of core material and air gap. The analysis of effects of core material on the magnetic flux distribution and the performance of induction motor is of immense importance to meet out the desirable performance. An increase in the air gap length will result in the air gap performance characteristics deterioration and decrease in air gap length will lead to serious mechanical balancing concern. So possibility of much variation in air gap beyond the limits on both sides is not feasible. For the optimized performance of the induction motor the core material plays a significant role. Using higher magnetic flux density, reduction on a magnetizing reactance and leakage of flux can be achieved. In this thesis work the analysis of single phase induction motor has been carried out with different core materials. The four models have been simulated using Ansys Maxwell 15.0. Higher flux density selection for same machine dimensions result into huge amount of reduction in iron core losses and thereby improve the efficiency. In this paper 2% higher efficiency has been achieved with Steel_1010 as compared to the machine using conventional D23 material. Out of four models result reflected by the machine using steel_1010 and steel_1008 are found to be better.

  15. Innovative research reactor core designed. Estimation and analysis of gamma heating distribution

    International Nuclear Information System (INIS)

    The Gamma heating value is an important factor needed for safety analysis of each experiments that will be realized on research reactor core. Gamma heat is internal heat source occurs in each irradiation facilities or any material irradiated in reactor core. This value should be determined correctly because of the safety related problems. The gamma heating value is in general depend on. reactor core characteristics, different one and other, and then each new reactor design should be completed by gamma heating data. The Innovative Research Reactor is one of the new reactor design that should be completed with any safety data, including the gamma heating value. For this reasons, calculation and analysis of gamma heating in the hole of reactor core and irradiation facilities in reflector had been done by using of modified and validated Gamset computer code. The result shown that gamma heating value of 11.75 W/g is the highest value at the center of reactor core, higher than gamma heating value of RSG-GAS. However, placement of all irradiation facilities in reflector show that safety characteristics for irradiation facilities of innovative research reactor more better than RSG-GAS reactor. Regarding the results obtained, and based on placement of irradiation facilities in reflector, can be concluded that innovative research reactor more safe for any irradiation used. (author)

  16. Degraded core accidents for the Sizewell PWR: A sensitivity analysis of the radiological consequences

    International Nuclear Information System (INIS)

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, particular to the releases analysed from Sizewell; for different releases from different locations the sensitivity may change significantly. In the earlier study and analysis was undertaken of the impact on the predicted consequences of potential overestimates in the release fractions of radionuclides. Since the results of that study were published some relatively minor numerical errors have been identified. While none of these affects the conclusions reached in that study the opportunity has been taken in this report to present revised values for those results known to be in error. This revised text and results are presented as an appendix to this report and they replace the corresponding material in the earlier study. (author)

  17. LMR core thermal-hydraulic analysis accounting for interassembly heat transfer =

    International Nuclear Information System (INIS)

    In is essential to have an accurate prediction of core coolant and fuel temperature distribution in the liquid metal reactor (LMR) core themal hydraulic design, because of the design limits are imposed on the maximum temperatures of claddings and fuel pins in the sodium cooled LMRs. Due to the high thermal conductivity of the sodium, the transverse interassembly heat transfer may have a significant effect on the temperature profile within the subassembly, especially when it is adjacent to considerably hotter or colder subassemblies. Therefore, the interassembly heat transfer calculation should be considered in the LMR core thermal hydraulic design and analysis. For multi-assembly analysis, the interassembly heat transfe model was added in the MATRA-LMR code and the code was extended a single assembly analysis to multi-assembly analysis, i. e., a whole core code. For the assessment of the development status with interassembly heat transfer, the benchmark calculations were performed with SLTHE and THI3D codes using the 7-assembly problems. It is founded that the subassembly mixed mean coolant temperature has been changed as an effect of the interassembly heat transfer. And the maximum temperature change was found in the non-fueled subassembly which is considerabl colder than the fueled subassemblies

  18. Microscopic analysis on showers recorded as single core on X-ray films

    International Nuclear Information System (INIS)

    Cosmic-ray particles recorded as single dark spots on X-ray films with use of the emulsion chamber data of Brazil-Japan Collaboration are studied. Some results of microscopic analysis of such single-core-like showers on nuclear emulsion plates are reported. (Author)

  19. A study on the core analysis methodology for SMART CEA ejection accident-I

    International Nuclear Information System (INIS)

    A methodology to analyze the fuel enthalpy is developed based on MASTER that is a time dependent 3 dimensional core analysis code. Using the proposed methodology, SMART CEA ejection accident is analyzed. Moreover, radiation doses are estimated at the exclusion area boundary and low population zone to confirm the criteria for the accident. (Author). 31 refs., 13 tabs., 18 figs

  20. Modeling Overlapping Laminations in Magnetic Core Materials Using 2-D Finite-Element Analysis

    DEFF Research Database (Denmark)

    Jensen, Bogi Bech; Guest, Emerson David; Mecrow, Barrie C.

    2015-01-01

    This paper describes a technique for modeling overlapping laminations in magnetic core materials using two-dimensional finite-element (2-D FE) analysis. The magnetizing characteristic of the overlapping region is captured using a simple 2-D FE model of the periodic overlapping geometry and a comp...

  1. On the Paleostress Analysis Using Kinematic Indicators Found on an Oriented Core

    Czech Academy of Sciences Publication Activity Database

    Nováková, Lucie; Brož, Milan

    2014-01-01

    Roč. 2, č. 2 (2014), s. 76-83. ISSN 2331-9593 R&D Projects: GA MPO(CZ) FR-TI1/367 Institutional support: RVO:67985891 Keywords : paleostress analysis * borehole core * kinematic indicators * bias sampling * recent stress Subject RIV: DC - Siesmology, Volcanology, Earth Structure http://www.hrpub.org/download/20140105/UJG6-13901884.pdf

  2. Core group approach to identify college students at risk for sexually transmitted infections "Core group" para identificar universitários em risco para infecções sexualmente transmissíveis

    Directory of Open Access Journals (Sweden)

    Miguel A Sánchez-Alemán

    2008-06-01

    Full Text Available OBJECTIVE: To analyze the core group for sexually transmitted infections (STI among college students. METHODS: Cross-sectional study carried out in a convenience sample comprising 711 college students of the public university of Morelos, Mexico, between 2001 and 2003. Sociodemographic and sexual behavior information were collected using self-applied questionnaires. Herpes simplex 2 (HSV-2 infection was tested in the blood. The number of sexual partners in the last year and cocaine consumption were used as indicators to construct the dependent variable "level of STI risk" in three categories: low, medium and high risk (core group. A multinomial analysis was conducted to evaluate whether different sex behaviors were associated with the variable "level of STI risk". RESULTS: There was significant association between HSV-2 seroprevalence and the variable "level of STI risk": 13%, 5.6% and 3.8% were found in high (core group, medium and low categories, respectively. There were gender differences regarding the core group. Men started having sexual intercourse earlier, had more sex partners, higher alcohol and drug consumption, higher frequency of sex intercourse with sex workers, exchanging sex for money, occasional and concurrent partners compared to women. CONCLUSIONS: The study findings suggest existing contextual characteristics in the study population that affect their sex behavior. In Mexico, the cultural conception of sexuality is determined mainly by gender differences where men engage in higher risky sexual behavior than women.OBJETIVO: Identificar al grupo core de infecciones de transmisión sexual (ITS en una población de estudiantes universitarios mexicanos. MÉTODOS: Se realizó un estudio transversal en una muestra por conveniencia que incluyó 711 estudiantes de una universidad pública de Morelos, México, entre 2001 y 2003. Las características sociodemográficas y de comportamiento sexual se obtuvieron mediante un cuestionario auto

  3. Tank 241-AZ-102 Privatization Push Mode Core Sampling and Analysis Plan

    International Nuclear Information System (INIS)

    This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AZ-102

  4. Hydraulic analysis of the emergency core cooling system of the RP-10 reactor

    International Nuclear Information System (INIS)

    This work shows calculation for the hydraulic analysis of the Emergency Core Cooling System (ECCS) of the RP-10 Reactor. This analysis is necessary for the design of such system. According to calculation results shown in the graphics, a pipe line of two inches of nominal diameter should be selected for such system and a maximum flow of 5 m3/h should be reached

  5. Core seismic analysis SNR-2 preliminary design and R and D program

    International Nuclear Information System (INIS)

    This paper reviews the work done at Belgonucleaire in the framework of the seismic analysis of FBR cores and subassemblies. The main features of the computer programme CLASH are first reviewed. Next, the main conclusions of the preliminary analysis of SNR-2 are examined. Finally, the objectives of future research are outlined and the status of the R and D program at Belgonucleaire is given. (author). 8 refs

  6. Optimization of High-Resolution Continuous Flow Analysis for Transient Climate Signals in Ice Cores

    DEFF Research Database (Denmark)

    Bigler, Matthias; Svensson, Anders; Kettner, Ernesto;

    2011-01-01

    Over the past two decades, continuous flow analysis (CFA) systems have been refined and widely used to measure aerosol constituents in polar and alpine ice cores in very high-depth resolution. Here we present a newly designed system consisting of sodium, ammonium, dust particles, and electrolytic...... meltwater conductivity detection modules. The system is optimized for high- resolution determination of transient signals in thin layers of deep polar ice cores. Based on standard measurements and by comparing sections of early Holocene and glacial ice from Greenland, we find that the new system features a...

  7. Consistency evaluation of JUPITER experiment and analysis for large FBR cores

    International Nuclear Information System (INIS)

    A series of critical experiments for study of large FBR cores, JUPITER, was analyzed with the latest analytical methods. These results were evaluated from various physical viewpoints by means of comparison with other cores or other nuclear characteristics by full use of sensitivity analysis, effect of different nuclear data libraries and application of most-detailed analytical tools. It is concluded that the JUPITER experiments and analytical results possess sufficient consistency on the whole, though there is some room for further improvements. The proper use of JUPITER data will enhance the accuracy and reliability of design work for the large FBRs. (author)

  8. Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core, OFI (onset of flow instability) and DNB (departure from nucleate boiling)

  9. Research reactor core conversion guidebook. V.2: Analysis (Appendices A-F)

    International Nuclear Information System (INIS)

    Volume 2 consists of detailed Appendices, covering safety analyses for generic 10 MW reactor, safety analysis - probabilistic methods, methods for preventing LOCA, radiological consequence analyses, examples of safety report amendments and safety specifications. Included in Volume 2 are example analyses for cores with with highly enriched uranium and low enriched uranium fuels showing differences that can be expected in the safety parameters and radiological consequences of postulated accidents. There are seven examples of licensing documents related to core conversion and two examples of methods for determining power limits for safety specifications in the document. Refs, figs, bibliographies and tabs

  10. Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, I.H. [SDTP User Group, Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore 45650, Islamabad (Pakistan)], E-mail: ihbokhari@yahoo.co.uk; Mahmood, T.; Chaudri, K.S. [SDTP User Group, Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore 45650, Islamabad (Pakistan)

    2007-10-15

    Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core, OFI (onset of flow instability) and DNB (departure from nucleate boiling)

  11. Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2015-01-01

    Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

  12. Development and assessment of a subchannel analysis code system for SMART core design

    International Nuclear Information System (INIS)

    A subchannel code system is developed for the thermal-hydraulic analysis of SMART core, and the applicability and accuracy of the code is assessed for various experimental data with rod bundles. MATRA is a subchannel analysis code calculating the enthalpy and flow distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. MATRA has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-IV-I. MATRA has been provided with an improved structure and code functions to give more convenient user environment. Improvement of various models enhances the convergence and accuracy of the code: those include the numerical solution scheme for the crossflow, the void fraction model, and the lateral transport model, and so on. A turbulent mixing model considering void drift phenomenon is devised by employing the two-phase mixing test data under PWR and BWR conditions. MATRA/SR-1 CHF correlation system is developed from local conditions of rod bundle CHF data calculated by MATRA. The optimized 1/8 core lumping models are developed for the analysis of the thermal margins of SMART core at steady-state and transient conditions

  13. Analysis on criticality properties and control strategies after reflooding of a damaged reactor core

    International Nuclear Information System (INIS)

    Highlights: • Criticality analyses based on a severe accident condition including a core meltdown. • The critical mass assessment of corium within a core. • The analysis of criticality with the change of molten level of corium. • The change of coolant void reactivity (CVR) following a core reflood. • The establishment of practical strategies for a criticality control using H3BO3. - Abstract: In order to evaluate the crucial physical phenomena in a re-criticality of the reactor corium following a core reflood, a variety of criticality analyses have been performed with the design of a hypothetical core based on a severe accident condition including a core melt-down. This study aims to assess the critical mass of the corium within the core, to analyze the criticality change with the molten level and the coolant void reactivity (CVR), and to establish of practical strategies for a criticality control using H3BO3. The MCNPX 2.5.0 code was used to design the hypothetical accident core and to calculate the effective multiplication factor (keff), and the US Peach Bottom and Japan Fukushima nuclear power plant (NPP) was chosen for a reference model. The critical level was set to 0.95 for a conservative evaluation, and the consequential critical mass was indicated to be about 60 kg only. The 548 fuel assemblies loaded in the core were assumed to be molten up to approximately 77%, the keff value therefore has a range of 1.03241 ± 0.00194 to 1.40801 ± 0.00157. The negative reactivity was increasingly inserted as the coolant density was decreased from 95% to 0%, and the keff value approaches the sub-critical state when the void fraction is above 30%. Finally, according to the boron injection into water, it was found that at least 600 ppm 10B is required to have the sufficient criticality safety and to assure the sub-criticality condition following a severe core damage accident

  14. Uncertainty analysis for the BEACON-COLSS core monitoring system application

    International Nuclear Information System (INIS)

    This paper will cover the measurement uncertainty analysis of BEACON-COLSS core monitoring system. The uncertainty evaluation is made by using a BEACON-COLSS simulation program. By simulating the BEACON on-line operation for analytically generated reactor conditions, accuracy of the 'Measured' results can be evaluated by comparing to analytically generated 'Truth'. The DNB power margin is evaluated based on the Combustion Engineering's Modified Statistical Combination of Uncertainties (MSCU) using the CETOPD code for the DNBR calculation. A BEACON-COLSS simulation program for the uncertainty evaluation function has been established for plant applications. Qualification work has been completed for two Combustion Engineering plants. Results of the BEACON-COLSS measured peaking factors and DNBR power margin are plant type dependent and are applicable to reload cores as long as the core geometry and detector layout are unchanged. (authors)

  15. High Level Analysis, Design and Validation of Distributed Mobile Systems with CoreASM

    Science.gov (United States)

    Farahbod, R.; Glässer, U.; Jackson, P. J.; Vajihollahi, M.

    System design is a creative activity calling for abstract models that facilitate reasoning about the key system attributes (desired requirements and resulting properties) so as to ensure these attributes are properly established prior to actually building a system. We explore here the practical side of using the abstract state machine (ASM) formalism in combination with the CoreASM open source tool environment for high-level design and experimental validation of complex distributed systems. Emphasizing the early phases of the design process, a guiding principle is to support freedom of experimentation by minimizing the need for encoding. CoreASM has been developed and tested building on a broad scope of applications, spanning computational criminology, maritime surveillance and situation analysis. We critically reexamine here the CoreASM project in light of three different application scenarios.

  16. CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

    International Nuclear Information System (INIS)

    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 that simulates a control-rod withdrawal at full power

  17. Precision measurement system and analysis of low core signal loss in DCF couplers

    Science.gov (United States)

    Yan, P.; Wang, X. J.; Fu, Ch; Li, D.; Sun, J. Y.; Gong, M. L.; Xiao, Q. R.

    2016-07-01

    In order to achieve higher output power of double cladding fiber lasers, low signal loss has become a focus in researches on optical technology, especially double-clad fiber (DCF) couplers. According to the analysis, DCF couplers with low core signal loss (less than 1%) are produced. To obtain higher precision, we use the first-proposed method for core signal transfer efficiency measurement based on the fiber propagation field image processing. To the best of our knowledge, we report, for the first time, the results of the core signal loss less than 1% in DCF coupler measured by our measurement with high stability and relative precision. The measurement values can assess the quality of DCF couplers and be used as a signal to suggest the improvement on the processing technology of our self-made DCF couplers.

  18. Statistical Analysis of ITRAX XRF Data to Identify Marine Incursion, Sediment Source, and Saltwater Leaching in Tsunami Deposits

    Science.gov (United States)

    Kain, C. L.; Chague-Goff, C.; Goff, J. R.; Gadd, P.

    2015-12-01

    Geochemical investigation of fine-grained tsunami sediments has found that a characteristic salinity signature can be commonly found in tsunami deposits and underlying soils following an event. We extend this method to assess historic and paleotsunami deposits and investigate a wider range of particle sizes, with the aim of identifying whether salinity signatures are present and determining the source material of the deposits. Geochemical and mineralogical investigation of seven short cores was undertaken at four sites in New Zealand, where historic and/or palaeotsunami deposits were present as sand, silt or gravel layers intercalated between soils. Geochemical signatures were measured using a high-resolution ITRAX X-ray fluorescence (XRF) core scanner and results were analysed using Principal Component Analysis (PCA) and Hierarchical Cluster Analysis (HCA). We were able to explore the signatures of historical events, verify that prehistoric deposits were laid down by tsunamis, and compare these sediments with the background depositional environment. A t-test of means was applied for salinity marker elements (S, Cl, Br) in the soils immediately above and below tsunami sand layers, to test for evidence of saltwater leaching. The dominant mineralogy of the sediments was determined using portable X-ray diffraction and the data used to consider source material and interpret the corresponding XRF data. Geochemical signatures were found to be site specific, depending primarily on the composition of the material. PCA and HCA results clearly distinguished the signature of the tsunami deposits from the background material at each individual site and were able to confirm or deny palaeodeposits as tsunami-related, by comparison with the signatures of known events in the same core or nearby.

  19. Water-quality trends in White Rock Creek Basin from 1912-1994 identified using sediment cores from White Rock Lake Reservoir, Dallas, Texas

    Science.gov (United States)

    Van Metre, P.C.; Callender, E.

    1997-01-01

    Historical trends in selected water-quality variables from 1912 to 1994 in White Rock Creek Basin were identified by dated sediment cores from White Rock Lake. White Rock Lake is a 4.4-km2 reservoir filled in 1912 and located on the north side of Dallas, Texas, with a drainage area of 259 km2. Agriculture dominated land use in White Rock Creek Basin before about 1950. By 1990, 72% of the basin was urban. Sediment cores were dated using cesium-137 and core lithology. Major element concentrations changed, and sedimentation rates and percentage of clay-sized particles in sediments decreased beginning in about 1952 in response to the change in land use. Lead concentrations, normalized with respect to aluminum, were six times larger in sediment deposited in about 1978 than in pre-1952 sediment. Following the introduction of unleaded gasoline in the 1970s, normalized lead concentrations in sediment declined and stabilized at about two and one-half times the pre-1952 level. Normalized zinc and arsenic concentrations increased 66 and 76%, respectively, from before 1952 to 1994. No organochlorine compounds were detected in sediments deposited prior to about 1940. Concentrations of polychlorinated biphenyls (PCB) and DDE (a metabolite of DDT) increased rapidly beginning in the 1940s and peaked in the 1960s at 21 and 20 ??g kg-1, respectively, which is coincident with their peak use in the United States. Concentrations of both declined about an order of magnitude from the 1960s to the 1990s to 3.0 and 2.0 ??g kg-1, respectively. Chlordane and dieldrin concentrations increased during the 1970s and 1980s. The largest chlordane concentration was 8.0 ??g kg-1 and occurred in a sediment sample deposited in about 1990. The largest dieldrin concentration was 0.7 ??g kg-1 and occurred in the most recent sample deposited in the early 1990s. Agricultural use of chlordane and dieldrin was restricted in the 1970s; however, both were used as termiticides, and urban use of chlordane

  20. Low-background neutron activation analysis. A powerful tool for atmospheric mineral dust analysis in ice cores

    International Nuclear Information System (INIS)

    The application of instrumental neutron activation analysis (INAA) for multi-elemental analysis of samples of extremely reduced mass such as dust samples extracted from ice cores requires specific efforts towards the development of a 'low level counting' analytical technique. An analytical protocol specifically designed for this kind of samples, based on low-background INAA (LBNAA) is here presented. A first application of the method was successfully performed on samples from the new alpine firn core NextData-LYS12. Sub-ng detection limits were reached for many elements. According to this point the technique is also potentially suitable to be applied to polar ice core samples. (author)

  1. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  2. PWR core and spent fuel pool analysis using scale and nestle

    International Nuclear Information System (INIS)

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  3. Improvement of core effective thermal conductivity model of GAMMA+ code based on CFD analysis

    International Nuclear Information System (INIS)

    Highlights: • We assessed the core effective thermal conductivity (ETC) model of GAMMA+ code. • The analytical model of GAMMA+ code was compared with the result of CFD analysis. • Effects of material property of composite and geometric configuration were studied. • The GAMMA+ model agreed with the CFD result when the fuel gap is ignored. • The GAMMA+ model was improved by the ETC model of fuel compact including fuel gap. - Abstract: The GAMMA+ code has been developed for the thermo-fluid and safety analyses of a high temperature gas-cooled reactor (HTGR). In order to calculate the core effective thermal conductivity, this code adopts a heterogeneous model derived from the Maxwell’s theory that accounts for three distinct materials in a fuel block of the reactor core. In this model, the fuel gap is neglected since the gap thickness is quite small. In addition, the configuration of the fuel block is assumed to be homogeneous, and the volume fraction and material properties of each component are taken into account. In the accident condition, the conduction and radiation are major heat transfer mechanism. Therefore, the core effective thermal conductivity model should be validated in order to estimate the heat transfer in the core appropriately. In this regard, the objective of this study is to validate the core effective thermal conductivity model of the GAMMA+ code by a computational fluid dynamics (CFD) analysis using a commercial CFD code, CFX-13. The effects of the temperature condition, material property and geometric modeling on the core effective thermal conductivity were investigated. When the fuel gap is not modeled in the CFD analysis, the result of the GAMMA+ code shows a good agreement with the CFD result. However, when the fuel gap is modeled, the GAMMA+ model overestimates the core effective thermal conductivity considerably for all cases. This is because of the increased thermal resistance by the fuel gap which is not taken into account in

  4. Toroidal superconducting transformer with cold magnetic core – results of analysis and measurements

    International Nuclear Information System (INIS)

    The paper is focused on a toroidal superconducting transformer with cold magnetic core. The transformer was developed aiming at the solution where magnetic core could operate immersed in LN2, not having much higher power losses than a core operating in room temperature. The second aim is concerned to the perpendicular component of the magnetic flux. It was assumed that it should be minimized. The third aim is required maximal coupling coefficient between windings. Magnetic material METGLAS 2605 SA1 (Metglas Ltd) was selected as the best taking electrical and mechanical properties into account. Numerical analysis of the transformer was carried using ANSYS software. It was assumed that magnetic core could be put inside windings and/or outside of it. The analysis and measurements yield that the best solution is the transformer with magnetic core put inside the windings. Such a construction reduces significantly perpendicular component of the magnetic field in HTS wire. Because outer winding is put on the inner one it gives maximal possible coupling coefficient. The primary and the secondary have the same number of turns. It means bifilar way of winding. The described transformer has the following constructional data: i) diameter of the main circle: 144 mm, ii) maximal diameter: 208 mm, iii) diameter of the small cross-section of the transformer (without external magnetic core): 50 mm and it results with the following parameters: output power 1.59 kVA, power density 700 VA/kg, efficiency: 99.5, coupling coefficient 0.99 at 100 ARMS of input current and maximum flux density 1.2 T.

  5. Determination of heavy metals and other elements by Neutron Activation Analysis in sediment cores of laguna Mar Chiquita (Cordoba, Argentina)

    International Nuclear Information System (INIS)

    Laguna Mar Chiquita is one of the largest water bodies of South America. It is a big lake of saline waters, and its geographic localization is SOMS' S, 62 deg 30' W, about 150 km Northwest of Cordoba, Argentina. Due to its large variability of hydrological budget, surface and water levels produced periods of low stands (LLP) and high stands (HLP). This fluctuation of water level also produces substantial changes in the water salinity. The principal tributary of the Laguna Mar Chiquita is Dulce River and also receives water from two other rivers: Suquia and Xanaes. The Suquia River drains in a small satellite lake, the Laguna del Plata. The purpose of the present work was to investigate the sediment composition of the Laguna Mar Chiquita (2 sediment cores) and del Plata (one sediment core) by using Neutron Activation Analysis (NAA) technique. The three 60 cm long sediment cores, sliced each 2 cm, were analyzed and 26 elements were determined (As, Ba, Br, Ce, Co, Cr, Cs, Eu, Fe, Hf, La, Lu, Na, Nd, Rb, Sb, Sc, Se, Sm, Ta, Tb, Th, U, Yb, Zn e Zr). Other complementary techniques were utilized: macro elements (Al, Ca, Fe, K, Mg, Mn, Na, P, Si, Ti) were determined by X-ray fluorescence, and the mineralogical composition of the sediments was determined by X-ray diffraction. The results obtained did not show any indication of anthropic contribution in the sediment composition, and concentration of the majority of the elements analyzed is uniform along the entire profile of the analyzed core. Statistical analysis of elemental concentrations (Cluster Analysis) reflects that in Laguna del Plata the fine fractions of the sediments is dominated by detrital minerals, while in the Laguna Mar Chiquita, the neo formed minerals are the principal components. In both lakes, it was possible to identify compositional variations in the sediment segments, which may correspond to temporal fluctuations in the sedimentation conditions. The statistical analysis associated to sedimentation

  6. Analysis of reaction rate and sample worth measured in simulated LMFBR meltdown cores

    International Nuclear Information System (INIS)

    An analysis of fission rate, fission rate ratio and sample worth has been made on FCA VIII-2 Assembly built to simulate LMFBR meltdown cores. To check the adoptability of computational methods used in analysing core disruptive accidents, the results obtained by the three methods, transport (Sn), conventional and modified diffusion methods, were compared with the measured ones. Group constants used in the analysis were prepared from JAERI Fast Set Version II. Conventional diffusion theory does not represent the measured fission rates both in the compacted and voided region. The results obtained by transport theory with S4P0 approximation agree fairly well with the measured ones. The use of modified diffusion theory, which changes the diffusion coefficient in the voided region does not significantly improve the agreement. For Pu sample worth, problems still remain partially which need a more detailed analysis, although the use of transport theory improves the agreement to a large extent. (author)

  7. Modeling and analysis of neutron noise from an ex-core detector at a pressurized water reactor

    International Nuclear Information System (INIS)

    Two applications of a noise diagnostic methodology were performed using ex-core neutron detector data from a pressurized water reactor (PWR). A feedback dynamics model of the neutron power spectral density (PSD) was derived from a low-order whole-plant physical model made stochastic using the Langevin technique. From a functional fit to plant data, the response of the dynamic system to changes in important physical parameters was evaluated by a direct sensitivity analysis. In addition, changes in monitored spectra were related to changes in physical parameters and detection thresholds using common surveillance discriminants were determined. A resonance model was developed from perturbation theory to give the ex-core neutron detector response for small in-core mechanical motions in terms of a pole-strength factor, a resonance asymmetry (or skewness) factor, a vibration damping factor, and a frequency of vibration. The mechanical motion parameters for several resonances were determined by a functional fit of the model to plant data taken at various times during a fuel cycle and were tracked to determine trends that indicated vibrational changes of reactor internals. In addition, the resonance model gave the ability to separate the resonant components of the PSD after the parameters had been identified. As a result, the behavior of several vibration peaks were monitored over a fuel cycle. 9 refs., 6 figs., 1 tab

  8. Analysis of severe hypothetical accidents for the SNR-300 MARK 1A core

    International Nuclear Information System (INIS)

    Two types of hypothetical accidents have been analysed for the fresh and irradiated SNR-300 MARK 1A cores: 1) Loss of flow accidents caused by a coast down of all primary pumps and simultaneous failure of both independent shutdown systems. 2) Transient overpower accidents caused by a reactivity input ramp and simultaneous failure of both independent shutdown systems. The analysis was done by using the CAPRI-2/KADIS computer-program-system which was developed at the Nuclear Research Center Karlsruhe. Detailed parametric variations were performed for both accident types in case of the fresh core to determine the most important and influential parameters. These parameter studies were also used to allow a conservative parameter choice for the reference cases within a reasonable parameter band. In addition the parametric variations gave some insight under which circumstances the accident would lead into energetic disassembly, early shutdown with in-place cooling possibility, or a transition phase with extended fuel motion. The thermal energy in the molten fuel at the end of the nuclear excursion is one important quantity for the severity of the accident. In case of the loss of flow accidents these energies were 3,239 MWs and 3,605 MWs for the reference cases of the fresh and irradiated cores, respectively. The corresponding energies for the reference transient overpower accidents caused by a 15 cent/sec reactivity input ramp were 1,182 MWs and 2,940 MWs for the fresh and irradiated cores respectively. Besides the thermal energy release the analysis provides much more information, for example, about the core conditions at the end of the nuclear excursion. More important the analysis automatically gives the input data for the computer programs which analyse the mechanical response of the reactor tank and the tankinternal mechanical structures. (orig./HP)

  9. Dating a tropical ice core by time-frequency analysis of ion concentration depth profiles

    Science.gov (United States)

    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.

  10. Experimental investigation and CFD analysis on cross flow in the core of PMR200

    International Nuclear Information System (INIS)

    Highlights: • An experimental facility for the cross flow in the core of PMR200 was constructed. • The cross flow experimental data were produced with wedge-shaped gap and parallel gap. • The experimental results were compared with CFD results. • The results of the CFD analysis and experimental data are in good agreement. • The pressure loss coefficient for the cross gap of PMR200 was derived. - Abstract: The Prismatic Modular Reactor (PMR) is one of the major Very High Temperature Reactor (VHTR) concepts, which consists of hexagonal prismatic fuel blocks and reflector blocks made of nuclear grade graphite. However, the shape of the graphite blocks could be easily changed by neutron damage during the reactor operation and the shape change can create gaps between the blocks inducing the bypass flow. In the VHTR core, two types of gaps, a vertical gap and a horizontal gap which are called bypass gap and cross gap, respectively, can be formed. The cross gap complicates the flow field in the reactor core by connecting the coolant channel to the bypass gap and it could lead to a loss of effective coolant flow in the fuel blocks. Thus, a cross flow experimental facility was constructed to investigate the cross flow phenomena in the core of the VHTR and a series of experiments were carried out under varying flow rates and gap sizes. The results of the experiments were compared with CFD (Computational Fluid Dynamics) analysis results in order to verify its prediction capability for the cross flow phenomena. Fairly good agreement was seen between experimental results and CFD predictions and the local characteristics of the cross flow was discussed in detail. Based on the calculation results, pressure loss coefficient across the cross gap was evaluated, which is necessary for the thermo-fluid analysis of the VHTR core using a lumped parameter code

  11. Analysis of core-melt states for the development of detection methods for filling level change and deformation of the core in PWR-type reactors

    International Nuclear Information System (INIS)

    The project ''noninvasive status monitoring of nuclear reactors for detection of filling level changes and core deformation'' (NIZUK) is aimed to develop a measuring system for the core status diagnosis during severe accidents in PWR-type reactors. For the development of an appropriate measuring technology the knowledge on the processes during the in-vessel phase of the accident sequence is of main importance. Using the analysis of the accident sequence nine in-vessel phases were defined that are the basis for the development of the measuring system. The differences between the individual core-melt states include the different core geometries and a varying gamma radiation distribution at the reactor pressure vessel outer surface. Especially the appearance of local flow-off paths during a late in-vessel phase requires that several measuring probes with gamma radiation sensors have to be installed around the reactor pressure vessel in order to detect the gamma radiation distribution at the outside. The definition of further core-melt states would be possible in case of a re-flooding of the reactor pressure vessel. However, the increasing filling level would not significantly change the core deformation and the gamma distribution at the outside.

  12. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

  13. A Numerical Procedure for Model Identifiability Analysis Applied to Enzyme Kinetics

    DEFF Research Database (Denmark)

    Daele, Timothy, Van; Van Hoey, Stijn; Gernaey, Krist;

    2015-01-01

    structure evaluation by assessing the local identifiability characteristics of the parameters. Moreover, such a procedure should be generic to make sure it can be applied independent from the structure of the model. We hereby apply a numerical identifiability approach which is based on the work of Walter...... and Pronzato (1997) and which can be easily set up for any type of model. In this paper the proposed approach is applied to the forward reaction rate of the enzyme kinetics proposed by Shin and Kim(1998). Structural identifiability analysis showed that no local structural model problems were occurring....... In contrast, the practical identifiability analysis revealed that high values of the forward rate parameter Vf led to identifiability problems. These problems were even more pronounced athigher substrate concentrations, which illustrates the importance of a proper experimental designto avoid...

  14. Spatially-resolved chemical analysis of frozen ice cores by cryo-cell-UV-laser-ablation-ICPMS

    Science.gov (United States)

    Müller, Wolfgang; Della Lunga, Damiano; Rasmussen, Sune O.; Svensson, Anders

    2015-04-01

    High-latitude ice cores have become the master records of late Pleistocene climate variability. Especially the high-resolution data from Greenland of the past ~125 ka reveal a remarkably changeable glacial climate, and these rapid climate oscillations have been shown to take place within a few years only [1, 2]. The requirement for an improvement in spatial resolution in ice core analysis arises from 1) the continuous thinning of annual layers in deep parts of ice cores to below what is routinely resolvable by continuous flow analysis and 2) the concomitant recrystallization of ice that potentially affects the location of impurities and thus the identification of annual layers. We developed a new technique to analyze elemental concentrations at ppb-levels in frozen ice cores at ~100 um (0.1 mm) resolution, which focuses on seasalt and dust tracers (e.g. Na, Mg, Ca, Al, Fe). It utilizes a custom-built, peltier-cooled cryo-sample holder fully compatible with the two-volume Laurin LA-cell of our RESOlution M-50 excimer (193 nm ArF) LA system, which is coupled to an Agilent 7500cs ICPMS, operated in reaction cell gas mode with H2 to eliminate 40Ar and 40Ar16O to access 40Ca and 56Fe [3]. Using 3 x 5 cm strips of ice cores per sample holder, this setup allows elemental concentrations to be acquired using both depth-profiling along (chains of) spots and/or as continuous lateral profiles, following surface cleaning with a major-element-free ceramic blade. Ice crystal boundaries can be observed with transmitted or reflected light illumination. We focus on NGRIP samples from Greenland Stadial 22 (GS22; ~84-88 ka; ~2695-2720 m) with its corresponding transitions. Owing to analysis in frozen ice, we can easily identify - relative to ice crystal boundaries - the location of cation impurities in both clear ice and so-called cloudy bands that are enriched in impurities. We find a remarkable difference in the location of impurities between these different ice domains [4]. Lower

  15. Consolidating metabolite identifiers to enable contextual and multi-platform metabolomics data analysis

    Directory of Open Access Journals (Sweden)

    Saito Kazuki

    2010-04-01

    Full Text Available Abstract Background Analysis of data from high-throughput experiments depends on the availability of well-structured data that describe the assayed biomolecules. Procedures for obtaining and organizing such meta-data on genes, transcripts and proteins have been streamlined in many data analysis packages, but are still lacking for metabolites. Chemical identifiers are notoriously incoherent, encompassing a wide range of different referencing schemes with varying scope and coverage. Online chemical databases use multiple types of identifiers in parallel but lack a common primary key for reliable database consolidation. Connecting identifiers of analytes found in experimental data with the identifiers of their parent metabolites in public databases can therefore be very laborious. Results Here we present a strategy and a software tool for integrating metabolite identifiers from local reference libraries and public databases that do not depend on a single common primary identifier. The program constructs groups of interconnected identifiers of analytes and metabolites to obtain a local metabolite-centric SQLite database. The created database can be used to map in-house identifiers and synonyms to external resources such as the KEGG database. New identifiers can be imported and directly integrated with existing data. Queries can be performed in a flexible way, both from the command line and from the statistical programming environment R, to obtain data set tailored identifier mappings. Conclusions Efficient cross-referencing of metabolite identifiers is a key technology for metabolomics data analysis. We provide a practical and flexible solution to this task and an open-source program, the metabolite masking tool (MetMask, available at http://metmask.sourceforge.net, that implements our ideas.

  16. Core-Shell Columns in High-Performance Liquid Chromatography: Food Analysis Applications

    Science.gov (United States)

    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

  17. Core-scale solute transport model selection using Monte Carlo analysis

    CERN Document Server

    Malama, Bwalya; James, Scott C

    2013-01-01

    Model applicability to core-scale solute transport is evaluated using breakthrough data from column experiments conducted with conservative tracers tritium (H-3) and sodium-22, and the retarding solute uranium-232. The three models considered are single-porosity, double-porosity with single-rate mobile-immobile mass-exchange, and the multirate model, which is a deterministic model that admits the statistics of a random mobile-immobile mass-exchange rate coefficient. The experiments were conducted on intact Culebra Dolomite core samples. Previously, data were analyzed using single- and double-porosity models although the Culebra Dolomite is known to possess multiple types and scales of porosity, and to exhibit multirate mobile-immobile-domain mass transfer characteristics at field scale. The data are reanalyzed here and null-space Monte Carlo analysis is used to facilitate objective model selection. Prediction (or residual) bias is adopted as a measure of the model structural error. The analysis clearly shows ...

  18. Application of noise analysis to investigate core degradation process during PHEBUS-FPT1 test

    International Nuclear Information System (INIS)

    Noise analysis has been performed for measurement data obtained during PHEBUS-FPT1 test. The purpose of the study is to evaluate the applicability of the noise analysis to the following problems: To get more knowledge about the physical processes going on during severe core conditions; To better understand the core melting process; To establish appropriate on-line shut-down data. Results of the study indicate that the noise analysis is quite promising as a tool for investigating physical processes during the experiment. Compared with conventional approach of evaluating the signal's mean value behaviour, the noise analysis can provide additional, more detailed information: It was found that the neutron flux signal is subjected to additional reactivity perturbations in conjunction with fuel melting and relocation. This can easily be detected by applying noise analysis for the neutron flux signal. It has been demonstrated that the method developed in the present study can provide more accurate estimates of the onset of fuel relocation than using temperature signals from thermocouples in the thermal shroud. Moreover, the result suggests a potential of the present method for tracking the whole process of relocation. The result of the data analysis suggests a possibility of sensor diagnostics which may be important for confirming the quality and reliability of the recorded data. Based on the results achieved it is believed that the combined use of noise analysis and thermocouple signals will provide reliable shut-down criteria for the experiment. 8 refs

  19. Application of noise analysis to investigate core degradation process during PHEBUS-FPT1 test

    Energy Technology Data Exchange (ETDEWEB)

    Oguma, Ritsuo [GSE Power Systems AB, Nykoeping (Sweden)

    1997-01-01

    Noise analysis has been performed for measurement data obtained during PHEBUS-FPT1 test. The purpose of the study is to evaluate the applicability of the noise analysis to the following problems: To get more knowledge about the physical processes going on during severe core conditions; To better understand the core melting process; To establish appropriate on-line shut-down data. Results of the study indicate that the noise analysis is quite promising as a tool for investigating physical processes during the experiment. Compared with conventional approach of evaluating the signal`s mean value behaviour, the noise analysis can provide additional, more detailed information: It was found that the neutron flux signal is subjected to additional reactivity perturbations in conjunction with fuel melting and relocation. This can easily be detected by applying noise analysis for the neutron flux signal. It has been demonstrated that the method developed in the present study can provide more accurate estimates of the onset of fuel relocation than using temperature signals from thermocouples in the thermal shroud. Moreover, the result suggests a potential of the present method for tracking the whole process of relocation. The result of the data analysis suggests a possibility of sensor diagnostics which may be important for confirming the quality and reliability of the recorded data. Based on the results achieved it is believed that the combined use of noise analysis and thermocouple signals will provide reliable shut-down criteria for the experiment. 8 refs.

  20. The adult intestinal core microbiota is determined by analysis depth and health status

    OpenAIRE

    Salonen, A.; Salojärvi, J.; Lahti, L.M.; De Vos

    2012-01-01

    High-throughput molecular methods are currently exploited to characterize the complex and highly individual intestinal microbiota in health and disease. Definition of the human intestinal core microbiota, i.e. the number and the identity of bacteria that are shared among different individuals, is currently one of the main research questions. Here we apply a high-throughput phylogenetic microarray, for a comprehensive and high-resolution microbiota analysis, and a novel computational approach ...

  1. Two-dimension calculation of proposed benchmark core analysis for the BN-600 hybrid reactor

    International Nuclear Information System (INIS)

    This paper presents primary calculation results of the proposed benchmark for a hybrid UOX/MOX fuelled core of the BN-600 reactor. The analysis in this paper uses a R-Z homogeneous model of the BN-600 reactor. Calculation results include effective multiplication factors obtained by both diffusion and Monte Carlo methods; fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution

  2. Three dimension calculation of proposed benchmark core analysis for the BN-600 hybrid reactor

    International Nuclear Information System (INIS)

    This paper presents primary calculation results of the proposed benchmark for a hybrid UOX/MOX fuelled core of the BN-600 reactor. The analysis in this paper uses a HEX-Z homogeneous model of the BN-600 reactor. Calculation results include effective multiplication factors obtained by both diffusion and Monte Carlo methods; fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution

  3. Analysis of core soil and water samples from the Cactus Crater Disposal Site at Enewetak atoll

    International Nuclear Information System (INIS)

    Core soil samples and water samples were collected from the Cactus Crater Disposal Site at Enewetak for analysis of 137Cs, 90Sr, 239+240Pu and 241Am by both gamma spectroscopy and, through a contractor laboratory, by wet chemistry procedures. The samples processing methods, the analytical methods and the analytical quality control are all procedures developed for the continuing Marshall Island radioecology and dose assessment work

  4. Analysis of the chondroitin sulfate proteoglycan core protein (CSPGCP) gene in achondroplasia and pseudoachondroplasia.

    OpenAIRE

    Finkelstein, J E; Doege, K; Yamada, Y; Pyeritz, R E; Graham, J M; Moeschler, J.B.; Pauli, R. M.; Hecht, J T; Francomano, C A

    1991-01-01

    Achondroplasia and pseudoachondroplasia are autosomal dominant skeletal dysplasias resulting in short-limbed dwarfism. Histologic and ultrastructural studies of the cartilage in pseudoachondroplasia and in homozygous achondroplasia have suggested a structural abnormality in chondroitin sulfate proteoglycan (CSPG), a major structural protein in the extra-cellular matrix. The gene encoding CSPG core protein (CSPGCP) is thus a logical "candidate gene" for analysis in these conditions. cDNA probe...

  5. Tank 241-TX-113 rotary mode core sampling and analysis plan

    International Nuclear Information System (INIS)

    This sampling and analysis plan (SAP) identities characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for push mode core samples from tank 241-TX-113 (TX-113). The Tank Characterization Technical Sampling Basis document identities Retrieval, Pretreatment and Immobilization as an issue that applies to tank TX-113. As a result, a 150 gram composite of solids shall be made and archived for that program. This tank is not on a Watch List

  6. Lumped parameter analysis of Pb-Bi cooled fast reactor PEACER core using MATRA

    International Nuclear Information System (INIS)

    A lead-bismuth (Pb-Bi) eutectic cooled fast reactor PEACER (Proliferation-resistant, Environment-friendly, Accident-tolerant, Continuable-energy and Economical Reactor) is under development at the Seoul National University. This study is intended to examine the liquid metal coolant behavior along the subchannels and to find out whether the given flux profiles and geometrical arrangement of fuel rods yield reasonable flow distribution during nominal operation using the subchannel analysis code MATRA (Multi-channel Analyzer for Transient and steady-state in Rod Arrays). MATRA was developed at the Korea Atomic Energy Research Institute based on the subchannel approach to calculate the enthalpy and flow distribution in fuel rod bundle elements for both steady-state and transient conditions. The best-estimate analysis was carried out utilizing MATRA for the PEACER-300 quadrant core under the nominal operation condition. Subchannel analysis was performed for the hottest assembly of the PEACER-300 core. The calculation result showed that during normal operation the core material temperature distribution stays well below the thermal design limits. Comparison of the code results with those by hand calculation resulted in good agreement. Hand calculations are in further progress to include the finite difference scheme in the radial direction

  7. Application of noise analysis for the study of core local instability at Forsmark 1

    International Nuclear Information System (INIS)

    Core local instability was recently experienced at Forsmark 1 BWR. The event has been studied by applying noise analysis to data collected in January 1997 for the stability test. The result indicated that there was a region in the left corner of the core which was subject to instability due to neutronic and thermal-hydraulic coupling. The result of the noise analysis suggested two types of disturbance source, one in the vicinity of the detector string LPRM10 having resonant oscillation at 0.5 Hz and another relatively wide band noise in the neighbourhood of LPRM18. Three hypotheses have been examined as the possible cause, operational factor, abnormal fuel assembly, and wide band low frequency disturbance. Although the real cause has not been made clear from the noise analysis, it is likely that the operational factor played an important role as the cause. Further investigations are expected to be performed in the future. In order to detect the local instability it is important to have a stability monitor with a capability of monitoring a sufficient number of LPRMs so as to cover the whole core. This is important since local instability is a type of anomaly which should not occur during reactor operation

  8. Analysis of a German BWR core with TRACE/PARCS using different cross section sets

    International Nuclear Information System (INIS)

    'Full text:' Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools as well as appropriate cross sections (XS) libraries depending on the individual thermal hydraulic state parameters. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running and user-friendly lattice codes such as Casco and Helios. At research institutions and universities, alternative tools to the commercial ones with full access to the source code as well as moderate cost are urgently needed. The Scale system is being developed and improved for lattice physics calculations of real core loading of Light Water Reactors (LWR). It represents a promising alternative to the commercial lattice codes. At Karlsruhe Institute of Technology (Kit) a computational route based on Scale/Triton/Newt for BWR core loading is under development. The generated XS-data sets have to be transformed in PMAXS-format for use in the reactor dynamic code PARCS. This task is performed by the module GenPMAXS being developed and tested at the Michigan University. To verify the computational route, a BWR fuel assembly depletion problem was calculated by PARCS and compared to the CASMO results. Since the SCALE/TRITON XS-file does actually not contain all required neutronic data, FORTRAN routines have been developed to incorporate the missing data e.g. the yields of Iodine, Xenon and Promethium into the XS-data sets in the PMAXS-format. The comparison of the results obtained with PARCS (using the corrected PMAXS file) and CASMO for the depletion problem exhibited a good agreement. Consequently, this approach was followed for the generation of a complete XS-set for a real BWR core to be used in subsequent transient analysis. Then 3D neutronic and thermal hydraulic core model were elaborated for a TRACE/PARCS analysis. The thermal hydraulic model is based on the 3D VESSEL

  9. Lumped parameter analysis of Pb-Bi cooled fast reactor PEACER core using MATRA

    International Nuclear Information System (INIS)

    PEACER core is designed to produce 1560MW of thermal output with electric output of 550MW. PEACER uses B4C control rods and lead-bismuth (Pb-Bi) coolant in the primary system. This work examines the Pb-Bi coolant behavior along the PEACER fuel channels and to check on whether the given heat flux profiles and geometrical arrangement of the fuel rods yield reasonable fluid dynamic distribution under nominal operation resorting to a subchannel approach using MATRA. MATRA is a thermal hydraulic analysis code based on the subchannel approach for calculating the enthalpy and flow distribution in the fuel rod bundle during steady-state and transient conditions. The calculational result revealed that the input data based on the current design of PEACER core yielded reasonable results mostly satisfying the thermal design limits. The calculation results, however, indicated a potential for fuel damage in the hottest assembly of the core. This was found to be mainly due to excessively conservative assumptions made in generating the input conditions. Work is underway to apply physically-based conditions of the PEACER core and more reliable rod-to-coolant heat transfer correlations. (author)

  10. Hydrodynamic and elastoplastic structural analysis of fast breeder reactor core accident

    International Nuclear Information System (INIS)

    This paper describes the principles and examples of applications of an explicit Lagrangian coupled finite difference-finite element code HEMP-ESI developed in order to calculate the structural consequences of hypothetical core disruptive accidents (HCDA) in nuclear reactors. The explicit solution algorithm of the finite difference scheme used to discretize the hydrodynamic fluid domains is shown to be very similar to that used for the solution of the finite element discretized shell structures, hence permitting an easy and efficient coupling. Two examples of simulation show the applicability of the method to nuclear reactor core safety analysis (test problem). Core explosion in a loop-type reactor including a shell containment: the calculation shows the energy absorbing function of the shell and enables the evaluation of the forces acting on the reactor containment. Hypothetical Core Disruptive Accident in a fast breeder reactor: the calculation shows the main features of this accident: lifting of the liquid sodium above the explosion and impact on the cover head inducing upward deformations; radial outflow of the sodium which induces large deformations of the inner and outer shell; zones of compressive circumferential stresses in the main shell at the junction of the spherical head and the cylindrical part

  11. Development of the enigma fuel performance code for whole core analysis and dry storage assessments

    Energy Technology Data Exchange (ETDEWEB)

    Rositer, Glyn [Fuel Cycle Solutions, UK National Nuclear Laboratory, Lancashire (United Kingdom)

    2011-11-15

    UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage-this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

  12. Reflooding of a severely damaged reactor core. Experimental analysis and modelling

    International Nuclear Information System (INIS)

    The understanding of the reflood process of a severely damaged reactor core represents a challenge in the prediction of safety margin of existing and future pressurized water reactors. After the TMI-2 accident, the understanding of coolability of severely damaged reactor core became an objective of many theoretical and experimental studies. Currently, the French Institute of Radioprotection and Nuclear Safety (IRSN) has started two experimental programs, PRELUDE and PEARL, to investigate the physical phenomena during a reflood process at high temperature and to provide relevant data in order to improve predictive models. The purpose of this paper is to propose a consistent thermo-hydraulic model of reflood of severely damaged reactor core. The presented model is based on the theory of heat transfer and two-phase flow in porous media and in small hydraulic diameter channels. The proposed model is implemented into the European computer code for severe accident analysis ICARE-CATHARE. The comparison of the calculations with PRELUDE experimental results is presented. Finally, the issue of transposition to the reactor scale is discussed and some answers are proposed using calculation results for a debris bed in a configuration similar to what could be expected in a severely damaged reactor core. (author)

  13. Safety analysis for core conversion (from HEU to LEU) of Pakistan research reactor-2 (PARR-2)

    International Nuclear Information System (INIS)

    PARR-2 (Pakistan Research Reactor-2), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. The present report presents thermal hydraulic and safety analyses of PARR-2 using existing HEU fuel as well as proposed LEU fuel. Presently, the core is comprised of 90.2% enriched UAl4-Al fuel. There are 344 fuel pins of 5.5 mm diameter. The core has a total of 994.8 g of U235. Standard computer code PARET/ANL (version 1992) was employed to perform steady-state and transient analyses. Various parameters were computed, which included: coolant outlet, maximum clad surface and maximum fuel centerline temperatures; and peak power and corresponding peak core temperatures resulting from a transient initiated by 4 mK positive reactivity insertion. Results were compared with the reported data in Final Safety Analysis Report (FSAR). It was found that the PARET results were in reasonable agreement with the manufacturer's results. Calculations were also carried out for the proposed LEU core with two suggested fuel pin sizes (5.5 mm and 5.1 mm diameter with 12.6% and 12.3% enrichment, respectively). Comparison of the LEU results with the existing HEU fuel has been made and discussed.

  14. Development and performance analysis of EPICS channel access server on FPGA based soft-core processor

    International Nuclear Information System (INIS)

    A soft core processor is a flexible hardware description language (HDL) model of a specific processor (CPU) that can be customized for a given application and synthesized for an FPGA as opposed to a hard core processor which is fixed in silicon. Combined with an on-board ethernet port, the technology incorporates integrating the IOC and digital control hardware within a single FPGA thus reducing the overall hardware complexities of field devices. In this paper, the technical details of porting EPICS Channel Access Server on MicroBlaze soft-core processor are explained. The EPICS performance on the MicroBlaze processor is analyzed. For this, the CPU load and server processing time for different numbers of Process Variables (PVs) have been studied for this platform. On the basis of the analysis, critical parameters of EPICS on this embedded platform have been derived and a few modifications in the channel access protocol are proposed for MicroBlaze soft-core processor. (author)

  15. Development of the enigma fuel performance code for whole core analysis and dry storage assessments

    International Nuclear Information System (INIS)

    UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage-this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

  16. Analysis of core damage frequency, Surry, Unit 1 internal events appendices

    International Nuclear Information System (INIS)

    This document contains the appendices for the accident sequence analyses of internally initiated events for the Surry Nuclear Station, Unit 1. This is one of the five plant analyses conducted as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 documents the risk of a selected group of nuclear power plants. The work performed is an extensive reanalysis of that published in November 1986 as NUREG/CR-4450, Volume 3. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved. The context and detail of this report are directed toward PRA practitioners who need to know how the work was performed and the details for use in further studies. The mean core damage frequency at Surry was calculated to be 4.0E-5 per year, with a 95% upper bound of 1.3E-4 and 5% lower bound of 6.8E-6 per year. Station blackout type accidents (loss of all AC power) were the largest contributors to the core damage frequency, accounting for approximately 68% of the total. The next type of dominant contributors were Loss of Coolant Accidents (LOCAs). These sequences account for 15% of core damage frequency. No other type of sequence accounts for more than 10% of core damage frequency

  17. Performance analysis of saturated iron core superconducting fault current limiter using Jiles–Atherton hysteresis model

    International Nuclear Information System (INIS)

    In this paper study of the Saturated Iron Core Superconducting Fault Current Limiter (SISFCL) has been carried out. Since in an SISFCL, the iron core plays a key role in distributing the magnetic flux, the hysteresis property of the core material has been introduced in a mathematical model to get a more accurate result. In this paper the Jiles–Atherton hysteresis model has been used for modeling the core. The equations are solved through numerical method and performances of SISFCL are analyzed for both normal and fault conditions. On further analysis it is observed that for suppression of higher value of fault current a high voltage develops across the DC source. Hence there is a chance of the DC source being damaged by the rise in voltage under fault condition. In order to protect the DC source, a shorted ring is introduced in the SISFCL circuit and its effects have been analyzed. It is noticed that the shorted ring has successfully reduced the voltage across the DC coil during fault condition while the performance of the limiter remains the same. - Highlights: • Mathematical modeling of SISFCL has been established. • Hysteresis effect has been included using J–A model. • Mathematical model of SISFCL using the Shorted Ring have been established. • Comparisons are drawn between the responses with and without shorted ring

  18. The accuracy of frozen section analysis in ultrasound- guided core needle biopsy of breast lesions

    Directory of Open Access Journals (Sweden)

    Riss Paul

    2009-09-01

    Full Text Available Abstract Background Limited data are available to evaluate the accuracy of frozen section analysis and ultrasound- guided core needle biopsy of the breast. Methods In a retrospective analysis data of 120 consecutive handheldultrasound- guided 14- gauge automated core needle biopsies (CNB in 109 consecutive patients with breast lesions between 2006 and 2007 were evaluated. Results In our outpatient clinic120 CNB were performed. In 59/120 (49.2% cases we compared histological diagnosis on frozen sections with those on paraffin sections of CNB and finally with the result of open biopsy. Of the cases 42/59 (71.2% were proved to be malignant and 17/59 (28.8% to be benign in the definitive histology. 2/59 (3.3% biopsies had a false negative frozen section result. No false positive results of the intraoperative frozen section analysis were obtained, resulting in a sensitivity, specificity and positive predicting value (PPV and negative predicting value (NPV of 95%, 100%, 100% and 90%, respectively. Histological and morphobiological parameters did not show up relevance for correct frozen section analysis. In cases of malignancy time between diagnosis and definitive treatment could not be reduced due to frozen section analysis. Conclusion The frozen section analysis of suspect breast lesions performed by CNB displays good sensitivity/specificity characteristics. Immediate investigations of CNB is an accurate diagnostic tool and an important step in reducing psychological strain by minimizing the period of uncertainty in patients with breast tumor.

  19. The accuracy of frozen section analysis in ultrasound- guided core needle biopsy of breast lesions

    International Nuclear Information System (INIS)

    Limited data are available to evaluate the accuracy of frozen section analysis and ultrasound- guided core needle biopsy of the breast. In a retrospective analysis data of 120 consecutive handheldultrasound- guided 14- gauge automated core needle biopsies (CNB) in 109 consecutive patients with breast lesions between 2006 and 2007 were evaluated. In our outpatient clinic120 CNB were performed. In 59/120 (49.2%) cases we compared histological diagnosis on frozen sections with those on paraffin sections of CNB and finally with the result of open biopsy. Of the cases 42/59 (71.2%) were proved to be malignant and 17/59 (28.8%) to be benign in the definitive histology. 2/59 (3.3%) biopsies had a false negative frozen section result. No false positive results of the intraoperative frozen section analysis were obtained, resulting in a sensitivity, specificity and positive predicting value (PPV) and negative predicting value (NPV) of 95%, 100%, 100% and 90%, respectively. Histological and morphobiological parameters did not show up relevance for correct frozen section analysis. In cases of malignancy time between diagnosis and definitive treatment could not be reduced due to frozen section analysis. The frozen section analysis of suspect breast lesions performed by CNB displays good sensitivity/specificity characteristics. Immediate investigations of CNB is an accurate diagnostic tool and an important step in reducing psychological strain by minimizing the period of uncertainty in patients with breast tumor

  20. An approach to model reactor core nodalization for deterministic safety analysis

    Science.gov (United States)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  1. An approach to model reactor core nodalization for deterministic safety analysis

    International Nuclear Information System (INIS)

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M

  2. An approach to model reactor core nodalization for deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my [Nuclear Energy Department, Regulatory Economics & Planning Division, Tenaga Nasional Berhad (Malaysia); Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my [Prototypes & Plant Development Center, Malaysian Nuclear Agency (Malaysia); Roslan, Ridha, E-mail: ridha@aelb.gov.my; Sadri, Abd Aziz [Nuclear Installation Divisions, Atomic Energy Licensing Board (Malaysia); Farid, Mohd Fairus Abd [Reactor Technology Center, Malaysian Nuclear Agency (Malaysia)

    2016-01-22

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  3. Accuracy of Percutaneous Core Biopsy in the Diagnosis of Small Renal Masses (≤4.0 cm): A Meta-analysis

    OpenAIRE

    Qiqi He; Hanzhang Wang; Jonathan Kenyon; Guiming Liu; Li Yang; Junqiang Tian; Zhongjin Yue; Zhiping Wang

    2015-01-01

    Objective To use meta-analysis to determine the accuracy of percutaneous core needle biopsy in the diagnosis of small renal masses (SMRs≤4.0 cm). Materials and Methods Studies were identified by searching PubMed, Embase, and the Cochrane Library database up to March 2013. Two of the authors independently assessed the study quality using QUADAS-2 tool and extracted data that met the inclusion criteria. The sensitivity, specificity, likelihood ratios, diagnostic odds ratio (DOR) and also summar...

  4. Reactivity accidents analysis during natural core cooling operation of ETRR-2

    International Nuclear Information System (INIS)

    One of the main features of Egypt Test and Research Reactor Number 2 (ETRR-2), MTR type, is a continuous steady-state operation at low power level, <=400 kW, with core cooling by natural water circulation. Two flapper valves mounted on the return core cooling pipe lines and long chimney encloses the reactor core and assure natural convection phenomena through the reactor core and reactor pool. Many tests and experiments are carried out during this state of operation. A possible occurrence of reactivity insertion accidents (RIA) may be expected over this operation. The present work studies two types of possible RIA: 1-fast reactivity insertion accident (FRIA) with rate 1.04$/s and 2-slow reactivity insertion accident (SRIA) with rate 0.023$/s which may occur due to fast/slow withdrawal of a control rod or sudden cooling of the core inlet water temperature. Failure or success of the reactor scram system during the transient operation is considered. A computer code TRAP22 is developed for such analysis. It is verified against CONVEC code and commissioning tests for steady state operation. The results of verification show good agreement. The study demonstrates that the reactor can be scrammed safely due to either FRIA or SRIA, whenever the maximum expected hot channel HC clad temperature lies within the range 70.73-71.85 deg. C. While, in case of failure of scram system the maximum (HC) clad temperature reaches the burn out value at time 1.175s for FRIA and at 46.36s for SRIA. At the burn out point the clad surface heat flux exceeds its design critical value which results in partial fuel melt

  5. False-negative results of breast core needle biopsies – retrospective analysis of 988 biopsies

    International Nuclear Information System (INIS)

    Breast cancer is the most common malignant neoplasm and the most common cause of death among women. The core needle biopsy is becoming a universal practice in diagnosing breast lesions suspected of malignancy. Unfortunately, breast core needle biopsies also bear the risk of having false-negative results. 988 core needle breast biopsies were performed at the Maria Skłodowska-Curie Memorial Cancer Center and Institute of Oncology, Gliwice Branch, between 01 March 2006 and 29 February 2008. Malignant lesions were diagnosed in 426/988 (43.12%) cases, atypical hyperplasia in 69/988 (6.98%), and benign lesions in 493/988 (49.90%) cases. Twenty-two out of 988 biopsies (2.23%) were found to be false negative. Histopathological assessment of tissue specimens was repeated in these cases. In 14/22 (64%) cases, the previous diagnosis of a benign lesion was changed. In 8/22 (36%) cases, the diagnosis of a benign lesion was confirmed. False-negative rate was calculated at 2.2%. The rate of false-negative diagnoses resulting from a radiological mistake was estimated at 36%. The rate of false-negative diagnoses, resulting from histopathological assessment, was 64%. False-negative results caused by a radiological error comprised 1.5% of all histopathologically diagnosed cancers and atypias (sensitivity of 98.5%). There were no false-positive results in our material - the specificity of the method was 100%. Histopathological interpretation is a substantial cause of false-negative results of breast core needle biopsy. Thus, in case of a radiological-histopathological divergence, histopathological analysis of biopsy specimens should be repeated. The main radiological causes of false-negative results of breast core needle biopsy are as follows: sampling from an inappropriate site and histopathological non-homogeneity of cancer infiltration

  6. Analysis of containment venting following a core damage at a BWR Mark I using THALES-2

    International Nuclear Information System (INIS)

    Analysis of containment venting following a core damage at a boiling water reactor (BWR) Mark I using THALES-2 was performed. In this analysis, the effect of various parameters, namely, the areas of the vent path, containment venting pressure, and accident sequences on the containment thermodynamic response, and radionuclide transport and release in the containment venting at a BWR was examined. The code THALES-2B developed by the Japan Atomic Energy Research Institute (JAERI) was used in this analysis. The model plant in this analysis was the Browns Ferry plant. From this analysis was found that the 4-inch pipe of containment venting flow path is sufficient to maintain the containment pressure in the specified range if the containment was pressurized by the decay heat power. The entrainment by the pool swelling as well as by the flashing was not occurred during the containment venting. The source terms are not sensitive to the variation of containment venting flow path area. The containment venting pressure operation setting point has important rule in the containment venting. In the containment venting, the source terms are not sensitive to the accident sequence, except for Sr source term. In order to get better understanding on the containment venting strategy, the following analyses are necessary. Analyses of accident sequence which has a high power such as anticipated transient without scram are necessary, as well as analyses of accident sequence which pressurize the containment before the core damage. (author)

  7. [Effective core formulae for lung cancer based on complex network and survival analysis].

    Science.gov (United States)

    Yang, Ming; Li, Jia-qi; Jiao, Li-jing; Chen, Pei-qi; Xu, Ling

    2015-11-01

    The study on the effective core formulae (CEF) not only summarized traditional chinese medicine (TCM) treatment experience, but also helped reveal the underlying knowledge in the formulation of TCM prescriptions. The aim of the present paper was to investigate the method of data mining for the discovery of core effective formulae for lung cancer. In the present study, a prescription fingerprint approach was used to characterize the staged prescription information of patients. The D index was used to screen potential beneficial herbs. Then, based on a herbal compatibility network, the maximal clique searching algorithm (BK algorithm) and survival analysis were applied to discover CEF for lung cancer, and a mining analysis was made for the 322 cases from Longhua hospital. The correlation between prescriptions and survival time was analyzed by prescription fingerprints. Forty-three potentially beneficial herbs were obtained, and two CEFs were significant for the survival time by a parametric survival model based on lognormal distribution, the results were verified by a multivariate survival model. The rules of combination of the two CEFs basically conform to TCM onco-therapeutic theory of strengthening the body resistance and the actual conditions in clinic. All results showed that the established approach was feasible for discovering the core effective formulae for lung cancer and mining survival data for complex TCM onco-therapy. PMID:27097428

  8. Development of computer code packages for molten salt reactor core analysis

    International Nuclear Information System (INIS)

    This paper presents the implementations of the Oak Ridge National Laboratory (ORNL) approach for Molten Salt Reactor (MSR) core analysis with two nuclear reactor core analysis computer code systems. The first code system has been set up with the MCNP6 Monte Carlo code, its depletion module CINDER90 and the PYTHON script language. The second code system has been set up with the NEWT transport calculation module and ORIGEN depletion module connected by TRITON sequence in SCALE code, and the PYTHON script language. The PYTHON script language is used for implementing the online reprocessing of molten-salt fuel, and feeding new fertile material in the computer code simulations. In this paper, simplified nuclear reactor core models of a Molten Salt Breeder Reactor (MSBR), designed by ORNL in the 1960's, and FUJI-U3 designed by Toyohashi University of Technology (TUT) in the 2000's, were analyzed by the two code systems. Using these, various reactor design parameters of the MSRs were compared, such as the multiplication factor, breeding ratio, amount of material, total feeding, neutron flux distribution, and temperature coefficient. (author)

  9. Neutronic Analysis of the Pavia University TRIGA Mark II Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Tigliole, A. Borio Di [University of Pavia, Department of Nuclear and Theoretical Physics, Via Bassi 4, 27100 Pavia (Italy); University of Pavia, Laboratory of Applied Nuclear Energy, Via Aselli 41, 27100 Pavia (Italy); Cammi, A. [Polytechnic of Milano, Department of Energy, Via La Masa 34, 20156 Milano (Italy); Chiesa, D.; Clemenza, M. [University of Milano-Bicocca, Physics Department, Piazza della Scienza 3, 20126, Milano (Italy); Pattavina, L.; Previtali, E. [University of Milano-Bicocca, Physics Department, Piazza della Scienza 3, 20126, Milano (Italy); INFN section of Milano-Bicocca, Piazza della Scienza 3, 20126, Milano (Italy); Scian, G. [University of Pavia, Laboratory of Applied Nuclear Energy, Via Aselli 41, 27100 Pavia (Italy)

    2011-07-01

    An advanced computational tool based on Monte Carlo code was developed for the neutronic analysis of the TRIGA Mark II research reactor of the University of Pavia and the results from simulations were benchmarked with the experimental data. The neutron analysis was performed by means of the 3-D continuous-energy Monte Carlo codes MCNP4C and MCNP5. All core components were reproduced in good detail and the model allows the description of different core and fuel configurations with good versatility. In a first step, the results from the simulations with fresh fuel at room temperature were compared with the experimental data about critical core parameters at zero power level, quoted in the reactor First Criticality Final Report; a very good agreement was found and the goodness of the model was proved. Afterwards, neutron flux and reactivity dependence on the fuel and moderator temperature was studied in detail; all the available methods for the simulation of thermal effects were analyzed in order to model the reactor in the stationary operating condition at 250 kW. The model was updated to the present configuration and the use of specific S({alpha},{beta}) cross sections was found to be of primary importance for a correct simulation of low-energy neutrons scattering within moderator. (author)

  10. Neutronic Analysis of the Pavia University TRIGA Mark II Reactor Core

    International Nuclear Information System (INIS)

    An advanced computational tool based on Monte Carlo code was developed for the neutronic analysis of the TRIGA Mark II research reactor of the University of Pavia and the results from simulations were benchmarked with the experimental data. The neutron analysis was performed by means of the 3-D continuous-energy Monte Carlo codes MCNP4C and MCNP5. All core components were reproduced in good detail and the model allows the description of different core and fuel configurations with good versatility. In a first step, the results from the simulations with fresh fuel at room temperature were compared with the experimental data about critical core parameters at zero power level, quoted in the reactor First Criticality Final Report; a very good agreement was found and the goodness of the model was proved. Afterwards, neutron flux and reactivity dependence on the fuel and moderator temperature was studied in detail; all the available methods for the simulation of thermal effects were analyzed in order to model the reactor in the stationary operating condition at 250 kW. The model was updated to the present configuration and the use of specific S(α,β) cross sections was found to be of primary importance for a correct simulation of low-energy neutrons scattering within moderator. (author)

  11. Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

    International Nuclear Information System (INIS)

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general

  12. A 3D transport-based core analysis code for research reactors with unstructured geometry

    International Nuclear Information System (INIS)

    Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal SN method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of keff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

  13. Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Kolaczkowski, A.M.; Lambright, J.A.; Ferrell, W.L.; Cathey, N.G.; Najafi, B.; Harper, F.T.

    1986-10-01

    This document contains the internal event initiated accident sequence analyses for Peach Bottom, Unit 2; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Peach Bottom, Unit 2, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provided additional insights regarding the dominant contributors to the Peach Bottom core damage frequency estimate. The mean core damage frequency at Peach Bottom was calculated to be 8.2E-6. Station blackout type accidents (loss of all ac power) were found to dominate the overall results. Anticipated Transient Without Scram accidents were also found to be non-negligible contributors. The numerical results are largely driven by common mode failure probability estimates and to some extent, human error. Because of significant data and analysis uncertainties in these two areas (important, for instance, to the most dominant scenario in this study), it is recommended that the results of the uncertainty and sensitivity analyses be considered before any actions are taken based on this analysis.

  14. Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

    International Nuclear Information System (INIS)

    This document contains the internal event initiated accident sequence analyses for Peach Bottom, Unit 2; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Peach Bottom, Unit 2, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provided additional insights regarding the dominant contributors to the Peach Bottom core damage frequency estimate. The mean core damage frequency at Peach Bottom was calculated to be 8.2E-6. Station blackout type accidents (loss of all ac power) were found to dominate the overall results. Anticipated Transient Without Scram accidents were also found to be non-negligible contributors. The numerical results are largely driven by common mode failure probability estimates and to some extent, human error. Because of significant data and analysis uncertainties in these two areas (important, for instance, to the most dominant scenario in this study), it is recommended that the results of the uncertainty and sensitivity analyses be considered before any actions are taken based on this analysis

  15. Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, Jonathan, E-mail: JMitchell16@slb.com; Fordham, Edmund J. [Schlumberger Gould Research, High Cross, Madingley Road, Cambridge CB3 0EL (United Kingdom)

    2014-11-15

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general.

  16. A new gene co-expression network analysis based on Core Structure Detection (CSD)

    OpenAIRE

    Brunet, A-C; Azais, J-M; Loubes, J-M; Amar, J; Burcelin, R

    2016-01-01

    We propose a novel method to cluster gene networks. Based on a dissimilarity built using correlation structures, we consider networks that connect all the genes based on the strength of their dissimilarity. The large number of genes require the use of the threshold to find sparse structures in the graph. in this work, using the notion of graph coreness, we identify clusters of genes which are central in the network. Then we estimate a network that has these genes as main hubs. We use this new...

  17. BGCore - A Comprehensive Package for Reactor Core and Fuel Storage Analysis

    International Nuclear Information System (INIS)

    Recent interest in Fast Gas Cooled Reactors requires major adaptations or evolutions of calculation tools to accommodate the innovative features of core design (new fuel and subassembly forms), fuel composition (homogeneous recycling of minor actinides). Fast neutron spectrum renders inadequate the familiar group reduction schemes and homogenization methods used in LWR analysis. In addition, the specificities of Fast Gas Cooled Reactors (materials, subassembly design, preferential direction for neutron leakage (streaming), high temperatures, particular reactivity effects, etc.) require at least an increase in the number of nuclides to be taken into account in the neutronic libraries with an extended tabulation in temperature. Enhancement of neutronic calculational tools is needed for S/A heterogeneity and anisotropy and to accurately model control elements and other non fueled regions. Several computational systems recently developed are widely used and several others are currently under development. All of the systems are based on Monte-Carlo codes for a 3 Dimensional representation of core and ORIGEN(8) code for fuel composition calculations. This abstract presents the outline and current progress of a development of a comprehensive calculational system for Fast Gas Cooled Reactors carried out in Ben-Gurion University. The flow chart of the system (temporarily) designated as BG-CORE is shown in Fig. 1. Our approach follows, in general, that adapted in other systems, i.e. interfacing the core model (MCNP) with a SARAF - an independently developed code for calculating fuel composition in-core and spent fuel emissions following discharge. Two novel features are added: temperature distributions and feedback, and fuel management. The following modules comprise the BG-CORE system: MCNP - Monte-Carlo code for 3 dimensional core representation using cross-section data sets based on JEF-2/JEF-3 and ENDFB-VI. This module provides Kerf, flux and power density distributions

  18. Subchannel analysis of a small ultra-long cycle fast reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Han; Kim, Ji Hyun; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2014-04-01

    Highlights: • The UCFR-100 is small-sized one of 60 years long-life nuclear reactors without refueling. • The design safety limits of the UCFR-100 are evaluated using MATRA-LMR. • The subchannel results are below the safety limits of general SFR design criteria. - Abstract: Thermal-hydraulic evaluation of a small ultra-long cycle fast reactor (UCFR) core is performed based on existing safety regulations. The UCFR is an innovative reactor newly designed with long-life core based on the breed-and-burn strategy and has a target electric power of 100 MWe (UCFR-100). Low enriched uranium (LEU) located at the bottom region of the core play the role of igniter to operate the UCFR for 60 years without refueling. A metallic form is selected as a burning fuel region material after the LEU location. HT-9 and sodium are used as cladding and coolant materials, respectively. In the present study, MATRA-LMR, subchannel analysis code, is used for evaluating the safety design limit of the UCFR-100 in terms of fuel, cladding, and coolant temperature distributions in the core as design criteria of a general fast reactor. The start-up period (0 year of operation), the middle of operating period (30 years of operation), and the end of operating cycle (60 years of operation) are analyzed and evaluated. The maximum cladding surface temperature (MCST) at the BOC (beginning of core life) is 498 °C on average and 551 °C when considering peaking factor, while the MCST at the MOC (middle of core life) is 498 °C on average and 548 °C in the hot channel, respectively, and the MCST at the EOC (end of core life) is 499 °C on average and 538 °C in the hot channel, respectively. The maximum cladding surface temperature over the long cycle is found at the BOC due to its high peaking factor. It is found that all results including fuel rods, cladding, and coolant exit temperature are below the safety limit of general SFR design criteria.

  19. Analysis of CHAMP scalar magnetic data to identify ocean circulation signals

    DEFF Research Database (Denmark)

    Manoj, C.; Maus, S.; Kuvshinov, Alexei;

    Unlike tidal ocean signals, the magnetic signal of ocean circulation has not yet been identified in satellite magnetic data. In particular, the steady signal of mean ocean flow is indistinguishable from time invariant crustal signals. One option, therefore, is to predict the seasonal and annual...... variations in the ocean flow signal from ocean circulation models and compare them with the corresponding variations in satellite magnetic residuals. We used the 11 year ECCO-1 simulation data to derive the ocean transport. A 3D EM induction code in its low frequency limit, was used to simulate the magnetic...... signals at satellite altitude. We predict annual variation amplitudes in the scalar anomaly of the order of 0.3 nT. We compare these predictions with the particularly quiet CHAMP night-time scalar data, subtracting core, mantle, crustal, ocean tidal, and magnetospheric contributions to the field. The...

  20. Low time resolution analysis of polar ice cores cannot detect impulsive nitrate events

    CERN Document Server

    Smart, D F; Melott, A L; Laird, C M

    2015-01-01

    Ice cores are archives of climate change and possibly large solar proton events (SPEs). Wolff et al. (2012) used a single event, a nitrate peak in the GISP2-H core, which McCracken et al. (2001a) time associated with the poorly quantified 1859 Carrington event, to discredit SPE-produced, impulsive nitrate deposition in polar ice. This is not the ideal test case. We critique the Wolff et al. analysis and demonstrate that the data they used cannot detect impulsive nitrate events because of resolution limitations. We suggest re-examination of the top of the Greenland ice sheet at key intervals over the last two millennia with attention to fine resolution and replicate sampling of multiple species. This will allow further insight into polar depositional processes on a sub-seasonal scale, including atmospheric sources, transport mechanisms to the ice sheet, post-depositional interactions, and a potential SPE association.

  1. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    International Nuclear Information System (INIS)

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  2. Joint European contribution to phase 4 of the BN-600 hybrid reactor benchmark core analysis

    International Nuclear Information System (INIS)

    Hybrid UOX/MOX fuelled core of BN-600 reactor was endorsed as an international benchmark. Phase 4 of the RCM benchmark studies consider full MOX core differentiated by design measures to reduce the sodium void worth. Parameters to be calculated were: fuel and steel Doppler coefficients; fuel density coefficient; sodium density coefficient; power distribution for fuel and non-fuelled regions; β effective and prompt neutron life time. Heterogeneity effects are evaluated. Analysis was carried out using ERANOS code and data system for fast reactors. Nuclear data library is based on JEF2.2. Accurate calculations of control rod heterogeneity effects with homogeneous equivalent cross sections for control rod absorbers were prepared using reactivity equivalence technique

  3. Thermal-Fluid and Safety Analysis of the TRU Deep-Burn MHR Core

    International Nuclear Information System (INIS)

    The DB-MHR (Deep Burn-Modular Helium Reactor) concept was proposed by GA to achieve a very high burnup of the LWR TRU fuel. To increase the TRU discharge burnup, the original GT-MHR of GA was modified for the DB-MHR core: a 5-fuel-ring configuration was adopted instead of the original 3-fueling concept. This paper describes the GAMMA+ thermal-fluid analysis of the 600MWth DB-MHR system at the steady state and the transient condition of LPCC (Low Pressure Conduction Cooling) event. The objective of this study is to characterize the DB-MHR core in terms of the fuel temperature during the nominal and LPCC conditions

  4. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    Energy Technology Data Exchange (ETDEWEB)

    Prabhu Gaunkar, N., E-mail: neelampg@iastate.edu; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C. [Department of Electrical and Computer Engineering, Iowa State University, Ames, Iowa 50011 (United States); Bulu, I.; Ganesan, K.; Song, Y. Q. [Schlumberger-Doll Research, Cambridge, Massachusetts 02139 (United States)

    2015-05-07

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors.

  5. Identifying Innovative Interventions to Promote Healthy Eating Using Consumption-Oriented Food Supply Chain Analysis

    OpenAIRE

    Hawkes, Corinna, ed.

    2009-01-01

    The mapping and analysis of supply chains is a technique increasingly used to address problems in the food system. Yet such supply chain management has not yet been applied as a means of encouraging healthier diets. Moreover, most policies recommended to promote healthy eating focus on the consumer end of the chain. This article proposes a consumption-oriented food supply chain analysis to identify the changes needed in the food supply chain to create a healthier food environment, measured in...

  6. Preliminary safety analysis for key design features of KALIMER with breakeven core

    International Nuclear Information System (INIS)

    KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model

  7. Identifying Effective Spelling Interventions Using a Brief Experimental Analysis and Extended Analysis

    Science.gov (United States)

    McCurdy, Merilee; Clure, Lynne F.; Bleck, Amanda A.; Schmitz, Stephanie L.

    2016-01-01

    Spelling is an important skill that is crucial to effective written communication. In this study, brief experimental analysis procedures were used to examine spelling instruction strategies (e.g., whole word correction; word study strategy; positive practice; and cover, copy, and compare) for four students. In addition, an extended analysis was…

  8. BEACONTM Core Monitoring and Analysis for Operations of the Westinghouse AP1000

    International Nuclear Information System (INIS)

    The Westinghouse AP1000 is a highly advanced Generation 3 pressurized water reactor (PWR). The reactor control strategy is considerably different from today's Westinghouse PWRs. AP1000 reactor control utilizes mechanical shim (MShim) for reactivity control of load changes and core depletion in conjunction with boron reactivity shim. The system is designed for several days of power operation without changing the reactor coolant system boron concentration. This along with an advanced reactor coolant pump design allows for a chemistry and volume control system that is much simplified from present PWRs. Furthermore, the nuclear renaissance provides the opportunity for the application of innovative technology in the establishment of behaviors in the operation of the reactor instead of modifying already learned and established practices. The presentation of reactor peaking factor limits, limit monitoring, nuclear data and reactor operations predictive capabilities to the control room will reflect the advanced core, new operating strategy and the opportunity of the AP1000. The proven BEACONTM core monitoring and analysis system will be used in the AP1000. BEACON is presently implemented in over 60 PWRs of different vendor origin on five continents. It is licensed by the regulatory agencies of several countries to provide surveillance of core thermal margins. In addition it provides accurate reactor operations predictions for evolutions like start-up and load changes. The AP1000 implementation of BEACON will build on this extensive experience. From vanadium fixed in-core detector signals and other online plant data, BEACON will continuously update a core model implemented with an advanced 3D nodal code. From this core model, BEACON will measure reactor power distribution, reactivity and shutdown margin and provide information to control room display and alarm presentation systems on margin to thermal and shutdown reactivity limits. Specifically, BEACON will provide

  9. Using Latent Class Analysis to Identify Academic and Behavioral Risk Status in Elementary Students

    Science.gov (United States)

    King, Kathleen R.; Lembke, Erica S.; Reinke, Wendy M.

    2016-01-01

    Identifying classes of children on the basis of academic and behavior risk may have important implications for the allocation of intervention resources within Response to Intervention (RTI) and Multi-Tiered System of Support (MTSS) models. Latent class analysis (LCA) was conducted with a sample of 517 third grade students. Fall screening scores in…

  10. Identifying Skill Requirements for GIS Positions: A Content Analysis of Job Advertisements

    Science.gov (United States)

    Hong, Jung Eun

    2016-01-01

    This study identifies the skill requirements for geographic information system (GIS) positions, including GIS analysts, programmers/developers/engineers, specialists, and technicians, through a content analysis of 946 GIS job advertisements from 2007-2014. The results indicated that GIS job applicants need to possess high levels of GIS analysis…

  11. Twelve type 2 diabetes susceptibility loci identified through large-scale association analysis

    DEFF Research Database (Denmark)

    Voight, Benjamin F; Scott, Laura J; Steinthorsdottir, Valgerdur;

    2010-01-01

    By combining genome-wide association data from 8,130 individuals with type 2 diabetes (T2D) and 38,987 controls of European descent and following up previously unidentified meta-analysis signals in a further 34,412 cases and 59,925 controls, we identified 12 new T2D association signals with combi...

  12. Identifying sustainability issues using participatory SWOT analysis - A case study of egg production in the Netherlands

    NARCIS (Netherlands)

    Mollenhorst, H.; Boer, de I.J.M.

    2004-01-01

    The aim of this paper was to demonstrate how participatory strengths, weaknesses, opportunities and threats (SWOT) analysis can be used to identify relevant economic, ecological and societal (EES) issues for the assessment of sustainable development. This is illustrated by the case of egg production

  13. Bioinformatics analysis identifies several intrinsically disordered human E3 ubiquitin-protein ligases

    DEFF Research Database (Denmark)

    Boomsma, Wouter; Nielsen, Sofie V; Lindorff-Larsen, Kresten;

    2016-01-01

    conduct a bioinformatics analysis to examine >600 human and S. cerevisiae E3 ligases to identify enzymes that are similar to San1 in terms of function and/or mechanism of substrate recognition. An initial sequence-based database search was found to detect candidates primarily based on the homology of...

  14. Application of Genome-Wide Expression Analysis To Identify Molecular Markers Useful in Monitoring Industrial Fermentations

    OpenAIRE

    Higgins, Vincent J.; Rogers, Peter J.; Dawes, Ian W.

    2003-01-01

    Genome-wide expression analysis of an industrial strain of Saccharomyces cerevisiae identified the YOR387c and YGL258w homologues as highly inducible in zinc-depleted conditions. Induction was specific for zinc deficiency and was dependent on Zap1p. The results indicate that these sequences may be valuable molecular markers for detecting zinc deficiency in industrial fermentations.

  15. Genome-wide association scan meta-analysis identifies three loci influencing adiposity and fat distribution

    NARCIS (Netherlands)

    C.M. Lindgren (Cecilia); I.M. Heid (Iris); J.C. Randall (Joshua); C. Lamina (Claudia); V. Steinthorsdottir (Valgerdur); L. Qi (Lu); E.K. Speliotes (Elizabeth); G. Thorleifsson (Gudmar); C.J. Willer (Cristen); B.M. Herrera (Blanca); A.U. Jackson (Anne); N. Lim (Noha); P. Scheet (Paul); N. Soranzo (Nicole); N. Amin (Najaf); Y.S. Aulchenko (Yurii); J.C. Chambers (John); A. Drong (Alexander); J. Luan; H.N. Lyon (Helen); F. Rivadeneira Ramirez (Fernando); S. Sanna (Serena); N. Timpson (Nicholas); M.C. Zillikens (Carola); H.Z. Jing; P. Almgren (Peter); S. Bandinelli (Stefania); A.J. Bennett (Amanda); R.N. Bergman (Richard); L.L. Bonnycastle (Lori); S. Bumpstead (Suzannah); S.J. Chanock (Stephen); L. Cherkas (Lynn); P.S. Chines (Peter); L. Coin (Lachlan); C. Cooper (Charles); G. Crawford (Gabe); A. Doering (Angela); A. Dominiczak (Anna); A.S.F. Doney (Alex); S. Ebrahim (Shanil); P. Elliott (Paul); M.R. Erdos (Michael); K. Estrada Gil (Karol); L. Ferrucci (Luigi); G. Fischer (Guido); N.G. Forouhi (Nita); C. Gieger (Christian); H. Grallert (Harald); C.J. Groves (Christopher); S.M. Grundy (Scott); C. Guiducci (Candace); D. Hadley (David); A. Hamsten (Anders); A.S. Havulinna (Aki); A. Hofman (Albert); R. Holle (Rolf); J.W. Holloway (John); T. Illig (Thomas); B. Isomaa (Bo); L.C. Jacobs (Leonie); K. Jameson (Karen); P. Jousilahti (Pekka); F. Karpe (Fredrik); J. Kuusisto (Johanna); J. Laitinen (Jaana); G.M. Lathrop (Mark); D.A. Lawlor (Debbie); M. Mangino (Massimo); W.L. McArdle (Wendy); T. Meitinger (Thomas); M.A. Morken (Mario); A.P. Morris (Andrew); P. Munroe (Patricia); N. Narisu (Narisu); A. Nordström (Anna); B.A. Oostra (Ben); C.N.A. Palmer (Colin); F. Payne (Felicity); J. Peden (John); I. Prokopenko (Inga); F. Renström (Frida); A. Ruokonen (Aimo); V. Salomaa (Veikko); M.S. Sandhu (Manjinder); L.J. Scott (Laura); A. Scuteri (Angelo); K. Silander (Kaisa); K. Song (Kijoung); X. Yuan (Xin); H.M. Stringham (Heather); A.J. Swift (Amy); T. Tuomi (Tiinamaija); M. Uda (Manuela); P. Vollenweider (Peter); G. Waeber (Gérard); C. Wallace (Chris); G.B. Walters (Bragi); M.N. Weedon (Michael); J.C.M. Witteman (Jacqueline); C. Zhang (Cuilin); M. Caulfield (Mark); F.S. Collins (Francis); G.D. Smith; I.N.M. Day (Ian); P.W. Franks (Paul); A.T. Hattersley (Andrew); F.B. Hu (Frank); M.R. Jarvelin; A. Kong (Augustine); J.S. Kooner (Jaspal); M. Laakso (Markku); E. Lakatta (Edward); V. Mooser (Vincent); L. Peltonen (Leena Johanna); N.J. Samani (Nilesh); T.D. Spector (Timothy); D.P. Strachan (David); T. Tanaka (Toshiko); J. Tuomilehto (Jaakko); A.G. Uitterlinden (André); P. Tikka-Kleemola (Päivi); N.J. Wareham (Nick); H. Watkins (Hugh); D. Waterworth (Dawn); M. Boehnke (Michael); P. Deloukas (Panagiotis); L. Groop (Leif); D.J. Hunter (David); U. Thorsteinsdottir (Unnur); D. Schlessinger (David); H.E. Wichmann (Erich); T.M. Frayling (Timothy); G.R. Abecasis (Gonçalo); J.N. Hirschhorn (Joel); R.J.F. Loos (Ruth); J-A. Zwart (John-Anker); K.L. Mohlke (Karen); I. Barroso (Inês); M.I. McCarthy (Mark)

    2009-01-01

    textabstractTo identify genetic loci influencing central obesity and fat distribution, we performed a meta-analysis of 16 genome-wide association studies (GWAS, N = 38,580) informative for adult waist circumference (WC) and waist-hip ratio (WHR). We selected 26 SNPs for follow-up, for which the evid

  16. Identifying Contingency Requirements using Obstacle Analysis on an Unpiloted Aerial Vehicle

    Science.gov (United States)

    Lutz, Robyn R.; Nelson, Stacy; Patterson-Hine, Ann; Frost, Chad R.; Tal, Doron

    2005-01-01

    This paper describes experience using Obstacle Analysis to identify contingency requirements on an unpiloted aerial vehicle. A contingency is an operational anomaly, and may or may not involve component failure. The challenges to this effort were: ( I ) rapid evolution of the system while operational, (2) incremental autonomy as capabilities were transferred from ground control to software control and (3) the eventual safety-criticality of such systems as they begin to fly over populated areas. The results reported here are preliminary but show that Obstacle Analysis helped (1) identify new contingencies that appeared as autonomy increased; (2) identify new alternatives for handling both previously known and new contingencies; and (3) investigate the continued validity of existing software requirements for contingency handling. Since many mobile, intelligent systems are built using a development process that poses the same challenges, the results appear to have applicability to other similar systems.

  17. Identifying the "Right Stuff": An Exploration-Focused Astronaut Job Analysis

    Science.gov (United States)

    Barrett, J. D.; Holland, A. W.; Vessey, W. B.

    2015-01-01

    Industrial and organizational (I/O) psychologists play a key role in NASA astronaut candidate selection through the identification of the competencies necessary to successfully engage in the astronaut job. A set of psychosocial competencies, developed by I/O psychologists during a prior job analysis conducted in 1996 and updated in 2003, were identified as necessary for individuals working and living in the space shuttle and on the International Space Station (ISS). This set of competencies applied to the space shuttle and applies to current ISS missions, but may not apply to longer-duration or long-distance exploration missions. With the 2015 launch of the first 12- month ISS mission and the shift in the 2020s to missions beyond low earth orbit, the type of missions that astronauts will conduct and the environment in which they do their work will change dramatically, leading to new challenges for these crews. To support future astronaut selection, training, and research, I/O psychologists in NASA's Behavioral Health and Performance (BHP) Operations and Research groups engaged in a joint effort to conduct an updated analysis of the astronaut job for current and future operations. This project will result in the identification of behavioral competencies critical to performing the astronaut job, along with relative weights for each of the identified competencies, through the application of job analysis techniques. While this job analysis is being conducted according to job analysis best practices, the project poses a number of novel challenges. These challenges include the need to identify competencies for multiple mission types simultaneously, to evaluate jobs that have no incumbents as they have never before been conducted, and working with a very limited population of subject matter experts. Given these challenges, under the guidance of job analysis experts, we used the following methods to conduct the job analysis and identify the key competencies for current and

  18. Development of whole core thermal-hydraulic analysis program ACT. 4. Simplified fuel assembly model and parallelization by MPI

    International Nuclear Information System (INIS)

    A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including the effect of the flow between wrapper-tube walls (inter-wrapper flow) under various reactor operation conditions. As appropriate boundary conditions in addition to a detailed modeling of the core are essential for accurate simulations of in-core thermal hydraulics, ACT consists of not only fuel assembly and inter-wrapper flow analysis modules but also a heat transport system analysis module that gives response of the plant dynamics to the core model. This report describes incorporation of a simplified model to the fuel assembly analysis module and program parallelization by a message passing method toward large-scale simulations. ACT has a fuel assembly analysis module which can simulate a whole fuel pin bundle in each fuel assembly of the core and, however, it may take much CPU time for a large-scale core simulation. Therefore, a simplified fuel assembly model that is thermal-hydraulically equivalent to the detailed one has been incorporated in order to save the simulation time and resources. This simplified model is applied to several parts of fuel assemblies in a core where the detailed simulation results are not required. With regard to the program parallelization, the calculation load and the data flow of ACT were analyzed and the optimum parallelization has been done including the improvement of the numerical simulation algorithm of ACT. Message Passing Interface (MPI) is applied to data communication between processes and synchronization in parallel calculations. Parallelized ACT was verified through a comparison simulation with the original one. In addition to the above works, input manuals of the core analysis module and the heat transport system analysis module have been prepared. (author)

  19. A finite element thermal analysis of various dowel and core materials

    Directory of Open Access Journals (Sweden)

    Shanti Varghese

    2012-01-01

    Conclusion: Non-metallic dowel and core materials such as fibre reinforced composite dowels (FRC generate greater stress than metallic dowel and core materials. This emphasized the preferable use of the metallic dowel and core materials in the oral environment.

  20. Development of spectral history methods for pin-by-pin core analysis method using three-dimensional direct response matrix

    International Nuclear Information System (INIS)

    Spectral history methods for pin-by-pin core analysis method using the three-dimensional direct response matrix have been developed. The direct response matrix is formalized by four sub-response matrices in order to respond to a core eigenvalue k and thus can be recomposed at each outer iteration in the core analysis. For core analysis, it is necessary to take into account the burn-up effect related to spectral history. One of the methods is to evaluate the nodal burn-up spectrum obtained using the out-going neutron current. The other is to correct the fuel rod neutron production rates obtained the pin-by-pin correction. These spectral history methods were tested in a heterogeneous system. The test results show that the neutron multiplication factor error can be reduced by half during burn-up, the nodal neutron production rates errors can be reduced by 30% or more. The root-mean-square differences between the relative fuel rod neutron production rate distributions can be reduced within 1.1% error. This means that these methods can accurately reflect the effects of intra- and inter-assembly heterogeneities during burn-up and can be used for core analysis. Core analysis with the DRM method was carried out for an ABWR quarter core and it was found that both thermal power and coolant-flow distributions were smoothly converged. (authors)

  1. Research fronts analysis : A bibliometric to identify emerging fields of research

    Science.gov (United States)

    Miwa, Sayaka; Ando, Satoko

    Research fronts analysis identifies emerging areas of research through observing co-clustering in highly-cited papers. This article introduces the concept of research fronts analysis, explains its methodology and provides case examples. It also demonstrates developing research fronts in Japan by looking at the past winners of Thomson Reuters Research Fronts Awards. Research front analysis is currently being used by the Japanese government to determine new trends in science and technology. Information professionals can also utilize this bibliometric as a research evaluation tool.

  2. Improving performance portability for GPU-specific OpenCL kernels on multi-core/many-core CPUs by analysis-based transformations*#

    Institute of Scientific and Technical Information of China (English)

    Mei WEN; Da-fei HUANG; Chang-qing XUN; Dong CHEN

    2015-01-01

    OpenCL is an open heterogeneous programming framework. Although OpenCL programs are func-tionally portable, they do not provide performance portability, so code transformation often plays an irreplaceable role. When adapting GPU-specifi c OpenCL kernels to run on multi-core/many-core CPUs, coarsening the thread granularity is necessary and thus has been extensively used. However, locality concerns exposed in GPU-specifi c OpenCL code are usually inherited without analysis, which may give side-effects on the CPU performance. Typi-cally, the use of OpenCL’s local memory on multi-core/many-core CPUs may lead to an opposite performance effect, because local-memory arrays no longer match well with the hardware and the associated synchronizations are costly. To solve this dilemma, we actively analyze the memory access patterns using array-access descriptors derived from GPU-specifi c kernels, which can thus be adapted for CPUs by (1) removing all the unwanted local-memory arrays together with the obsolete barrier statements and (2) optimizing the coalesced kernel code with vectorization and locality re-exploitation. Moreover, we have developed an automated tool chain that makes this transformation of GPU-specifi c OpenCL kernels into a CPU-friendly form, which is accompanied with a scheduler that forms a new OpenCL runtime. Experiments show that the automated transformation can improve OpenCL kernel performance on a multi-core CPU by an average factor of 3.24. Satisfactory performance improvements are also achieved on Intel’s many-integrated-core coprocessor. The resultant performance on both architectures is better than or comparable with the corresponding OpenMP performance.

  3. Metabolites production improvement by identifying minimal genomes and essential genes using flux balance analysis.

    Science.gov (United States)

    Salleh, Abdul Hakim Mohamed; Mohamad, Mohd Saberi; Deris, Safaai; Illias, Rosli Md

    2015-01-01

    With the advancement in metabolic engineering technologies, reconstruction of the genome of host organisms to achieve desired phenotypes can be made. However, due to the complexity and size of the genome scale metabolic network, significant components tend to be invisible. We proposed an approach to improve metabolite production that consists of two steps. First, we find the essential genes and identify the minimal genome by a single gene deletion process using Flux Balance Analysis (FBA) and second by identifying the significant pathway for the metabolite production using gene expression data. A genome scale model of Saccharomyces cerevisiae for production of vanillin and acetate is used to test this approach. The result has shown the reliability of this approach to find essential genes, reduce genome size and identify production pathway that can further optimise the production yield. The identified genes and pathways can be extendable to other applications especially in strain optimisation. PMID:26489144

  4. New families of human regulatory RNA structures identified by comparative analysis of vertebrate genomes

    DEFF Research Database (Denmark)

    Parker, Brian John; Moltke, Ida; Roth, Adam;

    2011-01-01

    comparative method, EvoFam, for genome-wide identification of families of regulatory RNA structures, based on primary sequence and secondary structure similarity. We apply EvoFam to a 41-way genomic vertebrate alignment. Genome-wide, we identify 220 human, high-confidence families outside protein......-coding regions comprising 725 individual structures, including 48 families with known structural RNA elements. Known families identified include both noncoding RNAs, e.g., miRNAs and the recently identified MALAT1/MEN β lincRNA family; and cis-regulatory structures, e.g., iron-responsive elements. We also...... identify tens of new families supported by strong evolutionary evidence and other statistical evidence, such as GO term enrichments. For some of these, detailed analysis has led to the formulation of specific functional hypotheses. Examples include two hypothesized auto-regulatory feedback mechanisms: one...

  5. Gene expression meta-analysis identifies metastatic pathways and transcription factors in breast cancer

    International Nuclear Information System (INIS)

    Metastasis is believed to progress in several steps including different pathways but the determination and understanding of these mechanisms is still fragmentary. Microarray analysis of gene expression patterns in breast tumors has been used to predict outcome in recent studies. Besides classification of outcome, these global expression patterns may reflect biological mechanisms involved in metastasis of breast cancer. Our purpose has been to investigate pathways and transcription factors involved in metastasis by use of gene expression data sets. We have analyzed 8 publicly available gene expression data sets. A global approach, 'gene set enrichment analysis' as well as an approach focusing on a subset of significantly differently regulated genes, GenMAPP, has been applied to rank pathway gene sets according to differential regulation in metastasizing tumors compared to non-metastasizing tumors. Meta-analysis has been used to determine overrepresentation of pathways and transcription factors targets, concordant deregulated in metastasizing breast tumors, in several data sets. The major findings are up-regulation of cell cycle pathways and a metabolic shift towards glucose metabolism reflected in several pathways in metastasizing tumors. Growth factor pathways seem to play dual roles; EGF and PDGF pathways are decreased, while VEGF and sex-hormone pathways are increased in tumors that metastasize. Furthermore, migration, proteasome, immune system, angiogenesis, DNA repair and several signal transduction pathways are associated to metastasis. Finally several transcription factors e.g. E2F, NFY, and YY1 are identified as being involved in metastasis. By pathway meta-analysis many biological mechanisms beyond major characteristics such as proliferation are identified. Transcription factor analysis identifies a number of key factors that support central pathways. Several previously proposed treatment targets are identified and several new pathways that may

  6. A cross-species genetic analysis identifies candidate genes for mouse anxiety and human bipolar disorder

    Directory of Open Access Journals (Sweden)

    David G Ashbrook

    2015-07-01

    Full Text Available Bipolar disorder (BD is a significant neuropsychiatric disorder with a lifetime prevalence of ~1%. To identify genetic variants underlying BD genome-wide association studies (GWAS have been carried out. While many variants of small effect associated with BD have been identified few have yet been confirmed, partly because of the low power of GWAS due to multiple comparisons being made. Complementary mapping studies using murine models have identified genetic variants for behavioral traits linked to BD, often with high power, but these identified regions often contain too many genes for clear identification of candidate genes. In the current study we have aligned human BD GWAS results and mouse linkage studies to help define and evaluate candidate genes linked to BD, seeking to use the power of the mouse mapping with the precision of GWAS. We use quantitative trait mapping for open field test and elevated zero maze data in the largest mammalian model system, the BXD recombinant inbred mouse population, to identify genomic regions associated with these BD-like phenotypes. We then investigate these regions in whole genome data from the Psychiatric Genomics Consortium’s bipolar disorder GWAS to identify candidate genes associated with BD. Finally we establish the biological relevance and pathways of these genes in a comprehensive systems genetics analysis.We identify four genes associated with both mouse anxiety and human BD. While TNR is a novel candidate for BD, we can confirm previously suggested associations with CMYA5, MCTP1 and RXRG. A cross-species, systems genetics analysis shows that MCTP1, RXRG and TNR coexpress with genes linked to psychiatric disorders and identify the striatum as a potential site of action. CMYA5, MCTP1, RXRG and TNR are associated with mouse anxiety and human BD. We hypothesize that MCTP1, RXRG and TNR influence intercellular signaling in the striatum.

  7. A cross-species genetic analysis identifies candidate genes for mouse anxiety and human bipolar disorder.

    Science.gov (United States)

    Ashbrook, David G; Williams, Robert W; Lu, Lu; Hager, Reinmar

    2015-01-01

    Bipolar disorder (BD) is a significant neuropsychiatric disorder with a lifetime prevalence of ~1%. To identify genetic variants underlying BD genome-wide association studies (GWAS) have been carried out. While many variants of small effect associated with BD have been identified few have yet been confirmed, partly because of the low power of GWAS due to multiple comparisons being made. Complementary mapping studies using murine models have identified genetic variants for behavioral traits linked to BD, often with high power, but these identified regions often contain too many genes for clear identification of candidate genes. In the current study we have aligned human BD GWAS results and mouse linkage studies to help define and evaluate candidate genes linked to BD, seeking to use the power of the mouse mapping with the precision of GWAS. We use quantitative trait mapping for open field test and elevated zero maze data in the largest mammalian model system, the BXD recombinant inbred mouse population, to identify genomic regions associated with these BD-like phenotypes. We then investigate these regions in whole genome data from the Psychiatric Genomics Consortium's bipolar disorder GWAS to identify candidate genes associated with BD. Finally we establish the biological relevance and pathways of these genes in a comprehensive systems genetics analysis. We identify four genes associated with both mouse anxiety and human BD. While TNR is a novel candidate for BD, we can confirm previously suggested associations with CMYA5, MCTP1, and RXRG. A cross-species, systems genetics analysis shows that MCTP1, RXRG, and TNR coexpress with genes linked to psychiatric disorders and identify the striatum as a potential site of action. CMYA5, MCTP1, RXRG, and TNR are associated with mouse anxiety and human BD. We hypothesize that MCTP1, RXRG, and TNR influence intercellular signaling in the striatum. PMID:26190982

  8. Analysis of promoter regions of co-expressed genes identified by microarray analysis

    Directory of Open Access Journals (Sweden)

    Höglund Mattias

    2006-08-01

    Full Text Available Abstract Background The use of global gene expression profiling to identify sets of genes with similar expression patterns is rapidly becoming a widespread approach for understanding biological processes. A logical and systematic approach to study co-expressed genes is to analyze their promoter sequences to identify transcription factors that may be involved in establishing specific profiles and that may be experimentally investigated. Results We introduce promoter clustering i.e. grouping of promoters with respect to their high scoring motif content, and show that this approach greatly enhances the identification of common and significant transcription factor binding sites (TFBS in co-expressed genes. We apply this method to two different dataset, one consisting of micro array data from 108 leukemias (AMLs and a second from a time series experiment, and show that biologically relevant promoter patterns may be obtained using phylogenetic foot-printing methodology. In addition, we also found that 15% of the analyzed promoter regions contained transcription factors start sites for additional genes transcribed in the opposite direction. Conclusion Promoter clustering based on global promoter features greatly improve the identification of shared TFBS in co-expressed genes. We believe that the outlined approach may be a useful first step to identify transcription factors that contribute to specific features of gene expression profiles.

  9. Analysis of two-phase flow and boiling heat transfer in inclined channel of core-catcher

    International Nuclear Information System (INIS)

    Passive Corium Cooling System (CCS) provides a function of ex-vessel debris cooling and molten core stabilization during a severe accident. CCS features inclined cooling channels arranged axi-symmetrically below the core-catcher basin. In order to estimate the coolability of the inclined cooling channel, it is indispensable to identify the flow pattern of the two-phase flow in the cooling channel. Several former studies for the two-phase flow pattern in the inclined channel are referred. Taitel and Dukler (1976) developed a prediction method of the flow pattern transition in horizontal and near horizontal tubes. Barnea et al. (1980) showed the flow pattern map of upward flow with 10 degrees inclination. Sakaguti et al. (1996) observed the two-phase flow patterns in the horizontal pipe connected with slightly upward pipe, in which the flow pattern in the pipe with a bending part was expressed by the combination of a basic flow pattern and some auxiliary flow patterns. Then we investigated these studies In order to identify the flow patterns observed in the inclined cooling channel of CCS. Furthermore we experimentally observed the flow patterns in the inclined cooling channel with various inlet conditions. As a result of the investigation and observation, typical flow patterns in the inclined cooling channel were identified. Two typical flow patterns were observed depending on the steam flow rate, one of which is 'elongated bubble 'flow, and the other is 'churn with collapsing backward and upward slug 'flow The flow and heat transfer in the inclined channel of CCS is analyzed by using a two-phase analysis code employing two-fluid model in which the constitutive equations for the two-phase flow in inclined channels are incorporated. That is, drift flux parameter for each of the elongated bubble flow, and the churn with collapsing backward and upward slug flow are incorporated to the two-phase analysis code, which are based on the rising velocity of the long bubble in

  10. Lumped parameter analysis of Pb-Bi cooled fast reactor PEACER core using MATRA

    International Nuclear Information System (INIS)

    The PEACER (Proliferation-resistant Environment-friendly Accident-tolerant Continuable energy Economical Reactor) system is under study to transmute long-lived fission products and actinides as well as to produce electricity. It is important to keep the temperature of the reactor core structures under certain criteria in order to prevent damage of fuel materials which can advance to severe situations such as radiation leakage, and even meltdown of the fuel. This study intends to examine the liquid metal coolant behavior along the PEACER fuel channels and to find out whether the given heat flux profiles and geometrical arrangement of the fuel rods yield reasonable flow distribution during the nominal operation by using subchannel approach. The subchannel analysis of the fuel assembly under nominal operation condition was performed using MATRA (Multi-channel Analyzer for Transient and steady-state in Rod Arrays). The result showed that the input data based on the current design of the PEACER core yielded reliable results satisfying the thermla and mechanical design limits. Typical results obtained include the hydrodynamic conditions of Pb-Bi in subchannels and the thermodynamic states of the core structures. (author)

  11. Design and Dynamic Analysis of Air-core Coil type Linear DC Motor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Gyu Hong; Hong, Jung Pyo; Kim, Gyu Tak [Chang Won National University (Korea); Ha, Kyeun Su; Jung, Joong Gi; Im, Tae Bin [Korea Electronics Technology Institute (Korea)

    2000-03-01

    This paper proposes a technique to design of air-core coil type Linear DC Motor(LDM) by using Equivalent Magnetizing Current(EMC) method and has performed its dynamic analysis. The magnetic flux density differ in accordance with airgap position due to the difference of mechanical and magnetic air gap length and the coil shape has an influence on the thrust. Therefore, the analysis of magnetic field due to the magnets is carried out by EMC. The phenomena according to the various coil shape under the same Magneto Motive Force(MMF) has been analyzed and its result is applied to the design process. The appropriateness of the proposed technique is confirmed by Finite Element Method(FEM) and its dynamic analysis is carried out from the coupling of the electrical circuit equation and mechanical kinetic equation. (author). 8 refs., 13 figs., 2 tabs.

  12. Optimization analysis of the nuclear fuel cycle transition to the last core

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L.; Blanco, J. [Union Fenosa Generacion, Madrid (Spain)

    2001-07-01

    The Zorita NPP was the first Spanish commercial nuclear reactor connected to the grid. It is a 160 MW one loop PWR, Westinghouse design, owned by UFG, in operation since 1968. The configuration of the reactor core is based on 69 fuel elements type 14 x 14, the standard reload of the present equilibrium cycle being based on 16 fuel elements with 3.6% enrichment in {sup 235}U. In order to properly plan the nuclear fuel management of the transition cycles to its end of life, presently foreseen by 2008, an based on the non-reprocessing option required by the policy of the Spanish Administration, a technical-economical optimization analysis has been performed. As a result, a fuel management strategy has been defined looking for getting simultaneously the minimum integral fuel cost of the transition from the present equilibrium cycle to the last core, as well as the minimum residual worth of the fuel remaining in the core after the final outage. Based on the ''lessons learned'' derived from the study, the time margin for the decision making has been determined, and a planning of the nuclear fuel supply for the transition reloads, specifying both the number of fuel elements and their enrichment in {sup 235}U, as been prepared. Finally, based on the calculated economical worth of the partially burned fuel of the last core, after the end of its operation cycle, a financial cover for yearly compensation from now on of the foreseen final lost has been elaborated. Most of the conceptual conclusions obtained are applicable to the other commercial nuclear reactors in operation owned by UFG, so that they are understood to be of general interest and broad application to commercial PWR. (author)

  13. Optimization analysis of the nuclear fuel cycle transition to the last core

    International Nuclear Information System (INIS)

    The Zorita NPP was the first Spanish commercial nuclear reactor connected to the grid. It is a 160 MW one loop PWR, Westinghouse design, owned by UFG, in operation since 1968. The configuration of the reactor core is based on 69 fuel elements type 14 x 14, the standard reload of the present equilibrium cycle being based on 16 fuel elements with 3.6% enrichment in 235U. In order to properly plan the nuclear fuel management of the transition cycles to its end of life, presently foreseen by 2008, an based on the non-reprocessing option required by the policy of the Spanish Administration, a technical-economical optimization analysis has been performed. As a result, a fuel management strategy has been defined looking for getting simultaneously the minimum integral fuel cost of the transition from the present equilibrium cycle to the last core, as well as the minimum residual worth of the fuel remaining in the core after the final outage. Based on the ''lessons learned'' derived from the study, the time margin for the decision making has been determined, and a planning of the nuclear fuel supply for the transition reloads, specifying both the number of fuel elements and their enrichment in 235U, as been prepared. Finally, based on the calculated economical worth of the partially burned fuel of the last core, after the end of its operation cycle, a financial cover for yearly compensation from now on of the foreseen final lost has been elaborated. Most of the conceptual conclusions obtained are applicable to the other commercial nuclear reactors in operation owned by UFG, so that they are understood to be of general interest and broad application to commercial PWR. (author)

  14. Safety analysis for core conversion (from HEU to LEU) of Pakistan research reactor-2 (PARR-2)

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, Ishtiaq Hussain, E-mail: ishtiaq@pinstech.org.p [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan); Pervez, Showket [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2010-01-15

    PARR-2 (Pakistan Research Reactor-2), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. The present report presents thermal hydraulic and safety analyses of PARR-2 using existing HEU fuel as well as proposed LEU fuel. Presently, the core is comprised of 90.2% enriched UAl{sub 4}-Al fuel. There are 344 fuel pins of 5.5 mm diameter. The core has a total of 994.8 g of U{sup 235}. Standard computer code PARET/ANL (version 1992) was employed to perform steady-state and transient analyses. Various parameters were computed, which included: coolant outlet, maximum clad surface and maximum fuel centerline temperatures; and peak power and corresponding peak core temperatures resulting from a transient initiated by 4 mK positive reactivity insertion. Results were compared with the reported data in Final Safety Analysis Report (FSAR). It was found that the PARET results were in reasonable agreement with the manufacturer's results. Calculations were also carried out for the proposed LEU core with two suggested fuel pin sizes (5.5 mm and 5.1 mm diameter with 12.6% and 12.3% enrichment, respectively). Comparison of the LEU results with the existing HEU fuel has been made and discussed.

  15. Microarray analysis in clinical oncology: pre-clinical optimization using needle core biopsies from xenograft tumors

    International Nuclear Information System (INIS)

    labeled, would generate representative array profiles compared to larger excisional biopsy material. In this analysis correlation coefficients were obtained ranging from 0.750–0.834 between U251 biopsy cores and excised tumors, and 0.812–0.846 between DU145 biopsy cores and excised tumors. These data suggest that needle core biopsies can be used as reliable tissue samples for tumor microarray analysis after linear amplification and either indirect or direct labeling of the starting RNA

  16. Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices

    International Nuclear Information System (INIS)

    This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. The mean core damage frequency is 4.5E-6 with 5% and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively. Station blackout type accidents (loss of all ac power) contributed about 46% of the core damage frequency with Anticipated Transient Without Scram (ATWS) accidents contributing another 42%. The numerical results are driven by loss of offsite power, transients with the power conversion system initially available operator errors, and mechanical failure to scram. 13 refs., 345 figs., 171 tabs

  17. Numerical analysis of flow distribution at the reactor core inlet of Qinshan phase-II reactor

    International Nuclear Information System (INIS)

    To improve the thermal hydraulic performance of pressurized water reactor (PWR), it is of great importance to obtain reliable flow distribution data at the core inlet. Through computational fluid dynamics (CFD) analysis the flow field in a 1/4 scale model of the 600 MW PWR was worked out numerically. The sensitivity analysis focused on factors such as the lower plenum geometry and 500, 880, 900 m3/h of flow rate for an inlet nozzle, respectively. The results provided a deep understanding of the flow behavior concerning the pressure vessel of the PWR. Numerical results indicated that flow distribution at the core inlet is not sensitive to the inlet nozzle flow rate under two-loop operation mode. Moreover, flow characteristics inside the pressure vessel under single loop operation mode differ significantly from those under two-loop operation mode. A dimensionless flow distribution subfactor of 0.05 is in good agreement with the prototype design. The validity of applying CFD methods in flow distribution for nuclear reactor is verified. The analysis results are useful for the thermal hydraulic design of the PWR. (authors)

  18. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  19. Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices

    Energy Technology Data Exchange (ETDEWEB)

    Kolaczkowski, A.M.; Cramond, W.R.; Sype, T.T.; Maloney, K.J.; Wheeler, T.A.; Daniel, S.L. (Science Applications International Corp., Albuquerque, NM (USA); Sandia National Labs., Albuquerque, NM (USA))

    1989-08-01

    This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. The mean core damage frequency is 4.5E-6 with 5% and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively. Station blackout type accidents (loss of all ac power) contributed about 46% of the core damage frequency with Anticipated Transient Without Scram (ATWS) accidents contributing another 42%. The numerical results are driven by loss of offsite power, transients with the power conversion system initially available operator errors, and mechanical failure to scram. 13 refs., 345 figs., 171 tabs.

  20. Monte Carlo Few-Group Constant Generation for CANDU 6 Core Analysis

    OpenAIRE

    Seung Yeol Yoo; Hyung Jin Shim; Chang Hyo Kim

    2015-01-01

    The current neutronics design methodology of CANDU-PHWRs based on the two-step calculations requires determining not only homogenized two-group constants for ordinary fuel bundle lattice cells by the WIMS-AECL lattice cell code but also incremental two-group constants arising from the penetration of control devices into the fuel bundle cells by a supercell analysis code like MULTICELL or DRAGON. As an alternative way to generate the two-group constants necessary for the CANDU-PHWR core analys...

  1. Genome-wide detection and analysis of hippocampus core promoters using DeepCAGE

    DEFF Research Database (Denmark)

    Valen, Eivind; Pascarella, Giovanni; Chalk, Alistair;

    2009-01-01

    given tissue. Here, we present a new method for high-throughput sequencing of 5' cDNA tags-DeepCAGE: merging the Cap Analysis of Gene Expression method with ultra-high-throughput sequence technology. We apply DeepCAGE to characterize 1.4 million sequenced TSS from mouse hippocampus and reveal a wealth...... of novel core promoters that are preferentially used in hippocampus: This is the most comprehensive promoter data set for any tissue to date. Using these data, we present evidence indicating a key role for the Arnt2 transcription factor in hippocampus gene regulation. DeepCAGE can also detect...

  2. Simultaneous stable isotope analysis of methane and nitrous oxide on ice core samples

    Directory of Open Access Journals (Sweden)

    C. J. Sapart

    2011-07-01

    Full Text Available Methane and nitrous oxide are important greenhouse gases which show a strong increase in atmospheric mixing ratios since pre-industrial time as well as large variations during past climate changes. The understanding of their biogeochemical cycles can be improved using stable isotope analysis. However, high-precision isotope measurements on air trapped in ice cores are challenging because of the high susceptibility to contamination and fractionation.

    Here, we present a dry extraction system for combined CH4 and N2O stable isotope analysis from ice core air, using an ice grating device. The system allows simultaneous analysis of δD(CH4 or δ13C(CH4, together with δ15N(N2O, δ18O(N2O and δ15N(NO+fragment on a single ice core sample, using two isotope mass spectrometry systems. The optimum quantity of ice for analysis is about 600g with typical "Holocene" mixing ratios for CH4 and N2O. In this case, the reproducibility (1σ is 2.1 ‰ for δD(CH4, 0.18 ‰ for δ13C(CH4, 0.51 ‰ for δ15N(N2O, 0.69 ‰ for δ18O(N2O and 1.12 ‰ for δ15N(NO+fragment. For smaller amounts of ice the standard deviation increases, particularly for N2O isotopologues. For both gases, small-scale intercalibrations using air and/or ice samples have been carried out with other institutes that are currently involved in isotope measurements of ice core air. Significant differences are shown between the calibration scales, but those offsets are consistent and can be corrected for.

  3. Simultaneous stable isotope analysis of methane and nitrous oxide on ice core samples

    Directory of Open Access Journals (Sweden)

    C. J. Sapart

    2011-12-01

    Full Text Available Methane and nitrous oxide are important greenhouse gases which show a strong increase in atmospheric mixing ratios since pre-industrial time as well as large variations during past climate changes. The understanding of their biogeochemical cycles can be improved using stable isotope analysis. However, high-precision isotope measurements on air trapped in ice cores are challenging because of the high susceptibility to contamination and fractionation.

    Here, we present a dry extraction system for combined CH4 and N2O stable isotope analysis from ice core air, using an ice grating device. The system allows simultaneous analysis of δD(CH4 or δ13C(CH4, together with δ15N(N2O, δ18O(N2O and δ15N(NO+ fragment on a single ice core sample, using two isotope mass spectrometry systems. The optimum quantity of ice for analysis is about 600 g with typical "Holocene" mixing ratios for CH4 and N2O. In this case, the reproducibility (1σ is 2.1‰ for δD(CH4, 0.18‰ for δ13C(CH4, 0.51‰ for δ15N(N2O, 0.69‰ for δ18O(N2O and 1.12‰ for δ15N(NO+ fragment. For smaller amounts of ice the standard deviation increases, particularly for N2O isotopologues. For both gases, small-scale intercalibrations using air and/or ice samples have been carried out in collaboration with other institutes that are currently involved in isotope measurements of ice core air. Significant differences are shown between the calibration scales, but those offsets are consistent and can therefore be corrected for.

  4. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  5. Whole-Core Thermal Analysis of Prismatic Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Nam Il; Kim, Min Hwan; Lim, Hong Sik; Jun, Ji Su; Jo, Chang Keun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    A new method for thermal analysis of prismatic fuel blocks in a very high temperature reactor (VHTR) was developed to overcome the demerits of computational fluid dynamics (CFD) and system calculations. The developed method solves three dimensional heat conduction in prismatic fuel blocks like a CFD code. For the fluid, however, the method adopts one-dimensional conservation equations like a system code. Such a combination enables significantly reduced computational efforts with reasonable computational accuracy. In this paper, the new method has been applied to whole core of PMR200 under full power operating conditions

  6. TRAC-BF1/NEM stability methodology for BWR core wide and regional stability analysis

    International Nuclear Information System (INIS)

    A time-series analysis stability methodology is presented based on the TRAC-BF1/NEM coupled code. The methodology presented has a potential application for BWR core-wide and regional stability studies allowed by the 3D capabilities of the code. The stability analysis is performed at two different levels: using the TRAC-BF1 point kinetics model and employing the three-dimensional neutronic transient capability of the NEM code. Point kinetics calculations show power fluctuations when white noise is applied to the inlet mass flow rate of each of the channel components. These fluctuations contain information about the system stability, and are subsequently studied with time-series analysis methods. The analysis performed showed that the reactor core has a low-frequency resonance typical of BWRs. Analysis of preliminary three-dimensional calculations indicates that the power fluctuations do not contain the typical resonance at low frequency. This fact may be related to the limitation of the thermal-hydraulic (T-H) feedback representation through the use of two-dimensional tables for the cross-sections needed for 3D kinetics calculations. The results suggest that a more accurate table look-up should be used, which includes a three-dimensional representation of the feedback parameters (namely, average fuel temperature, average moderator temperature, and void fraction of the T-H cell of interest). Further research is being conducted on improving the cross-section modeling methodology, used to feed the neutron kinetics code for both steady state and transient cases. Also a comprehensive analysis of the code transient solution is being conducted to investigate the nature of the weak dependence of the power response on T-H variations during the performed 3D stability transient calculations

  7. Development of core design/analysis technology for integral reactor; verification of SMART nuclear design by Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Hong, In Seob; Han, Beom Seok; Jeong, Jong Seong [Seoul National University, Seoul (Korea)

    2002-03-01

    The objective of this project is to verify neutronics characteristics of the SMART core design as to compare computational results of the MCNAP code with those of the MASTER code. To achieve this goal, we will analyze neutronics characteristics of the SMART core using the MCNAP code and compare these results with results of the MASTER code. We improved parallel computing module and developed error analysis module of the MCNAP code. We analyzed mechanism of the error propagation through depletion computation and developed a calculation module for quantifying these errors. We performed depletion analysis for fuel pins and assemblies of the SMART core. We modeled a 3-D structure of the SMART core and considered a variation of material compositions by control rods operation and performed depletion analysis for the SMART core. We computed control-rod worths of assemblies and a reactor core for operation of individual control-rod groups. We computed core reactivity coefficients-MTC, FTC and compared these results with computational results of the MASTER code. To verify error analysis module of the MCNAP code, we analyzed error propagation through depletion of the SMART B-type assembly. 18 refs., 102 figs., 36 tabs. (Author)

  8. Probabilistic approach to identify sensitive parameter distributions in multimedia pathway analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Kamboj, S.; Gnanapragasam, E.; LePoire, D.; Biwer, B. M.; Cheng, J.; Arnish, J.; Yu, C.; Chen, S. Y.; Mo, T.; Abu-Eid, R.; Thaggard, M.; Environmental Assessment; NRC

    2002-01-01

    Sensitive parameter distributions were identified with the use of probabilistic analysis in the RESRAD computer code. RESRAD is a multimedia pathway analysis code designed to evaluate radiological exposures resulting from radiological contamination in soil. The dose distribution was obtained by using a set of default parameter distribution/values. Most of the variations in the output dose distribution could be attributed to uncertainty in a small set of input parameters that could be considered as sensitive parameter distributions. The identification of the sensitive parameters is a first step in the prioritization of future research and information gathering. When site-specific parameter distribution/values are available for an actual site, the same process should be used with these site-specific data. Regression analysis used to identify sensitive parameters indicated that the dominant pathways depended on the radionuclide and source configurations. However, two parameter distributions were sensitive for many radionuclides: the external shielding factor when external exposure was the dominant pathway and the plant transfer factor when plant ingestion was the dominant pathway. No single correlation or regression coefficient can be used alone to identify sensitive parameters in all the cases. The coefficients are useful guides, but they have to be used in conjunction with other aids, such as scatter plots, and should undergo further analysis.

  9. Discovery of core biotic stress responsive genes in Arabidopsis by weighted gene co-expression network analysis.

    Science.gov (United States)

    Amrine, Katherine C H; Blanco-Ulate, Barbara; Cantu, Dario

    2015-01-01

    Intricate signal networks and transcriptional regulators translate the recognition of pathogens into defense responses. In this study, we carried out a gene co-expression analysis of all currently publicly available microarray data, which were generated in experiments that studied the interaction of the model plant Arabidopsis thaliana with microbial pathogens. This work was conducted to identify (i) modules of functionally related co-expressed genes that are differentially expressed in response to multiple biotic stresses, and (ii) hub genes that may function as core regulators of disease responses. Using Weighted Gene Co-expression Network Analysis (WGCNA) we constructed an undirected network leveraging a rich curated expression dataset comprising 272 microarrays that involved microbial infections of Arabidopsis plants with a wide array of fungal and bacterial pathogens with biotrophic, hemibiotrophic, and necrotrophic lifestyles. WGCNA produced a network with scale-free and small-world properties composed of 205 distinct clusters of co-expressed genes. Modules of functionally related co-expressed genes that are differentially regulated in response to multiple pathogens were identified by integrating differential gene expression testing with functional enrichment analyses of gene ontology terms, known disease associated genes, transcriptional regulators, and cis-regulatory elements. The significance of functional enrichments was validated by comparisons with randomly generated networks. Network topology was then analyzed to identify intra- and inter-modular gene hubs. Based on high connectivity, and centrality in meta-modules that are clearly enriched in defense responses, we propose a list of 66 target genes for reverse genetic experiments to further dissect the Arabidopsis immune system. Our results show that statistical-based data trimming prior to network analysis allows the integration of expression datasets generated by different groups, under different

  10. Discovery of core biotic stress responsive genes in Arabidopsis by weighted gene co-expression network analysis.

    Directory of Open Access Journals (Sweden)

    Katherine C H Amrine

    Full Text Available Intricate signal networks and transcriptional regulators translate the recognition of pathogens into defense responses. In this study, we carried out a gene co-expression analysis of all currently publicly available microarray data, which were generated in experiments that studied the interaction of the model plant Arabidopsis thaliana with microbial pathogens. This work was conducted to identify (i modules of functionally related co-expressed genes that are differentially expressed in response to multiple biotic stresses, and (ii hub genes that may function as core regulators of disease responses. Using Weighted Gene Co-expression Network Analysis (WGCNA we constructed an undirected network leveraging a rich curated expression dataset comprising 272 microarrays that involved microbial infections of Arabidopsis plants with a wide array of fungal and bacterial pathogens with biotrophic, hemibiotrophic, and necrotrophic lifestyles. WGCNA produced a network with scale-free and small-world properties composed of 205 distinct clusters of co-expressed genes. Modules of functionally related co-expressed genes that are differentially regulated in response to multiple pathogens were identified by integrating differential gene expression testing with functional enrichment analyses of gene ontology terms, known disease associated genes, transcriptional regulators, and cis-regulatory elements. The significance of functional enrichments was validated by comparisons with randomly generated networks. Network topology was then analyzed to identify intra- and inter-modular gene hubs. Based on high connectivity, and centrality in meta-modules that are clearly enriched in defense responses, we propose a list of 66 target genes for reverse genetic experiments to further dissect the Arabidopsis immune system. Our results show that statistical-based data trimming prior to network analysis allows the integration of expression datasets generated by different groups

  11. Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

    International Nuclear Information System (INIS)

    The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

  12. Improvement of supercomputing based core design process with parallel estimations and statistical analysis

    International Nuclear Information System (INIS)

    Several new challenges appeared to design new generation of sodium cooled Fast Reactor (SFR). These new reactors should resist, without severe consequence, to most critical accidents such as loss of cooling pump (ULOx transients). Another focus is to design a reactor easy to operate and cost effective. To achieve the above targets, many constraints and parameters have to be considered during the optimization process. The complexity of the process resides in the non intuitive interactions between different parameters. Classical design approach consists of sequential analysis of neutrons core shielding, fuel behavior, sub-assembly mechanics and thermal-hydraulics (core and system), usually performed iteratively by separated experts/services. This iterative scheme is mainly based on designer's knowledge and experiences. Retained options are neither optimum nor rigorously justified by macroscopic indicators. Moreover in case of new space parameters introduction, or new optimization constraints and objectives, designers have to perform new iterations to find a new compromise. This process is time consuming especially in the early stage of design when parameters are subject to frequent changes. The use of supercomputing capacity allows complementary approaches based on optimization algorithm. This way explores multi-physical design space parameters. Nowadays it is not possible to make an optimization directly with best estimates codes, too many evaluations are required. To emulate codes behavior we use advanced correlation technique (such as polynomial, neural network, Kriging...) based on learning data base. TRIAD (Tools for Reactor optimization Analysis and Design) software is developed at CEA and offers all features to develop and implement these methods for SFR core designers. (authors)

  13. Safety analysis for key design features of KALIMER with breakeven core

    International Nuclear Information System (INIS)

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term nuclear R and D Program. In this report, key safety design features are described and safety analyses results for typical ATWS accidents in the KALIMER design with breakeven core are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the event categorization and acceptance criteria for the KALIMER safety analysis are described in chapter 2. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER system response to the events. They are categorized as Bounding Events (BEs) because of their low probability of occurrence. In chapter 4, the performance analysis results of the KALIMER containment dome are described along with the HCDA accident scenario and source terms. The major containment parameters of peak pressure and peak temperature have been calculated using the CONTAIN-LMR code. Radiological consequence has been evaluated by the MACCS code. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using SCHAMBETA code developed in the framework of the modified bethe-tait method. Work energy potentials based arising from the sodium expansion as well as the isentropic fuel expansion are then calculated to evaluate the structural integrity of the reactor vessel, reactor internals and primary coolant system of KALIMER

  14. Meta-Analysis of Placental Transcriptome Data Identifies a Novel Molecular Pathway Related to Preeclampsia.

    Science.gov (United States)

    van Uitert, Miranda; Moerland, Perry D; Enquobahrie, Daniel A; Laivuori, Hannele; van der Post, Joris A M; Ris-Stalpers, Carrie; Afink, Gijs B

    2015-01-01

    Studies using the placental transcriptome to identify key molecules relevant for preeclampsia are hampered by a relatively small sample size. In addition, they use a variety of bioinformatics and statistical methods, making comparison of findings challenging. To generate a more robust preeclampsia gene expression signature, we performed a meta-analysis on the original data of 11 placenta RNA microarray experiments, representing 139 normotensive and 116 preeclamptic pregnancies. Microarray data were pre-processed and analyzed using standardized bioinformatics and statistical procedures and the effect sizes were combined using an inverse-variance random-effects model. Interactions between genes in the resulting gene expression signature were identified by pathway analysis (Ingenuity Pathway Analysis, Gene Set Enrichment Analysis, Graphite) and protein-protein associations (STRING). This approach has resulted in a comprehensive list of differentially expressed genes that led to a 388-gene meta-signature of preeclamptic placenta. Pathway analysis highlights the involvement of the previously identified hypoxia/HIF1A pathway in the establishment of the preeclamptic gene expression profile, while analysis of protein interaction networks indicates CREBBP/EP300 as a novel element central to the preeclamptic placental transcriptome. In addition, there is an apparent high incidence of preeclampsia in women carrying a child with a mutation in CREBBP/EP300 (Rubinstein-Taybi Syndrome). The 388-gene preeclampsia meta-signature offers a vital starting point for further studies into the relevance of these genes (in particular CREBBP/EP300) and their concomitant pathways as biomarkers or functional molecules in preeclampsia. This will result in a better understanding of the molecular basis of this disease and opens up the opportunity to develop rational therapies targeting the placental dysfunction causal to preeclampsia. PMID:26171964

  15. Meta-Analysis of Placental Transcriptome Data Identifies a Novel Molecular Pathway Related to Preeclampsia.

    Directory of Open Access Journals (Sweden)

    Miranda van Uitert

    Full Text Available Studies using the placental transcriptome to identify key molecules relevant for preeclampsia are hampered by a relatively small sample size. In addition, they use a variety of bioinformatics and statistical methods, making comparison of findings challenging. To generate a more robust preeclampsia gene expression signature, we performed a meta-analysis on the original data of 11 placenta RNA microarray experiments, representing 139 normotensive and 116 preeclamptic pregnancies. Microarray data were pre-processed and analyzed using standardized bioinformatics and statistical procedures and the effect sizes were combined using an inverse-variance random-effects model. Interactions between genes in the resulting gene expression signature were identified by pathway analysis (Ingenuity Pathway Analysis, Gene Set Enrichment Analysis, Graphite and protein-protein associations (STRING. This approach has resulted in a comprehensive list of differentially expressed genes that led to a 388-gene meta-signature of preeclamptic placenta. Pathway analysis highlights the involvement of the previously identified hypoxia/HIF1A pathway in the establishment of the preeclamptic gene expression profile, while analysis of protein interaction networks indicates CREBBP/EP300 as a novel element central to the preeclamptic placental transcriptome. In addition, there is an apparent high incidence of preeclampsia in women carrying a child with a mutation in CREBBP/EP300 (Rubinstein-Taybi Syndrome. The 388-gene preeclampsia meta-signature offers a vital starting point for further studies into the relevance of these genes (in particular CREBBP/EP300 and their concomitant pathways as biomarkers or functional molecules in preeclampsia. This will result in a better understanding of the molecular basis of this disease and opens up the opportunity to develop rational therapies targeting the placental dysfunction causal to preeclampsia.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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, SESSA spectral modeling indicated that washed Au/Ag-core shell NPs were stabilized with a 0.8 nm l

  17. Scattering loss analysis and structure optimization of hollow-core photonic bandgap fiber

    Science.gov (United States)

    Song, Jingming; Wu, Rong; Sun, Kang; Xu, Xiaoliang

    2016-06-01

    Effects of core structure in 7 cell hollow-core photonic bandgap fibers (HC-PBGFs) on scattering loss are analyzed by means of investigating normalized interface field intensity. Fibers with different core wall thickness, core radius and rounding corner of air hole are simulated. Results show that with thick core wall and expanded core radius, scattering loss could be greatly reduced. The scattering loss of the HC-PBGFs in the wavelength range of 1.5-1.56 μm could be decreased by about 50 % of the present level with optimized core structure design.

  18. Identifying Population Groups with Low Palliative Care Program Enrolment Using Classification and Regression Tree Analysis

    Science.gov (United States)

    Gao, Jun; Lavergne, M. Ruth; McIntyre, Paul

    2013-01-01

    Classification and regression tree (CART) analysis was used to identify subpopulations with lower palliative care program (PCP) enrolment rates. CART analysis uses recursive partitioning to group predictors. The PCP enrolment rate was 72 percent for the 6,892 adults who died of cancer from 2000 and 2005 in two counties in Nova Scotia, Canada. The lowest PCP enrolment rates were for nursing home residents over 82 years (27 percent), a group residing more than 43 kilometres from the PCP (31 percent), and another group living less than two weeks after their cancer diagnosis (37 percent). The highest rate (86 percent) was for the 2,118 persons who received palliative radiation. Findings from multiple logistic regression (MLR) were provided for comparison. CART findings identified low PCP enrolment subpopulations that were defined by interactions among demographic, social, medical, and health system predictors. PMID:21805944

  19. Analysis of regulatory protease sequences identified through bioinformatic data mining of the Schistosoma mansoni genome

    Directory of Open Access Journals (Sweden)

    Minchella Dennis J

    2009-10-01

    Full Text Available Abstract Background New chemotherapeutic agents against Schistosoma mansoni, an etiological agent of human schistosomiasis, are a priority due to the emerging drug resistance and the inability of current drug treatments to prevent reinfection. Proteases have been under scrutiny as targets of immunological or chemotherapeutic anti-Schistosoma agents because of their vital role in many stages of the parasitic life cycle. Function has been established for only a handful of identified S. mansoni proteases, and the vast majority of these are the digestive proteases; very few of the conserved classes of regulatory proteases have been identified from Schistosoma species, despite their vital role in numerous cellular processes. To that end, we identified protease protein coding genes from the S. mansoni genome project and EST library. Results We identified 255 protease sequences from five catalytic classes using predicted proteins of the S. mansoni genome. The vast majority of these show significant similarity to proteins in KEGG and the Conserved Domain Database. Proteases include calpains, caspases, cytosolic and mitochondrial signal peptidases, proteases that interact with ubiquitin and ubiquitin-like molecules, and proteases that perform regulated intramembrane proteolysis. Comparative analysis of classes of important regulatory proteases find conserved active site domains, and where appropriate, signal peptides and transmembrane helices. Phylogenetic analysis provides support for inferring functional divergence among regulatory aspartic, cysteine, and serine proteases. Conclusion Numerous proteases are identified for the first time in S. mansoni. We characterized important regulatory proteases and focus analysis on these proteases to complement the growing knowledge base of digestive proteases. This work provides a foundation for expanding knowledge of proteases in Schistosoma species and examining their diverse function and potential as targets

  20. Thermohydraulic and mechanical analysis of the research reactor Munich II Compact-Core

    International Nuclear Information System (INIS)

    The new research reactor Munich II (Forschungsreaktor Muenchen II, FRM-II), which is under construction at the Technical University of Munich, Germany, contains a compact reactor core consisting of one single fuel element, assembled by two concentric tubes between which 113 involutely bent fuel plates are located rotationally symmetric. In order to perform the hydraulic and mechanical testing of the FRM-II fuel element, two test facilities have been built at the Department for Nuclear and New Energy Systems of the Ruhr University Bochum. The first mocks up the central region of the reactor coolant system of the FRM-II in a 1:1 scale with emphasis on the fuel element and the inflow and discharge section in order to enable the analysis of the FRM-II core. In the course of the testing the vibration behaviour and the flow resistance of the core were investigated. Likewise start-up and shut down tests of the main pump unit were simulated and the flow profile at the outlet of the element as well as the flow division inside the core were determined. Furthermore an endurance test lasting 60 days (equivalent to 12 operating cycles) was performed, too. Tests including blockages of parts of the reactor cooling system cross section at the core entrance sieve proved the efficiency of the cooling capacity. No major resonances occurred during operation and an endurance test neither showed any incidents nor irregularities. In order to investigate the concept of the decay heat removal in the FRM-II a second test facility was built. This facility simulates the thermohydraulic conditions in one cooling channel of the FRM-II by means of an electrically heated test section, which enables different operating conditions of the decay heat removal system as well as enhanced safety investigations. In the FRM-II the decay heat, which is produced after a shutdown, is removed by means of decay heat removal pumps, which maintain a downward flow in the fuel element for at least three hours

  1. Periostin identified as a potential biomarker of prostate cancer by iTRAQ-proteomics analysis of prostate biopsy

    Directory of Open Access Journals (Sweden)

    Tong Shijun

    2011-04-01

    Full Text Available Abstract Background Proteomics may help us better understand the changes of multiple proteins involved in oncogenesis and progression of prostate cancer(PCa and identify more diagnostic and prognostic biomarkers. The aim of this study was to screen biomarkers of PCa by the proteomics analysis using isobaric tags for relative and absolute quantification(iTRAQ. Methods The patients undergoing prostate biopsies were classified into 3 groups according to pathological results: benign prostate hyperplasia (BPH, n = 20, PCa(n = 20 and BPH with local prostatic intraepithelial neoplasm(PIN, n = 10. Then, all the specimens from these patients were analyzed by iTRAQ and two-dimensional liquid chromatography-tandem mass spectrometry (2DLC-MS/MS. The Gene Ontology(GO function and the transcription regulation networks of the differentially expressed were analyzed by MetaCore software. Western blotting and Immunohistochemical staining were used to analyze the interesting proteins. Result A total of 760 proteins were identified from 13787 distinct peptides, including two common proteins that enjoy clinical application: prostate specific antigen (PSA and prostatic acid phosphatase(PAP. Proteins that expressed differentially between PCa and BPH group were further analyzed. Compared with BPH, 20 proteins were significantly differentially up-regulated (>1.5-fold while 26 were significantly down-regulated in PCa( Conclusion Our study indicates that the iTRAQ technology is a new strategy for global proteomics analysis of the tissues of PCa. A significant up-regulation of periostin in PCa compared to BPH may provide clues for not only a promising biomarker for the prognosis of PCa but also a potential target for therapeutical intervention.

  2. A subchannel code for LMR core thermal hydraulic design and analysis with inter-assembly heat transfer

    International Nuclear Information System (INIS)

    The core design for liquid metal reactor (LMR) requires accurate prediction of the thermal hydraulic behaviors in the subassemblies and the core, to ensure that certain economic and safety considerations will be met. Many of these are related to fuel and cladding maximum temperatures and subassembly coolant outlet mixed mean temperatures, for both steady state and transient conditions. A detailed analysis code MATRA-LMR was developed for LMR core thermal hydraulic analysis, based on COBRA-IV-I and MATRA codes, which use a subchannel approach for calculating the enthalpy and flow distribution in nuclear fuel rod bundle elements. MATRA-LMR was used only for a single subassembly subchannel analysis. But it has been developed for the multi-assembly calculations with inter-assembly heat transfer modeling into the code. It will be extended for the multi-assembly whole core calculations for LMR core design and analysis. This paper summarizes the inter-assembly heat transfer development and some validation works. The validation of MATRA-LMR has been done by the benchmark analysis with the experimental data. The major calculation results of the conceptual design of the KALIMER core have been compared with MATRA-LMR, THI3D and SLTHEN codes. KALIMER, which is under design optimization study at KAERI, is a 150 MWe (392 MWt) pool-type sodium cooled prototype reactor. (author)

  3. Physics analysis on the NRU core for an accident scenario of a loop pressure tube crack

    International Nuclear Information System (INIS)

    The Nuclear Research Universal (NRU) reactor loops are high temperature, high pressure test facilities, designed for power reactor fuel development and materials testing within the core of the NRU reactor. The loops allow test material to be subject to neutron flux, temperature and pressure conditions typical of a power reactor. This paper describes the physics analysis on the NRU core for an accident scenario of a loop pressure tube crack with a concurrent liner tube failure. After the crack has occurred, thermal-hydraulic analysis predicts the formation of a steam bubble of 50 cm radius in the D2O moderator/coolant around the loop test section. The steam displaces the D2O moderator and has a negative reactivity effect. This negative reactivity effect is large enough to overcome the positive loop void reactivity such that the reactor is shut down and reactor safety is not compromised. The paper also describes the sensitivity of steam bubble densities on the reactivity effect and presents results for subsequent reductions of fluxes and channel powers around the loop site. (author)

  4. Physics analysis on the NRU core for an accident scenario of a loop pressure tube crack

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C., E-mail: leungt@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2010-07-01

    The Nuclear Research Universal (NRU) reactor loops are high temperature, high pressure test facilities, designed for power reactor fuel development and materials testing within the core of the NRU reactor. The loops allow test material to be subject to neutron flux, temperature and pressure conditions typical of a power reactor. This paper describes the physics analysis on the NRU core for an accident scenario of a loop pressure tube crack with a concurrent liner tube failure. After the crack has occurred, thermal-hydraulic analysis predicts the formation of a steam bubble of 50 cm radius in the D{sub 2}O moderator/coolant around the loop test section. The steam displaces the D{sub 2}O moderator and has a negative reactivity effect. This negative reactivity effect is large enough to overcome the positive loop void reactivity such that the reactor is shut down and reactor safety is not compromised. The paper also describes the sensitivity of steam bubble densities on the reactivity effect and presents results for subsequent reductions of fluxes and channel powers around the loop site. (author)

  5. Criticality qualification of a new Monte Carlo code for reactor core analysis

    International Nuclear Information System (INIS)

    In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.

  6. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Gehin, Jess C [ORNL

    2009-04-01

    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  7. Coupling the core analysis program DeCART to the fuel performance application BISON

    Energy Technology Data Exchange (ETDEWEB)

    Gleicher, F. N.; Spencer, B.; Novascone, S.; Williamson, R.; Martineau, R. C. [Idaho National Laboratory, 2525 N. Fremont Avenue, Idaho Falls, ID 83415 (United States); Rose, M.; Downar, T. J.; Collins, B. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd., Ann Arbor, MI 48105 (United States)

    2013-07-01

    The 3D neutron transport and core analysis program DeCART was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the method of characteristics) to a high fidelity fuel performance program, both of which can simulate 3D problems. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during burnup or a fast transient. BISON implicitly solves coupled thermomechanical equations for the fuel on a sub-millimeter level finite element mesh. A method was developed for mapping the fission rate density and fast neutron flux from DeCART to BISON. Multiple depletion cases were run with one-way data transfer from DeCART to BISON. The one-way data transfer of fission rate density is shown to agree with the fission rate density obtained from an internal Lassman-style model in BISON. One-way data transfer was also demonstrated in a 3D case in which azimuthal asymmetry was induced in the fission rate density profile of a fuel rod modeled in DeCART. Two-way data transfer was established by mapping the temperature distribution from BISON to DeCART. A Picard iterative algorithm was developed for the loose coupling with two-way data transfer. (authors)

  8. Coupling the core analysis program DeCART to the fuel performance application BISON

    International Nuclear Information System (INIS)

    The 3D neutron transport and core analysis program DeCART was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the method of characteristics) to a high fidelity fuel performance program, both of which can simulate 3D problems. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during burnup or a fast transient. BISON implicitly solves coupled thermomechanical equations for the fuel on a sub-millimeter level finite element mesh. A method was developed for mapping the fission rate density and fast neutron flux from DeCART to BISON. Multiple depletion cases were run with one-way data transfer from DeCART to BISON. The one-way data transfer of fission rate density is shown to agree with the fission rate density obtained from an internal Lassman-style model in BISON. One-way data transfer was also demonstrated in a 3D case in which azimuthal asymmetry was induced in the fission rate density profile of a fuel rod modeled in DeCART. Two-way data transfer was established by mapping the temperature distribution from BISON to DeCART. A Picard iterative algorithm was developed for the loose coupling with two-way data transfer. (authors)

  9. Criticality qualification of a new Monte Carlo code for reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Catsaros, N. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Gaveau, B. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Jaekel, M. [Laboratoire de Physique Theorique, Ecole Normale Superieure, 24 rue Lhomond, 75231 Paris (France); Maillard, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); CNRS-IDRIS, Bt 506, BP167, 91403 Orsay (France); CNRS-IN2P3, 3 rue Michel Ange, 75794 Paris (France); Maurel, G. [Faculte de Medecine, Universite Paris VI, 27 rue de Chaligny, 75012 Paris (France); MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Savva, P., E-mail: savvapan@ipta.demokritos.g [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Silva, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Varvayanni, M.; Zisis, Th. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece)

    2009-11-15

    In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.

  10. Emergent team roles in organizational meetings: Identifying communication patterns via cluster analysis.

    OpenAIRE

    Lehmann-Willenbrock, N.K.; Beck, S.J.; Kauffeld, S.

    2016-01-01

    Previous team role taxonomies have largely relied on self-report data, focused on functional roles, and described individual predispositions or personality traits. Instead, this study takes a communicative approach and proposes that team roles are produced, shaped, and sustained in communicative behaviors. To identify team roles communicatively, 59 regular organizational meetings were videotaped and analyzed. Cluster analysis revealed five emergent roles: the solution seeker, the problem anal...

  11. Identifying Gender-Preferred Communication Styles within Online Cancer Communities: A Retrospective, Longitudinal Analysis

    OpenAIRE

    Durant, Kathleen T.; McCray, Alexa T.; Charles Safran

    2012-01-01

    BACKGROUND: The goal of this research is to determine if different gender-preferred social styles can be observed within the user interactions at an online cancer community. To achieve this goal, we identify and measure variables that pertain to each gender-specific social style. METHODS AND FINDINGS: We perform social network and statistical analysis on the communication flow of 8,388 members at six different cancer forums over eight years. Kruskal-Wallis tests were conducted to measure the ...

  12. Network analysis identifies protein clusters of functional importance in juvenile idiopathic arthritis

    OpenAIRE

    Stevens, Adam; Meyer, Stefan; Hanson, Daniel; Clayton, Peter; Donn, Rachelle

    2014-01-01

    Introduction Our objective was to utilise network analysis to identify protein clusters of greatest potential functional relevance in the pathogenesis of oligoarticular and rheumatoid factor negative (RF-ve) polyarticular juvenile idiopathic arthritis (JIA). Methods JIA genetic association data were used to build an interactome network model in BioGRID 3.2.99. The top 10% of this protein:protein JIA Interactome was used to generate a minimal essential network (MEN). Reactome FI Cytoscape 2.83...

  13. Integrative Omics Analysis of Rheumatoid Arthritis Identifies Non-Obvious Therapeutic Targets

    OpenAIRE

    Whitaker, John W.; Boyle, David L.; Bartok, Beatrix; Ball, Scott T.; Gay, Steffen; Wang, Wei; Firestein, Gary S.

    2015-01-01

    Identifying novel therapeutic targets for the treatment of disease is challenging. To this end, we developed a genome-wide approach of candidate gene prioritization. We independently collocated sets of genes that were implicated in rheumatoid arthritis (RA) pathogenicity through three genome-wide assays: (i) genome-wide association studies (GWAS), (ii) differentially expression in RA fibroblast-like synoviocytes (FLS), and (iii) differentially methylation in RA FLS. Integrated analysis of the...

  14. Integrative omics analysis of rheumatoid arthritis identifies non-obvious therapeutic targets

    OpenAIRE

    Whitaker, John W.; Boyle, David L.; Bartok, Beatrix; Ball, Scott T.; Gay, Steffen; Wang, Wei; Firestein, Gary S.

    2015-01-01

    Identifying novel therapeutic targets for the treatment of disease is challenging. To this end, we developed a genome-wide approach of candidate gene prioritization. We independently collocated sets of genes that were implicated in rheumatoid arthritis (RA) pathogenicity through three genome-wide assays: (i) genome-wide association studies (GWAS), (ii) differentially expression in RA fibroblast-like synoviocytes (FLS), and (iii) differentially methylation in RA FLS. Integrated analysis of the...

  15. Robust Microarray Meta-Analysis Identifies Differentially Expressed Genes for Clinical Prediction

    OpenAIRE

    Phan, John H.; Andrew N. Young; Wang, May D.

    2012-01-01

    Combining multiple microarray datasets increases sample size and leads to improved reproducibility in identification of informative genes and subsequent clinical prediction. Although microarrays have increased the rate of genomic data collection, sample size is still a major issue when identifying informative genetic biomarkers. Because of this, feature selection methods often suffer from false discoveries, resulting in poorly performing predictive models. We develop a simple meta-analysis-ba...

  16. Identifying patterns in treatment response profiles in acute bipolar mania: a cluster analysis approach

    OpenAIRE

    Houston John P; Lipkovich Ilya A; Ahl Jonna

    2008-01-01

    Abstract Background Patients with acute mania respond differentially to treatment and, in many cases, fail to obtain or sustain symptom remission. The objective of this exploratory analysis was to characterize response in bipolar disorder by identifying groups of patients with similar manic symptom response profiles. Methods Patients (n = 222) were selected from a randomized, double-blind study of treatment with olanzapine or divalproex in bipolar I disorder, manic or mixed episode, with or w...

  17. Automated Source Code Analysis to Identify and Remove Software Security Vulnerabilities: Case Studies on Java Programs

    OpenAIRE

    2013-01-01

    The high-level contribution of this paper is to illustrate the development of generic solution strategies to remove software security vulnerabilities that could be identified using automated tools for source code analysis on software programs (developed in Java). We use the Source Code Analyzer and Audit Workbench automated tools, developed by HP Fortify Inc., for our testing purposes. We present case studies involving a file writer program embedded with features for password validation, and ...

  18. System reliability analysis using dominant failure modes identified by selective searching technique

    International Nuclear Information System (INIS)

    The failure of a redundant structural system is often described by innumerable system failure modes such as combinations or sequences of local failures. An efficient approach is proposed to identify dominant failure modes in the space of random variables, and then perform system reliability analysis to compute the system failure probability. To identify dominant failure modes in the decreasing order of their contributions to the system failure probability, a new simulation-based selective searching technique is developed using a genetic algorithm. The system failure probability is computed by a multi-scale matrix-based system reliability (MSR) method. Lower-scale MSR analyses evaluate the probabilities of the identified failure modes and their statistical dependence. A higher-scale MSR analysis evaluates the system failure probability based on the results of the lower-scale analyses. Three illustrative examples demonstrate the efficiency and accuracy of the approach through comparison with existing methods and Monte Carlo simulations. The results show that the proposed method skillfully identifies the dominant failure modes, including those neglected by existing approaches. The multi-scale MSR method accurately evaluates the system failure probability with statistical dependence fully considered. The decoupling between the failure mode identification and the system reliability evaluation allows for effective applications to larger structural systems

  19. Nonlinear Analysis of Core Competence for Real Estate Enterprise Using Evolutionary Fuzzy Neural Inference Model

    OpenAIRE

    Huawang Shi; Wei Hou

    2011-01-01

    The real estate development trade is a pillar industry of our national economy. The real estate companies must focus on core competence in order to succeed in the fierce competition. This paper presents a probabilistic model for core competence estimation of real estate firms. According to the nonlinear feature of real estate core competence, this paper establishes the fuzzy and neural network model to core competence assessment of real estate. We analyze the features of core competence, thus...

  20. Hot spot analysis applied to identify ecosystem services potential in Lithuania

    Science.gov (United States)

    Pereira, Paulo; Depellegrin, Daniel; Misiune, Ieva

    2016-04-01

    Hot spot analysis are very useful to identify areas with similar characteristics. This is important for a sustainable use of the territory, since we can identify areas that need to be protected, or restored. This is a great advantage in terms of land use planning and management, since we can allocate resources, reduce the economical costs and do a better intervention in the landscape. Ecosystem services (ES) are different according land use. Since landscape is very heterogeneous, it is of major importance understand their spatial pattern and where are located the areas that provide better ES and the others that provide less services. The objective of this work is to use hot-spot analysis to identify areas with the most valuable ES in Lithuania. CORINE land-cover (CLC) of 2006 was used as the main spatial information. This classification uses a grid of 100 m resolution and extracted a total of 31 land use types. ES ranking was carried out based on expert knowledge. They were asked to evaluate the ES potential of each different CLC from 0 (no potential) to 5 (very high potential). Hot spot analysis were evaluated using the Getis-ord test, which identifies cluster analysis available in ArcGIS toolbox. This tool identifies areas with significantly high low values and significant high values at a p level of 0.05. In this work we used hot spot analysis to assess the distribution of providing, regulating cultural and total (sum of the previous 3) ES. The Z value calculated from Getis-ord was used to statistical analysis to access the clusters of providing, regulating cultural and total ES. ES with high Z value show that they have a high number of cluster areas with high potential of ES. The results showed that the Z-score was significantly different among services (Kruskal Wallis ANOVA =834. 607, pcultural (0.080±1.979) and regulating (0.076±1.961). These results suggested that providing services are more clustered than the remaining. Ecosystem Services Z score were

  1. Decay heat analysis of MNSR reactor core using ORIGEN-2 code

    International Nuclear Information System (INIS)

    Highlights: • Analysis of the decay heat parameters of the MNSR reactor was performed using the ORIGEN-2 code. • A new one-group cross-section data base of the ORlGEN-2 computer code for the MNSR was developed. • It is recommended to adopt the results of the ORIGIEN-2 code for future reactor safety analysis of MNSR. - Abstract: The knowledge of the decay heat of nuclear fuel is necessary for performing the reactor safety analysis, determining the heating load in fuel pools, shielding requirements on fuel discharge and transport routes when irradiated reactor fuel is transferred from the reactor, via some intermediate storage location, to the final disposal or the chemical reprocessing plant. In this study, analysis of the decay heat parameters of the Miniature Neutron Source Reactor (MNSR) including radioactivity, decay heat and the isotopic mass variation with time since reactor shutdown for the potential Low Enriched Uranium (LEU) (UO2-Zircaloy and U3Si-Al) and the standard Highly Enriched Uranium (HEU) (HEU-Al4) cores has been performed using the ORIGEN-2 code. For this purpose, a new one-group cross-section data base of the ORlGEN-2 computer code for the MNSR with LEU and HEU fuels has been developed using the MCNP-4C code. The variation of fission products, actinides and daughters and activation products with post shutdown time for the standard core and the potential LEU cores have been considered in the analysis of the decay heat power resources. It was found that, all the three types of MNSR fuels show close agreement in the total decay heat, which is mainly due to the fission products. This behavior continued for about 1.0E05 days. After this time, the fission products decay heat became comparable with the corresponding actinides decay heat in which the standard HEU UAl4-Al fuel showed the smallest decay heat values while the potential LEU-UO2 fuel had the highest decay heat followed by the LEU-U3Si fuel. The time variation of the total radioactivity

  2. Identifying Innovative Interventions to Promote Healthy Eating Using Consumption-Oriented Food Supply Chain Analysis.

    Science.gov (United States)

    Hawkes, Corinna

    2009-07-01

    The mapping and analysis of supply chains is a technique increasingly used to address problems in the food system. Yet such supply chain management has not yet been applied as a means of encouraging healthier diets. Moreover, most policies recommended to promote healthy eating focus on the consumer end of the chain. This article proposes a consumption-oriented food supply chain analysis to identify the changes needed in the food supply chain to create a healthier food environment, measured in terms of food availability, prices, and marketing. Along with established forms of supply chain analysis, the method is informed by a historical overview of how food supply chains have changed over time. The method posits that the actors and actions in the chain are affected by organizational, financial, technological, and policy incentives and disincentives, which can in turn be levered for change. It presents a preliminary example of the supply of Coca-Cola beverages into school vending machines and identifies further potential applications. These include fruit and vegetable supply chains, local food chains, supply chains for health-promoting versions of food products, and identifying financial incentives in supply chains for healthier eating. PMID:23144674

  3. Combination of meta-analysis and graph clustering to identify prognostic markers of ESCC

    Directory of Open Access Journals (Sweden)

    Hongyun Gao

    2012-01-01

    Full Text Available Esophageal squamous cell carcinoma (ESCC is one of the most malignant gastrointestinal cancers and occurs at a high frequency rate in China and other Asian countries. Recently, several molecular markers were identified for predicting ESCC. Notwithstanding, additional prognostic markers, with a clear understanding of their underlying roles, are still required. Through bioinformatics, a graph-clustering method by DPClus was used to detect co-expressed modules. The aim was to identify a set of discriminating genes that could be used for predicting ESCC through graph-clustering and GO-term analysis. The results showed that CXCL12, CYP2C9, TGM3, MAL, S100A9, EMP-1 and SPRR3 were highly associated with ESCC development. In our study, all their predicted roles were in line with previous reports, whereby the assumption that a combination of meta-analysis, graph-clustering and GO-term analysis is effective for both identifying differentially expressed genes, and reflecting on their functions in ESCC.

  4. Failure analysis of bolted joints in foam-core sandwich composites

    DEFF Research Database (Denmark)

    Zabihpoor, M.; Moslemian, Ramin; Afshin, M.;

    2008-01-01

    This study represents an effort to predict the bearing strength, failure modes, and failure load of bolted joints in foam-core sandwich composites. The studied joints have been used in a light full composite airplane. By using solid laminates, a new design for the joint zone is developed. These...... using 3D FEM in ANSYS commercial code. Tsai-Wu failure criterion is used in the failure analysis. The results indicate that the most important parameter in the proposed joint zone design is the foam -solid laminate interface angle which plays an important role on the value of failure criterion (damage...... analysis, the increase of solid laminate size or interface angle could result in considerable higher buckling strength....

  5. A multi-cycle BWR [boiling water reactor] core reload design analysis system (MCAS)

    International Nuclear Information System (INIS)

    This paper describes the design, construction, and application of a software system (MCAS) for performing boiling water reactor reload core design analysis. MCAS provides for the execution of studies which analyze alternative reload strategies over a range of cycles. Studies are performed by preparing and executing sequential SIMULATE-E Haling depletions and storing the results on a data base for subsequent reporting and analysis. Application of MCAS has shown that the ability of efficiently and accurately predict the effects of next cycle design decisions on future cycles is a valuable capability. This capability results in the proper selection of BWR [boiling water reactor] reload fuel bundle enrichment and batch size as necessary for reload fuel supply planning and early identification and resolution of design problems which would prove expensive if discovered at a later time

  6. CERN antiproton target: Hydrocode analysis of its core material dynamic response under proton beam impact

    Science.gov (United States)

    Martin, Claudio Torregrosa; Perillo-Marcone, Antonio; Calviani, Marco; Muñoz-Cobo, José-Luis

    2016-07-01

    Antiprotons are produced at CERN by colliding a 26 GeV /c proton beam with a fixed target made of a 3 mm diameter, 55 mm length iridium core. The inherent characteristics of antiproton production involve extremely high energy depositions inside the target when impacted by each primary proton beam, making it one of the most dynamically demanding among high energy solid targets in the world, with a rise temperature above 2000 °C after each pulse impact and successive dynamic pressure waves of the order of GPa's. An optimized redesign of the current target is foreseen for the next 20 years of operation. As a first step in the design procedure, this numerical study delves into the fundamental phenomena present in the target material core under proton pulse impact and subsequent pressure wave propagation by the use of hydrocodes. Three major phenomena have been identified, (i) the dominance of a high frequency radial wave which produces destructive compressive-to-tensile pressure response (ii) The existence of end-of-pulse tensile waves and its relevance on the overall response (iii) A reduction of 44% in tensile pressure could be obtained by the use of a high density tantalum cladding.

  7. The CERN antiproton target: hydrocode analysis of its core material dynamic response under proton beam impact

    CERN Document Server

    Martin, Claudio Torregrosa; Calviani, Marco; Muñoz-Cobo, José-Luis

    2016-01-01

    Antiprotons are produced at CERN by colliding a 26 GeV/c proton beam with a fixed target made of a 3 mm diameter, 55 mm length iridium core. The inherent characteristics of antiproton production involve extremely high energy depositions inside the target when impacted by each primary proton beam, making it one of the most dynamically demanding among high energy solid targets in the world, with a rise temperature above 2000 {\\deg}C after each pulse impact and successive dynamic pressure waves of the order of GPa's. An optimized redesign of the current target is foreseen for the next 20 years of operation. As a first step in the design procedure, this numerical study delves into the fundamental phenomena present in the target material core under proton pulse impact and subsequent pressure wave propagation by the use of hydrocodes. Three major phenomena have been identified, (i) the dominance of a high frequency radial wave which produces destructive compressive-to-tensile pressure response (ii) The existence of...

  8. Three-dimensional vortex analysis and aeroacoustic source characterization of jet core breakdown

    Science.gov (United States)

    Violato, Daniele; Scarano, Fulvio

    2013-01-01

    The three-dimensional behavior of jet core breakdown is investigated with experiments conducted on a free water jet at Re = 5000 by time-resolved tomographic particle image velocimetry (TR-TOMO PIV). The investigated domain encompasses the range between 0 and 10 jet diameters. The characteristic pulsatile motion of vortex ring shedding and pairing culminates with the growth of four primary in-plane and out-of-plane azimuthal waves and leads to the formation of streamwise vortices. Vortex ring humps are tilted and ejected along the axial direction as they are subjected to higher axial velocities. By the end of the potential core, this process causes the breakdown of the vortex ring regime and the onset of streamwise filaments oriented at 30°-45° to the jet axis and "C" shaped peripheral structures. The latter re-organize further downstream in filaments oriented along the azimuthal direction at the jet periphery. Instead, in the vicinity of the jet axis the filaments do not exhibit any preferential direction resembling the isotropic turbulent regime. Following Powell's aeroacoustic analogy, the instantaneous spatial distribution of the acoustic source term is mapped by the second time derivative of the Lamb vector, revealing the highest activity during vortex ring breakdown. A three-dimensional modal analysis of velocity, vorticity, Lamb vector, and Lamb vector second time derivative fields is conducted by proper orthogonal decomposition (POD) within the first 10 modes. The decomposed velocity fluctuations describe a helical organization in the region of the jet core-breakdown and, further downstream, jet axis flapping and precession motions. By the end of the potential core, vorticity modes show that vortex rings are dominated by travelling waves of radial and axial vorticity with a characteristic 40°-45° inclination to the jet axis. The Lamb vector and the Lamb vector second time derivative modes exhibit similar patterns for the azimuthal component, whereas the

  9. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    CERN Document Server

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

    1982-01-01

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  10. Analysis of nascent RNA identifies a unified architecture of initiation regions at mammalian promoters and enhancers.

    Science.gov (United States)

    Core, Leighton J; Martins, André L; Danko, Charles G; Waters, Colin T; Siepel, Adam; Lis, John T

    2014-12-01

    Despite the conventional distinction between them, promoters and enhancers share many features in mammals, including divergent transcription and similar modes of transcription factor binding. Here we examine the architecture of transcription initiation through comprehensive mapping of transcription start sites (TSSs) in human lymphoblastoid B cell (GM12878) and chronic myelogenous leukemic (K562) ENCODE Tier 1 cell lines. Using a nuclear run-on protocol called GRO-cap, which captures TSSs for both stable and unstable transcripts, we conduct detailed comparisons of thousands of promoters and enhancers in human cells. These analyses identify a common architecture of initiation, including tightly spaced (110 bp apart) divergent initiation, similar frequencies of core promoter sequence elements, highly positioned flanking nucleosomes and two modes of transcription factor binding. Post-initiation transcript stability provides a more fundamental distinction between promoters and enhancers than patterns of histone modification and association of transcription factors or co-activators. These results support a unified model of transcription initiation at promoters and enhancers. PMID:25383968

  11. The Analysis of Surrounding Structure Effect on the Core Degradation Progress with COMPASS Code

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Jun Ho; Son, Dong Gun; Kim, Jong Tae; Park, Rae Jun; Kim, Dong Ha [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In line with the importance of severe accident analysis after Fukushima accident, the development of integrated severe accident code has been launched by the collaboration of three institutes in Korea. KAERI is responsible to develop modules related to the in-vessel phenomena, while other institutes are to the containment and severe accident mitigation facility, respectively. In the first phase, the individual severe accident module has been developed and the construction of integrated analysis code is planned to perform in the second phase. The basic strategy is to extend the design basis analysis codes of SPACE and CAP, which are being validated in Korea for the severe accident analysis. In the first phase, KAERI has targeted to develop the framework of severe accident code, COMPASS (COre Meltdown Progression Accident Simulation Software), covering the severe accident progression in a vessel from a core heat-up to a vessel failure as a stand-alone fashion. In order to analyze the effect of surrounding structure, the melt progression has been compared between the central zone and the most outer zone under the condition of constant radial power peaking factor. Figure 2 and 3 shows the fuel element temperature and the clad mass at the central zone, respectively. Due to the axial power peaking factor, the axial node No.3 has the highest temperature, while the top and bottom nodes have the lowest temperature. When the clad temperature reaches to the Zr melting temperature (2129.15K), the Zr starts to melt. The axial node No.2 reaches to the fuel melting temperature about 5000 sec and the molten fuel relocates to the node No.1, which results to the blockage of flow area in node No.1. The blocked flow area becomes to open about 6100 sec due to the molten ZrO{sub 2} mass relocation to core support plate. Figure 4 and 5 shows the fuel element temperature and the clad mass at the most outer zone, respectively. It is shown that the fuel temperature increase more slowly

  12. Protein functional links in Trypanosoma brucei, identified by gene fusion analysis

    Directory of Open Access Journals (Sweden)

    Trimpalis Philip

    2011-07-01

    Full Text Available Abstract Background Domain or gene fusion analysis is a bioinformatics method for detecting gene fusions in one organism by comparing its genome to that of other organisms. The occurrence of gene fusions suggests that the two original genes that participated in the fusion are functionally linked, i.e. their gene products interact either as part of a multi-subunit protein complex, or in a metabolic pathway. Gene fusion analysis has been used to identify protein functional links in prokaryotes as well as in eukaryotic model organisms, such as yeast and Drosophila. Results In this study we have extended this approach to include a number of recently sequenced protists, four of which are pathogenic, to identify fusion linked proteins in Trypanosoma brucei, the causative agent of African sleeping sickness. We have also examined the evolution of the gene fusion events identified, to determine whether they can be attributed to fusion or fission, by looking at the conservation of the fused genes and of the individual component genes across the major eukaryotic and prokaryotic lineages. We find relatively limited occurrence of gene fusions/fissions within the protist lineages examined. Our results point to two trypanosome-specific gene fissions, which have recently been experimentally confirmed, one fusion involving proteins involved in the same metabolic pathway, as well as two novel putative functional links between fusion-linked protein pairs. Conclusions This is the first study of protein functional links in T. brucei identified by gene fusion analysis. We have used strict thresholds and only discuss results which are highly likely to be genuine and which either have already been or can be experimentally verified. We discuss the possible impact of the identification of these novel putative protein-protein interactions, to the development of new trypanosome therapeutic drugs.

  13. Gene expression signature analysis identifies vorinostat as a candidate therapy for gastric cancer.

    Directory of Open Access Journals (Sweden)

    Sofie Claerhout

    Full Text Available BACKGROUND: Gastric cancer continues to be one of the deadliest cancers in the world and therefore identification of new drugs targeting this type of cancer is thus of significant importance. The purpose of this study was to identify and validate a therapeutic agent which might improve the outcomes for gastric cancer patients in the future. METHODOLOGY/PRINCIPAL FINDINGS: Using microarray technology, we generated a gene expression profile of human gastric cancer-specific genes from human gastric cancer tissue samples. We used this profile in the Broad Institute's Connectivity Map analysis to identify candidate therapeutic compounds for gastric cancer. We found the histone deacetylase inhibitor vorinostat as the lead compound and thus a potential therapeutic drug for gastric cancer. Vorinostat induced both apoptosis and autophagy in gastric cancer cell lines. Pharmacological and genetic inhibition of autophagy however, increased the therapeutic efficacy of vorinostat, indicating that a combination of vorinostat with autophagy inhibitors may therapeutically be more beneficial. Moreover, gene expression analysis of gastric cancer identified a collection of genes (ITGB5, TYMS, MYB, APOC1, CBX5, PLA2G2A, and KIF20A whose expression was elevated in gastric tumor tissue and downregulated more than 2-fold by vorinostat treatment in gastric cancer cell lines. In contrast, SCGB2A1, TCN1, CFD, APLP1, and NQO1 manifested a reversed pattern. CONCLUSIONS/SIGNIFICANCE: We showed that analysis of gene expression signature may represent an emerging approach to discover therapeutic agents for gastric cancer, such as vorinostat. The observation of altered gene expression after vorinostat treatment may provide the clue to identify the molecular mechanism of vorinostat and those patients likely to benefit from vorinostat treatment.

  14. Measurement and analysis of fractures in vertical, slant, and horizontal core, with examples from the Mesaverde formation

    Energy Technology Data Exchange (ETDEWEB)

    Lorenz, J.C. (Sandia National Labs., Albuquerque, NM (United States)); Hill, R.E. (CER Corp., Las Vegas, NV (United States))

    1991-01-01

    Optimum analysis of natural fracture characteristics and distributions in reservoirs requires conscientious supervision of coring operations, on-site core processing, careful layout and marketing of the core, and detailed measurement of fracture characteristics. Natural fractures provide information on the in situ permeability system, and coring-induced fractures provide data on the in situ stresses. Fracture data derived from vertical core should include fracture height, type and location of fracture terminations with respect to lithologic heterogeneity, fracture planatary and roughness, and distribution with depth. Fractures in core from either a vertical or a deviated well will yield information on dip, dip azimuth, strike, mineralization, and the orientation of fractures relative to the in situ stresses. Only measurements of fractures in core from a deviated/horizontal well will provide estimates of fracture spacing and porosity. These data can be graphed and cross-plotted to yield semi-quantitative fracture characteristics for reservoir models. Data on the orientations of fractures relative to each other in unoriented core can be nearly as useful as the absolute orientations of fractures. A deviated pilot hole is recommended for fracture assessment prior to a drilling horizontal production well because it significantly enhances the chances of fracture intersection, and therefore of fracture characterization. 35 refs., 20 figs., 2 tabs.

  15. Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids

    Science.gov (United States)

    Bernnat, W.; Mattes, M.; Guilliard, N.; Lapins, J.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2014-06-01

    Power reactors are composed of assemblies with fuel pin lattices or other repeated structures with several grid levels, which can be modeled in detail by Monte Carlo neutronics codes such as MCNP6 using corresponding lattice options, even for large cores. Except for fresh cores at beginning of life, there is a varying material distribution due to burnup in the different fuel pins. Additionally, for power states the fuel and moderator temperatures and moderator densities vary according to the power distribution and cooling conditions. Therefore, a coupling of the neutronics code with a thermal hydraulics code is necessary. Depending on the level of detail of the analysis, a very large number of cells with different materials and temperatures must be regarded. The assignment of different material properties to all elements of a multilevel grid is very elaborate and may exceed program limits if the standard input procedure is used. Therefore, an internal assignment is used which overrides uniform input parameters. The temperature dependency of continuous energy cross sections, probability tables for the unresolved resonance region and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. The method is applied with MCNP6 and proven for several full core reactor models. For the coupling of MCNP6 with thermal hydraulics appropriate interfaces were developed for the GRS system code ATHLET for liquid coolant and the IKE thermal hydraulics code ATTICA-3D for gaseous coolant. Examples will be shown for different applications for PWRs with square and hexagonal lattices, fast reactors (SFR) with hexagonal lattices and HTRs with pebble bed and prismatic lattices.

  16. PIXE analysis as a tool for dating of ice cores from the Greenland ice sheet

    International Nuclear Information System (INIS)

    Sections from the 2037 m long Dye 3 ice core drilled in 1979-1981 in the ice sheet of Southern Greenland were analysed with PIXE. The seven selected sections were from depths between 1778 and 1813 m, which corresponds to a time interval between about 8 500 and 10 000 years B.C. at the end of the last Ice Age. During this time period, fast climatic changes of several degrees centrigrade per century are known to have taken place. The exact time scales of these changes need yet to be verified by renewed measurements using nonconventional stratigraphic dating techniques such as PIXE. The problem is highly relevant for the prediction of climatic changes in our present age. A new sample preparation technique was developed which enables the determination of annual thicknesses of the parts of the ice core representing 10 000-40 000 years before present, where the thickness of the annual ice layers are believed to be less than 2.5 cm. More commonly used techniques of dating, such as measurements of oxygen and hydrogen isotopes δ18O and δD, nitrate, acidity or conductivity all have difficulties in resolving annual cycles in thicknesses of less than about 2 cm. The new technique involves sublimation of 18 cm long ice sections, after which the material contained in the ice is deposited on the thin backing. In this way, the material to be analysed is preconcentrated through the removal of the H2O, while still retaining the spatial distribution pattern of the various water soluble and insoluble components along the ice core. The resulting spatial resolution of the sublimation technique is estimated to be ±1 mm. A PIXE analysis was performed in contiguous millimeter steps across the sublimated ice sections. Estimations of annual ice layer thicknesses were based on the patterns of seasonal variation along the ice sections for several major and minor elements quantified with PIXE. (orig./TW)

  17. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  18. Analysis of full core steam flooding experiments for the Phase II GCFR critical assembly

    International Nuclear Information System (INIS)

    The initial program of bench mark critical experiments conducted on behalf of the design and safety evaluations for the 300 MW(e) gas cooled fast breeder reactor demonstration plant included extensive measurements of the reactivity effects of accidental steam ingress. Insertions of polyethylene (CH2) foam into all of the void channels in the 1250-liter (l) core, the radial blankets, and the axial blankets of the Phase II GCFR critical assembly gave simulated floodings of up to 2.25% steam in the coolant. The report presents results of General Atomic Company (GA) analyses of the Phase II steam entry experiments, giving comparisons of calculated and measured flooding worths under various conditions, including changes in core geometry and introduction of control rod poisoning. Also studied were the effects of steam flooding on control material worth and other physics parameters. Calculated worths of hydrogenous materials were found to be significantly sensitive to variations in analytical models and methods. Good agreement with experiments was obtained by a 28-group analysis when a rigorous regeneration of cross sections, cell-heterogeneity factors, and directional diffusion coefficients was provided at each specific flooding density to account for the moderated spectra. Steam worths in a rodded core can be similarly well predicted provided that rod shielding effects are re-evaluated in the steam environment. Extrapolations based on these experiments clearly suggest that should a steam leak occur, it would not be a major safety concern, even in a small GCFR demonstration plant. Details of the analytical procedures and models utilized are presented

  19. Genome-wide association analysis identifies novel loci for chronotype in 100,420 individuals from the UK Biobank.

    Science.gov (United States)

    Lane, Jacqueline M; Vlasac, Irma; Anderson, Simon G; Kyle, Simon D; Dixon, William G; Bechtold, David A; Gill, Shubhroz; Little, Max A; Luik, Annemarie; Loudon, Andrew; Emsley, Richard; Scheer, Frank A J L; Lawlor, Deborah A; Redline, Susan; Ray, David W; Rutter, Martin K; Saxena, Richa

    2016-01-01

    Our sleep timing preference, or chronotype, is a manifestation of our internal biological clock. Variation in chronotype has been linked to sleep disorders, cognitive and physical performance, and chronic disease. Here we perform a genome-wide association study of self-reported chronotype within the UK Biobank cohort (n=100,420). We identify 12 new genetic loci that implicate known components of the circadian clock machinery and point to previously unstudied genetic variants and candidate genes that might modulate core circadian rhythms or light-sensing pathways. Pathway analyses highlight central nervous and ocular systems and fear-response-related processes. Genetic correlation analysis suggests chronotype shares underlying genetic pathways with schizophrenia, educational attainment and possibly BMI. Further, Mendelian randomization suggests that evening chronotype relates to higher educational attainment. These results not only expand our knowledge of the circadian system in humans but also expose the influence of circadian characteristics over human health and life-history variables such as educational attainment. PMID:26955885

  20. Genome wide association analysis of a founder population identified TAF3 as a gene for MCHC in humans.

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    Giorgio Pistis

    Full Text Available The red blood cell related traits are highly heritable but their genetics are poorly defined. Only 5-10% of the total observed variance is explained by the genetic loci found to date, suggesting that additional loci should be searched using approaches alternative to large meta analysis. GWAS (Genome Wide Association Study for red blood cell traits in a founder population cohort from Northern Italy identified a new locus for mean corpuscular hemoglobin concentration (MCHC in the TAF3 gene. The association was replicated in two cohorts (rs1887582, P = 4.25E-09. TAF3 encodes a transcription cofactor that participates in core promoter recognition complex, and is involved in zebrafish and mouse erythropoiesis. We show here that TAF3 is required for transcription of the SPTA1 gene, encoding alpha spectrin, one of the proteins that link the plasma membrane to the actin cytoskeleton. Mutations in SPTA1 are responsible for hereditary spherocytosis, a monogenic disorder of MCHC, as well as for the normal MCHC level. Based on our results, we propose that TAF3 is required for normal erythropoiesis in human and that it might have a role in controlling the ratio between hemoglobin (Hb and cell volume and in the dynamics of RBC maturation in healthy individuals. Finally, TAF3 represents a potential candidate or a modifier gene for disorders of red cell membrane.