NPP-VIIRS DNB-based reallocating subpopulations to mercury in Urumqi city cluster, central Asia

Science.gov (United States)

Zhou, X.; Feng, X. B.; Dai, W.; Li, P.; Ju, C. Y.; Bao, Z. D.; Han, Y. L.

2017-02-01

Accurate and update assignment of population-related environmental matters onto fine grid cells in oasis cities of arid areas remains challenging. We present the approach based on Suomi National Polar-orbiting Partnership (S-NPP) -Visible Infrared Imaging Radiometer Suite (VIIRS) Day/Night Band (DNB) to reallocate population onto a regular finer surface. The number of potential population to the mercury were reallocated onto 0.1x0.1 km reference grid in Urumqi city cluster of China’s Xinjiang, central Asia. The result of Monte Carlo modelling indicated that the range of 0.5 to 2.4 million people was reliable. The study highlights that the NPP-VIIRS DNB-based multi-layered, dasymetric, spatial method enhances our abilities to remotely estimate the distribution and size of target population at the street-level scale and has the potential to transform control strategies for epidemiology, public policy and other socioeconomic fields.

76 FR 24883 - DNB Exports LLC, and AFI Elektromekanikanik Ve Elektronik San. Tic. Ltd. Sti. v. Barsan Global...

Science.gov (United States)

2011-05-03

... FEDERAL MARITIME COMMISSION [Docket No. 11-07] DNB Exports LLC, and AFI Elektromekanikanik Ve Elektronik San. Tic. Ltd. Sti. v. Barsan Global Lojistiks Ve Gumruk Musavirligi A.S., Barsan International... AFI Elektromekanikanik Ve Elektronik San. Tic. Ltd. Sti. (``AFI''), hereinafter ``Complainants...

Thermal Decomposition of 1,5-Dinitrobiuret (DNB): Direct Dynamics Trajectory Simulations and Statistical Modeling

Science.gov (United States)

2011-05-03

Vibrational frequencies and zero-point energies ( ZPE ) were scaled by a factor of 0.955 and 0.981,20 respectively. The corrected ZPE were added to the...PE) during the trajectory. The oscillations in the PE reflect the vibration of the DNB molecule and the products, including ZPE . The time scale of...Energetics of complexes, TSs, and products are derived from B3LYP/6-31++G** calculations, including ZPE . For TSs, vibrational modes corresponding

Design development of bellows for the DNB beam source

International Nuclear Information System (INIS)

Singh, Dhananjay Kumar; Venkata Nagaraju, M.; Joshi, Jaydeep; Patel, Hitesh; Yadav, Ashish; Pillai, Suraj; Singh, Mahendrajit; Bandyopadhyay, Mainak; Chakraborty, A.K.; Sharma, Dheeraj

2017-01-01

Establishing a procedure and mechanism for alignment of Ion beams in Neutral Beam (NB) sources for ITER like systems are complex due to large traversal distances (∼21 m) and restricted use of flexible elements into the system. For the beam source of DNB, movement requirements for beam alignment are the combination of tilting (±9mrad), rotation (±9mrad) and translation (±25mm). The present work describes the design development of a system composed of three single ply ‘Gimbal’ type bellow system, placed in series, in L-shaped hydraulic lines (size DN50, DN20 and DN15). The paper shall detail out the generation of initial requirements, transformation of movements at bellow locations, selection of bellows/combination of bellows, minimizing the induced movements by optimization of bellows location, estimation of movements through CEASAR II and the design compliance with respect to EJMA code

Impact on DNB predictions of mixing models implemented into the three-dimensional thermal-hydraulic code Thyc; Impact de modeles de melange implantes dans le code de thermohydraulique Thyc sur les predictions de flux critique

Energy Technology Data Exchange (ETDEWEB)

Banner, D

1993-10-01

The objective of this paper is to point out how departure from nucleate boiling (DNB) predictions can be improved by the THYC software. The EPRI/Columbia University E161 data base has been used for this study. In a first step, three thermal-hydraulic mixing models have been implemented into the code in order to obtain more accurate calculations of local void fractions at the DNB location. The three investigated models (A, B and C) are presented by growing complexity. Model A assumes a constant turbulent viscosity throughout the flow. In model B, a k-L turbulence transport equation has been implemented to model generation and decay of turbulence in the DNB test section. Model C is obtained by representing oriented transverse flows due to mixing vanes in addition to the k-L equation. A parametric study carried out with the three mixing models exhibits the most significant parameters. The occurrence of departure from nucleate boiling is then predicted by using a DNB correlation. Similar results are obtained as long as the DNB correlation is kept unchanged. In a second step, an attempt to substitute correlations by another statistical approach (pseudo-cubic thin-plate type Spline method) has been done. It is then shown that standard deviations of P/M (predicted to measured) ratios can be greatly improved by advanced statistics. (author). 7 figs., 2 tabs., 9 refs.

An Estimate of the Pixel-Level Connection between Visible Infrared Imaging Radiometer Suite Day/Night Band (VIIRS DNB Nighttime Lights and Land Features across China

Directory of Open Access Journals (Sweden)

Ting Ma

2018-05-01

Full Text Available Satellite-derived nighttime light images are increasingly used for various studies in relation to demographic, socioeconomic and urbanization dynamics because of the salient relationships between anthropogenic lighting signals at night and statistical variables at multiple scales. Owing to a higher spatial resolution and fewer over-glow and saturation effects, the new generation of nighttime light data derived from the Visible Infrared Imaging Radiometer Suite (VIIRS day/night band (DNB, which is located on board the Suomi National Polar-Orbiting Partnership (Suomi-NPP satellite, is expected to facilitate the performance of nocturnal luminosity-based investigations of human activity in a spatially explicit manner. In spite of the importance of the spatial connection between the VIIRS DNB nighttime light radiance (NTL and the land surface type at a fine scale, the crucial role of NTL-based investigations of human settlements is not well understood. In this study, we investigated the pixel-level relationship between the VIIRS DNB-derived NTL, a Landsat-derived land-use/land-cover dataset, and the map of point of interest (POI density over China, especially with respect to the identification of artificial surfaces in urban land. Our estimates suggest that notable differences in the NTL between urban (man-made surfaces and other types of land surfaces likely allow us to spatially identify most of the urban pixels with relatively high radiance values in VIIRS DNB images. Our results also suggest that current nighttime light data have a limited capability for detecting rural residential areas and explaining pixel-level variations in the POI density at a large scale. Moreover, the impact of non-man-made surfaces on the partitioned results appears inevitable because of the spatial heterogeneity of human settlements and the nature of remotely sensed nighttime light data. Using receiver operating characteristic (ROC curve-based analysis, we obtained

Modelling of a DNB mechanism by dry-out of a nucleation site

International Nuclear Information System (INIS)

Bricard, P.

1995-10-01

This study deals with the modelling of a nucleation site dry-out DNB mechanism which unifies those of Kirby et al. (1967) and Fiori and Bergles (1970). A first model based on a simplified heat balance in the wall at the location of the dry spot is developed and a set of closure relations is proposed. The model is then quantitatively and qualitatively compared to CHF data. In order to support the likelihood of the mechanism, we develop a more elaborated model which couples the unsteady thermal behavior of the wall and the thermal-hydraulics of the fluid described by the different phases of the nucleation cycle. The conditions which enable the boiling crisis to be reached are given

DNB heat flux on inner side of a vertical pipe in forced flow of liquid hydrogen and liquid nitrogen

Science.gov (United States)

Shirai, Yasuyuki; Tatsumoto, Hideki; Shiotsu, Masahiro; Hata, Koichi; Kobayashi, Hiroaki; Naruo, Yoshihiro; Inatani, Yoshifumi

2018-06-01

Heat transfer from inner side of a heated vertical pipe to liquid hydrogen flowing upward was measured at the pressures of 0.4, 0.7 and 1.1 MPa for wide ranges of flow rate and liquid temperature. Nine test heaters with different inner diameters of 3, 4, 6 and 9 mm and the lengths of 50, 100, 150, 200, 250 and 300 mm were used. The DNB (departure from nucleate boiling) heat fluxes in forced flow of liquid hydrogen were measured for various subcoolings and flow velocities at pressures of 0.4, 0.7 and 1.1 MPa. Effect of L/d (ratio of heater length to diameter) was clarified for the range of L / d ⩽ 50 . A new correlation of DNB heat flux was presented based on a simple model and the experimental data. Similar experiments were performed for liquid nitrogen at pressures of 0.5 MPa and 1.0 MPa by using the same experimental system and some of the test heaters. It was confirmed that the new correlation can describe not only the hydrogen data, but also the data of liquid nitrogen.

Preliminary design of bellows for the DNB beam source by EJMA and FE linear analysis

International Nuclear Information System (INIS)

Trapasiya, Shobhit; Muvvala, Venkata Nagaraju; Rambilas, P.; Gangadharan, Roopesh; Rotti, Chandramouli; Chakraborty, Arun Kumar; Sharma, Dheeraj Kumar

2015-01-01

In piping system, U-shaped Bellows are widely used among flexible elements. In general, bellows are typically design for Fatigue behavior according to the EJMA standard based on empirically generated fatigue curves. The present work proposes a methodology in the design of bellows by design by analyses and validates its design by EJMA standard. A linear FE approach is chosen to in line with the EJMA standard. The proposed methodology is benchmarked with the available literatures. The same practice is implemented in the preliminary design of a U-shaped bellows in the water line circuits of DNB beam source. DNB Beam Source is a negative ion source-based neutral beam generator for ITER operates at 100KV. The beam divergence (intrinsic) and magnetic fields from ITER torus causes deflection of beams. This calls for beam optic alignment, which are assured by BS Movement mechanism system. To accomplish the above movement requirements, bellows, which is a stringent of its kind (± 22 mm axial, ± 45 mm lateral within 400mm available space with single ply), is designed between the beam source and possible rigid interface-cooling lines coming from HVB. The paper describes right from conceptual stage to preliminary design. Optimization tools are adopted in the selecting bellow dimensions using MATLAB. At the end a coordinated approach between FE based assessment (in ANSYS) and widely applied code, EJMA is implemented for the validation of design and found FE approach is a very conservative than later in the present case. (author)

Measurement of neutral beam power and beam profile distribution on DNB

International Nuclear Information System (INIS)

Liu Zhimin; Liu Sheng; Song Shihua; Han Xiaopu; Li Jun; Hu Chundong; Hu Liqun; Xie Jun

2005-01-01

The injection power of a diagnostic neutral beam (DNB) can be obtained with the thermocouple probe measurement system on the Hefei superconducting Tokamak-7 (HT-7). With the 49 kv, 6 A, 100 ms pulse charge of an acceleration electrode, a thermocouple probe measurement system with 13 thermocouples crossly distributed on a coppery heat target was used to measure the temperature rise of the target, and the maximum measured temperature rise was 14 degree C. And the neutral beam power of 160 kW and beam profile distribution was obtained by calculation. The total neutral beam power of 130 kW was also obtained by integral calculation with the temperature rise on the heat section board. The difference between the two means was analyzed. The experiment results shows that the method of heat section board with thermocouple probe is one of the effective ways to measure the beam power and beam profile distribution. (authors)

Design optimization of the 100 kV HV bushing for ITER-DNB

Energy Technology Data Exchange (ETDEWEB)

Shah, Sejal, E-mail: sshah@iter-india.org [ITER-India, Institute for Plasma Research, Gandhinagar 382025, Gujarat (India); Rajesh, S. [ITER-India, Institute for Plasma Research, Gandhinagar 382025, Gujarat (India); Microelectronics and Materials Physics Laboratories, P.O. Box 4500, FIN-90014 University of Oulu (Finland); Srusti, B. [DesignTech Systems Ltd, Banjara Hills, Hyderabad, Andhra Pradesh 500034 (India); Bandyopadhyay, M.; Rotti, C.; Singh, M.J.; Roopesh, G.; Chakraborty, A.K. [ITER-India, Institute for Plasma Research, Gandhinagar 382025, Gujarat (India); Schunke, B.; Hemsworth, R.; Chareyre, J. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France)

2011-10-15

The 100 kV bushing of the Diagnostic Neutral Beam (DNB) injector is a cylindrical feedthrough which forms the interface between the gas insulated transmission line and the torus primary vacuum and provides all necessary services to the beam source. All conventions for safety, voltage holding requirement, vacuum compatibility and the choice of materials have been addressed in the design. Finite Element Analyses (FEAs) for the electrostatic and the structural configuration is carried out to validate the design of the High Voltage Bushing (HVB). Several iterations and optimizations of the stress shields are carried out to meet electrostatic criteria, especially at the triple point (the ceramic, metal and vacuum joint), which is critical for good voltage holding. Structural analyses is carried out to assess the stress distribution in the fiber reinforced plastic insulator, the alumina insulator and the integrated HVB for different load cases like operational orientation (horizontal), normal operation and accidental case. Design is further validated for seismic conditions for Seismic Loading-2 (SL-2).

Design optimization of the 100 kV HV bushing for ITER-DNB

International Nuclear Information System (INIS)

Shah, Sejal; Rajesh, S.; Srusti, B.; Bandyopadhyay, M.; Rotti, C.; Singh, M.J.; Roopesh, G.; Chakraborty, A.K.; Schunke, B.; Hemsworth, R.; Chareyre, J.

2011-01-01

The 100 kV bushing of the Diagnostic Neutral Beam (DNB) injector is a cylindrical feedthrough which forms the interface between the gas insulated transmission line and the torus primary vacuum and provides all necessary services to the beam source. All conventions for safety, voltage holding requirement, vacuum compatibility and the choice of materials have been addressed in the design. Finite Element Analyses (FEAs) for the electrostatic and the structural configuration is carried out to validate the design of the High Voltage Bushing (HVB). Several iterations and optimizations of the stress shields are carried out to meet electrostatic criteria, especially at the triple point (the ceramic, metal and vacuum joint), which is critical for good voltage holding. Structural analyses is carried out to assess the stress distribution in the fiber reinforced plastic insulator, the alumina insulator and the integrated HVB for different load cases like operational orientation (horizontal), normal operation and accidental case. Design is further validated for seismic conditions for Seismic Loading-2 (SL-2).

Vibrational signatures in the THz spectrum of 1,3-DNB: A first-principles and experimental study

Science.gov (United States)

Ahmed, Towfiq; Azad, Abul K.; Chellappa, Raja; Higginbotham-Duque, Amanda; Dattelbaum, Dana M.; Zhu, Jian-Xin; Moore, David; Graf, Matthias J.

2016-05-01

Understanding the fundamental processes of light-matter interaction is important for detection of explosives and other energetic materials, which are active in the infrared and terahertz (THz) region. We report a comprehensive study on electronic and vibrational lattice properties of structurally similar 1,3-dinitrobenzene (1,3-DNB) crystals through first-principles electronic structure calculations and THz spectroscopy measurements on polycrystalline samples. Starting from reported x-ray crystal structures, we use density-functional theory (DFT) with periodic boundary conditions to optimize the structures and perform linear response calculations of the vibrational properties at zero phonon momentum. The theoretically identified normal modes agree qualitatively with those obtained experimentally in a frequency range up to 2.5 THz and quantitatively at much higher frequencies. The latter frequencies are set by intra-molecular forces. Our results suggest that van der Waals dispersion forces need to be included to improve the agreement between theory and experiment in the THz region, which is dominated by intermolecular modes and sensitive to details in the DFT calculation. An improved comparison is needed to assess and distinguish between intra- and intermolecular vibrational modes characteristic of energetic materials.

Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

International Nuclear Information System (INIS)

Mesquita, Amir Z.; Rezende, Hugo C.; Santos, Andre Augusto C.; Silva, Vitor Vasconcelos A.; Campolina, Daniel de Almeida M.

2013-01-01

In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Amir Z.; Rezende, Hugo C.; Santos, Andre Augusto C.; Silva, Vitor Vasconcelos A.; Campolina, Daniel de Almeida M., E-mail: amir@cdtn.br, E-mail: hcr@cdtn.br, E-mail: aacs@cdtn.br, E-mail: vitors@cdtn.br, E-mail: campolina@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/UFMG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel Artur P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2013-07-01

In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

Descriptive normative beliefs and the self-regulation in alcohol use among Slovak university students.

Science.gov (United States)

Brutovská, Monika; Orosova, Olga; Kalina, Ondrej; Šebeňa, René

2015-12-01

This study aims (i) to understand how descriptive normative beliefs (DNB) about typical students' alcohol use and self-regulation (SRG) are related to alcohol use (AU) by exploring the indirect effect of SRG on AU through DNB and (ii) to explore gender differences and the differences between universities in DNB, SRG and AU. The cross-sectional data were collected online from 817 Slovak university students from four universities (75.22% females; Mage = 19.61; SD = 1.42), who filled in the AUDIT-C items, items measuring the DNB about typical students' AU and SRG. T-tests, one-way Anova and structural equation modelling were used for data analysis. Gender differences in AU and DNB were found with males having higher levels of both AU and DNB. The tested model of AU fits the data well. A significant association was found between DNB and (i) AU (positive) and (ii) SRG (negative). The analysis confirmed the existence of an indirect effect of SRG on AU through DNB. The study contributes to research concerning AU by the way in which DNB and SRG are linked to AU among Slovak university students. The research findings can also be used in developing prevention and intervention programs. © The Author 2014. Published by Oxford University Press on behalf of Faculty of Public Health. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

Safety criterion for burnout of the plate-type fuel in pressurized conditions

International Nuclear Information System (INIS)

Komori, Y.; Kaminaga, M.; Sakurai, F.; Ando, H.; Sudo, Y.; Saito, M.; Futamura, Y.

1992-01-01

The reduced enrichment program for JMTR is now underway and the core conversion to LEU (Low Enrichment Uranium) is scheduled to be made in 1993. Consistent with the safety guide which have been recently developed for research and test reactors in Japan, the safety analysis for the JMTR LEU conversion was conducted. In the safety analysis, DNB (Departure from Nucleate Boiling) heat flux correlation for the JMTR downflow condition was reconsidered because recent studies on burnout show that DNB heat fluxes with thin rectangular channels under low flow rate and low pressure conditions are much lower than predicted values by conventional DNB correlations. Available DNB data, however, are very limited for the JMTR operation pressure range, so that DNB experiments were conducted simulating the JMTR fuel subchannel. Based mainly on the present experimental data, the DNB correlations scheme composed of three correlations was selected for the JMTR safety analysis. Errors of the correlations scheme with experimental data were evaluated in order to determine the allowable limit of the minimum DNB ratio for preventing fuel failure. (author)

Mixed Inhibition of Adenosine Deaminase Activity by 1,3-Dinitrobenzene: A Model for Understanding Cell-Selective Neurotoxicity in Chemically-Induced Energy Deprivation Syndromes in Brain

Science.gov (United States)

Wang, Yipei; Liu, Xin; Schneider, Brandon; Zverina, Elaina A.; Russ, Kristen; Wijeyesakere, Sanjeeva J.; Fierke, Carol A.; Richardson, Rudy J.; Philbert, Martin A.

2012-01-01

Astrocytes are acutely sensitive to 1,3-dinitrobenzene (1,3-DNB) while adjacent neurons are relatively unaffected, consistent with other chemically-induced energy deprivation syndromes. Previous studies have investigated the role of astrocytes in protecting neurons from hypoxia and chemical injury via adenosine release. Adenosine is considered neuroprotective, but it is rapidly removed by extracellular deaminases such as adenosine deaminase (ADA). The present study tested the hypothesis that ADA is inhibited by 1,3-DNB as a substrate mimic, thereby preventing adenosine catabolism. ADA was inhibited by 1,3-DNB with an IC50 of 284μM, Hill slope, n = 4.8 ± 0.4. Native gel electrophoresis showed that 1,3-DNB did not denature ADA. Furthermore, adding Triton X-100 (0.01–0.05%, wt/vol), Nonidet P-40 (0.0015–0.0036%, wt/vol), or bovine serum albumin (0.05 mg/ml or changing [ADA] (0.2 and 2nM) did not substantially alter the 1,3-DNB IC50 value. Likewise, dynamic light scattering showed no particle formation over a (1,3-DNB) range of 149–1043μM. Kinetics revealed mixed inhibition with 1,3-DNB binding to ADA (KI = 520 ± 100μM, n = 1 ± 0.6) and the ADA-adenosine complex (KIS = 262 ± 7μM, n = 6 ± 0.6, indicating positive cooperativity). In accord with the kinetics, docking predicted binding of 1,3-DNB to the active site and three peripheral sites. In addition, exposure of DI TNC-1 astrocytes to 10–500μM 1,3-DNB produced concentration-dependent increases in extracellular adenosine at 24 h. Overall, the results demonstrate that 1,3-DNB is a mixed inhibitor of ADA and may thus lead to increases in extracellular adenosine. The finding may provide insights to guide future work on chemically-induced energy deprivation. PMID:22106038

Full scale mock-up tests for rod bundle thermal-hydraulics in Japan

International Nuclear Information System (INIS)

Sugawara, S.

1995-01-01

This poster describes tests aimed at development and validation of principal design methodology of rod bundle thermal-hydraulics correlations. The works are based on domestic data base using the full-scale mock-up test facilities. The scope of the tests comprises DNB heat flux, transient DNB heat flux, post DNB heat transfer, pressure drop and void distribution. The works have been performed under collaboration among electric facilities, NPP vendors, universities, governmental corporations. 1 tab., 14 figs

The S-NPP VIIRS Day-Night Band On-Orbit Calibration/Characterization and Current State of SDR Products

Directory of Open Access Journals (Sweden)

Shihyan Lee

2014-12-01

Full Text Available The launch of VIIRS on-board the Suomi-National Polar-orbiting Partnership (S-NPP on 28 October 2011, marked the beginning of the next chapter on nighttime lights observation started by the Defense Meteorological Satellite Program’s (DMSP OLS sensor more than two decades ago. The VIIRS observes the nighttime lights on Earth through its day-night band (DNB, a panchromatic channel covering the wavelengths from 500 nm to 900 nm. Compared to its predecessors, the VIIRS DNB has a much improved spatial/temporal resolution, radiometric sensitivity and, more importantly, continuous calibration using on-board calibrators (OBCs. In this paper, we describe the current state of the NASA calibration and characterization methodology used in supporting mission data quality assurance and producing consistent mission-wide sensor data records (SDRs through NASA’s Land Product Evaluation and Analysis Tool Element (Land PEATE. The NASA calibration method utilizes the OBCs to determine gains, offset drift and sign-to-noise ratio (SNR over the entire mission. In gain determination, the time-dependent relative spectral response (RSR is used to correct the optical throughput change over time. A deep space view acquired during an S-NPP pitch maneuver is used to compute the airglow free dark offset for DNB’s high gain stage. The DNB stray light is estimated each month from new-moon dark Earth surface observations to remove the excessive stray light over the day-night terminators. As the VIIRS DNB on-orbit calibration is the first of its kind, the evolution of the calibration methodology is evident when the S-NPP VIIRS’s official calibrations are compared with our latest mission-wide reprocessing. In the future, the DNB calibration methodology is likely to continue evolving, and the mission-wide reprocessing is a key to providing consistently calibrated DNB SDRs for the user community. In the meantime, the NASA Land PEATE provides an alternative source to obtain

Interactions among K+-Ca2+ exchange, sorption of m-dinitrobenzene, and smectite quasicrystal dynamics.

Science.gov (United States)

Chatterjee, Ritushree; Laird, David A; Thompson, Michael L

2008-12-15

The fate of organic contaminants in soils and sediments is influenced by sorption of the compounds to surfaces of soil materials. We investigated the interaction among sorption of an organic compound, cation exchange reactions, and both the size and swelling of smectite quasicrystals. Two reference smectites that vary in location and amount of layer charge, SPV (a Wyoming bentonite) and SAz-1 were initially Ca- and K-saturated and then equilibrated with mixed 0.01 M KCl and 0.005 M CaCl2 salt solutions both with and without the presence of 200 mg L(-1) m-dinitrobenzene (m-DNB). In general, sorption of m-DNB increased with the amount of K+ in the system for both clays, and the SPV sorbed more m-DNB than the SAz-1. Sorption of m-DNB increased the preference of Ca-SPV for K+ relative to Ca2+ but had little effect on K+-Ca2+ selectivity for K-SPV. Selectivity for K+ relative to Ca2+ was slightly higher for both K-SAz-1 and Ca-SAz-1 in the presence of m-DNB than in its absence. Distinct hysteresis loops were observed for the K+-Ca2+ cation exchange reactions for both clays, and the legacy of having been initially Ca- or K-saturated influenced sorption of m-DNB by SPV but had little effect for SAz-1. Suspension X-ray diffraction was used to measure changes in d-spacing and the relative thickness of smectite quasicrystals during the cation exchange and m-DNB sorption reactions. The results suggest that interactions among cation exchange and organic sorption reactions are controlled byan inherently hysteretic complex feedback process that is regulated by changes in the size and extent of swelling of smectite quasicrystals.

Improvements to Lunar BRDF-Corrected Nighttime Satellite Imagery: Uses and Applications

Science.gov (United States)

Cole, Tony A.; Molthan, Andrew L.; Schultz, Lori A.; Roman, Miguel O.; Wanik, David W.

2016-01-01

Observations made by the VIIRS day/night band (DNB) provide daily, nighttime measurements to monitor Earth surface processes.However, these observations are impacted by variations in reflected solar radiation on the moon's surface. As the moon transitions from new to full phase, increasing radiance is reflected to the Earth's surface and contributes additional reflected moonlight from clouds and land surface, in addition to emissions from other light sources observed by the DNB. The introduction of a bi-directional reflectance distribution function (BRDF) algorithm serves to remove these lunar variations and normalize observed radiances. Provided by the Terrestrial Information Systems Laboratory at Goddard Space Flight Center, a 1 km gridded lunar BRDF-corrected DNB product and VIIRS cloud mask can be used for a multitude of nighttime applications without influence from the moon. Such applications include the detection of power outages following severe weather events using pre-and post-event DNB imagery, as well as the identification of boat features to curtail illegal fishing practices. This presentation will provide context on the importance of the lunar BRDF correction algorithm and explore the aforementioned uses of this improved DNB product for applied science applications.

Improvements to Lunar BRDF-Corrected Nighttime Satellite Imagery: Uses and Applications

Science.gov (United States)

Cole, T.; Molthan, A.; Schultz, L. A.; Roman, M. O.; Wanik, D. W.

2016-12-01

Observations made by the VIIRS day/night band (DNB) provide daily, nighttime measurements to monitor Earth surface processes. However, these observations are impacted by variations in reflected solar radiation on the moon's surface. As the moon transitions from new to full phase, increasing radiance is reflected to the Earth's surface and contributes additional reflected moonlight from clouds and land surface, in addition to emissions from other light sources observed by the DNB. The introduction of a bi-directional reflectance distribution function (BRDF) algorithm serves to remove these lunar variations and normalize observed radiances. Provided by the Terrestrial Information Systems Laboratory at Goddard Space Flight Center, a 1 km gridded lunar BRDF-corrected DNB product and VIIRS cloud mask can be used for a multitude of nighttime applications without influence from the moon. Such applications include the detection of power outages following severe weather events using pre- and post-event DNB imagery, as well as the identification of boat features to curtail illegal fishing practices. This presentation will provide context on the importance of the lunar BRDF correction algorithm and explore the aforementioned uses of this improved DNB product for applied science applications.

Measuring equipment for limit-value control of pressurized water reactors. Pt. 1

International Nuclear Information System (INIS)

Aleite, W.; Mertens, U.

1975-08-01

A test installation of a Siemens 101 process computer in the Stade nuclear power station is used to monitor the onset of excessive local power densities or excessively low DNB ratios. A linearised W3 formula complemented by signals derived from the core power distribution detectors, constitutes the major part of the DNB system. Operating experience yielded by the installations has been used for the formulation of proposals for a DNB criterium including detailled error analysis. Some aspects are given of closed loop operation of a computer-based monitoring system. (orig.) [de

Evaluation of the dependence of heat transfer coefficient on the particle diameter of a metal porous medium in a heat removal system using liquid nitrogen

International Nuclear Information System (INIS)

Sasaki, Shunsuke; Ito, Satoshi; Hashizume, Hidetoshi

2015-01-01

Cryogenic cooling system using a bronze-particle-sintered porous medium has been studied for a re mountable high-temperature superconducting magnet. This study evaluates boiling curve of subcooled liquid nitrogen as flowing in a bronze porous medium as a function of the particle diameter of the medium. We obtained Departure from Nuclear Boiling (Dnb) point from the boiling curve and discussed growth of nitrogen vapor bubble inferred from measured pressure drop. The pressure drop decreased significantly at wall superheat before reaching the DNB point whereas that slightly decreased after reaching the DNB point compared to the smallest wall superheat. This result could consider DNB rises with an increase in the particle diameter because larger particle makes vapor to move easily from the heated pore region. The influence of the particle diameter on the heat transfer performance is larger than that of coolant's degree of subcooling. (author)

Suomi-NPP VIIRS Day-Night Band On-Orbit Calibration and Performance

Science.gov (United States)

Chen, Hongda; Xiong, Xiaoxiong; Sun, Chengbo; Chen, Xuexia; Chiang, Kwofu

2017-01-01

The Suomi national polar-orbiting partnership Visible Infrared Imaging Radiometer Suite (VIIRS) instrument has successfully operated since its launch in October 2011. The VIIRS day-night band (DNB) is a panchromatic channel covering wavelengths from 0.5 to 0.9 microns that is capable of observing Earth scenes during both daytime and nighttime at a spatial resolution of 750 m. To cover the large dynamic range, the DNB operates at low-, middle-, and high-gain stages, and it uses an on-board solar diffuser (SD) for its low-gain stage calibration. The SD observations also provide a means to compute the gain ratios of low-to-middle and middle-to-high gain stages. This paper describes the DNB on-orbit calibration methodology used by the VIIRS characterization support team in supporting the NASA Earth science community with consistent VIIRS sensor data records made available by the land science investigator-led processing systems. It provides an assessment and update of the DNB on-orbit performance, including the SD degradation in the DNB spectral range, detector gain and gain ratio trending, and stray-light contamination and its correction. Also presented in this paper are performance validations based on Earth scenes and lunar observations, and comparisons to the calibration methodology used by the operational interface data processing segment.

VIIRS day-night band gain and offset determination and performance

Science.gov (United States)

Geis, J.; Florio, C.; Moyer, D.; Rausch, K.; De Luccia, F. J.

2012-09-01

On October 28th, 2011, the Visible-Infrared Imaging Radiometer Suite (VIIRS) was launched on-board the Suomi National Polar-orbiting Partnership (NPP) spacecraft. The instrument has 22 spectral bands: 14 reflective solar bands (RSB), 7 thermal emissive bands (TEB), and a Day Night Band (DNB). The DNB is a panchromatic, solar reflective band that provides visible through near infrared (IR) imagery of earth scenes with radiances spanning 7 orders of magnitude. In order to function over this large dynamic range, the DNB employs a focal plane array (FPA) consisting of three gain stages: the low gain stage (LGS), the medium gain stage (MGS), and the high gain stage (HGS). The final product generated from a DNB raw data record (RDR) is a radiance sensor data record (SDR). Generation of the SDR requires accurate knowledge of the dark offsets and gain coefficients for each DNB stage. These are measured on-orbit and stored in lookup tables (LUT) that are used during ground processing. This paper will discuss the details of the offset and gain measurement, data analysis methodologies, the operational LUT update process, and results to date including a first look at trending of these parameters over the early life of the instrument.

Development of a DNBR evaluation method for the CEA ejection accident in SMART core

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Yoo, Y. J.; In, W. K.; Chang, M. H. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

A methodology applicable to the analysis of the CEA ejection accident in SMART is developed for the evaluation of the fraction of fuel failure caused by DNB. The transient behavior of the core thermal-hydraulic conditions is calculated by the subchannel analysis code MATRA. The minimum DNBR during the accident is calculated by KRB-1 CHF correlation considering the 1/8 symmetry of hot assembly. The variation of hot assembly power during the accident is simulated by the LTC(Limiting transient Curve) which is determined from the analysis of power distribution data resulting from the three-dimensional core dynamics calculations. The initial condition of the accident is determined by considering LOC(Limiting Conditions for Operation) of SMART core. Two different methodologies for the evaluation of DNB failure rate are established; a deterministic method based on the DNB envelope, and a probabilistic method based on the DNB probability of each fuel rod. The methodology developed in this study is applied to the analysis of CEA ejection accident in the preliminary design core of SMART. As the result, the fractions of DNB fuel failure by the deterministic method and the probabilistic method are calculated as 38.7% and 7.8%, respectively. 16 refs., 16 figs., 5 tabs. (Author)

Comparison between the Suomi-NPP Day-Night Band and DMSP-OLS for Correlating Socio-Economic Variables at the Provincial Level in China

Directory of Open Access Journals (Sweden)

Xin Jing

2015-12-01

Full Text Available Nighttime light imagery offers a unique view of the Earth’s surface. In the past, the nighttime light data collected by the DMSP-OLS sensors have been used as an efficient means to correlate regional and global socio-economic activities. With the launch of the Suomi National Polar-orbiting Partnership (Suomi-NPP satellite in 2011, the day-night band (DNB of the Visible Infrared Imaging Radiometer Suite (VIIRS onboard represents a major advancement in nighttime imaging capabilities, because it surpasses its predecessor DMSP-OLS in radiometric accuracy, spatial resolution and geometric quality. In this paper, four variables (total night light, light area, average night light and log average night light are extracted from nighttime radiance data observed by the VIIRS-DNB composite in 2013 and nighttime digital number (DN data from the DMSP-OLS stable dataset in 2012, respectively, and correlated with 12 socio-economic parameters at the provincial level in mainland China during the corresponding period. Background noise of DNB composite data is removed using either a masking method or an optimal threshold method. In general, the correlation of these socio-economic data with the total night light and light area of VIIRS-DNB composite data is better than with the DMSP-OLS stable data. The correlations between total night light of denoised DNB composite data and built-up area, gross regional product (GRP and power consumption are higher than 0.9 and so are the correlations between the light area of denoised DNB composite data and city and town population, built-up area, GRP, power consumption and waste water discharge. However, the correlations of socio-economic data with the average night light and log average night light of VIIRS-DNB composite data are not as good as with the DMSP-OLS stable data. To quantitatively analyze the reasons for the correlation difference, a cubic regression method is developed to correct the saturation effect of the DMSP

Using Ground Targets to Validate S-NPP VIIRS Day-Night Band Calibration

Science.gov (United States)

Chen, Xuexia; Wu, Aisheng; Xiong, Xiaoxiong; Lei, Ning; Wang, Zhipeng; Chiang, Kwofu

2016-01-01

In this study, the observations from S-NPP VIIRS Day-Night band (DNB) and Moderate resolution bands (M bands) of Libya 4 and Dome C over the first four years of the mission are used to assess the DNB low gain calibration stability. The Sensor Data Records produced by NASA Land Product Evaluation and Algorithm Testing Element (PEATE) are acquired from nearly nadir overpasses for Libya 4 desert and Dome C snow surfaces. A kernel-driven bidirectional reflectance distribution function (BRDF) correction model is used for both Libya 4 and Dome C sites to correct the surface BRDF influence. At both sites, the simulated top-of-atmosphere (TOA) DNB reflectances based on SCIAMACHY spectral data are compared with Land PEATE TOA reflectances based on modulated Relative Spectral Response (RSR). In the Libya 4 site, the results indicate a decrease of 1.03% in Land PEATE TOA reflectance and a decrease of 1.01% in SCIAMACHY derived TOA reflectance over the period from April 2012 to January 2016. In the Dome C site, the decreases are 0.29% and 0.14%, respectively. The consistency between SCIAMACHY and Land PEATE data trends is good. The small difference between SCIAMACHY and Land PEATE derived TOA reflectances could be caused by changes in the surface targets, atmosphere status, and on-orbit calibration. The reflectances and radiances of Land PEATE DNB are also compared with matching M bands and the integral M bands based on M4, M5, and M7. The fitting trends of the DNB to integral M bands ratios indicate a 0.75% decrease at the Libya 4 site and a 1.89% decrease at the Dome C site. Part of the difference is due to an insufficient number of sampled bands available within the DNB wavelength range. The above results indicate that the Land PEATE VIIRS DNB product is accurate and stable. The methods used in this study can be used on other satellite instruments to provide quantitative assessments for calibration stability.

Using Ground Targets to Validate S-NPP VIIRS Day-Night Band Calibration

Directory of Open Access Journals (Sweden)

Xuexia Chen

2016-11-01

Full Text Available In this study, the observations from S-NPP VIIRS Day-Night band (DNB and Moderate resolution bands (M bands of Libya 4 and Dome C over the first four years of the mission are used to assess the DNB low gain calibration stability. The Sensor Data Records produced by NASA Land Product Evaluation and Algorithm Testing Element (PEATE are acquired from nearly nadir overpasses for Libya 4 desert and Dome C snow surfaces. A kernel-driven bidirectional reflectance distribution function (BRDF correction model is used for both Libya 4 and Dome C sites to correct the surface BRDF influence. At both sites, the simulated top-of-atmosphere (TOA DNB reflectances based on SCIAMACHY spectral data are compared with Land PEATE TOA reflectances based on modulated Relative Spectral Response (RSR. In the Libya 4 site, the results indicate a decrease of 1.03% in Land PEATE TOA reflectance and a decrease of 1.01% in SCIAMACHY derived TOA reflectance over the period from April 2012 to January 2016. In the Dome C site, the decreases are 0.29% and 0.14%, respectively. The consistency between SCIAMACHY and Land PEATE data trends is good. The small difference between SCIAMACHY and Land PEATE derived TOA reflectances could be caused by changes in the surface targets, atmosphere status, and on-orbit calibration. The reflectances and radiances of Land PEATE DNB are also compared with matching M bands and the integral M bands based on M4, M5, and M7. The fitting trends of the DNB to integral M bands ratios indicate a 0.75% decrease at the Libya 4 site and a 1.89% decrease at the Dome C site. Part of the difference is due to an insufficient number of sampled bands available within the DNB wavelength range. The above results indicate that the Land PEATE VIIRS DNB product is accurate and stable. The methods used in this study can be used on other satellite instruments to provide quantitative assessments for calibration stability.

Development of MHI PWR fuel assembly with high thermal performance

International Nuclear Information System (INIS)

Yasushi Makino; Masaya Hoshi; Masaji Mori; Hidetoshi Kido; Kazuo Ikeda

2005-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a PWR fuel assembly to meet the needs of Japanese fuel market with mainly improving its reliability such as a mechanical strength, a seismic strength and endurance. For burn-up extension of the fuel to 55 GWd/t, MHI has introduced a Zircaloy spacer grid with better neutron economics with retaining the reliability in an operating core. However, for a future power up-rating and a longer cycle operation, a higher thermal performance is required for PWR fuel assembly. To meet the needs of fuel market, MHI has developed an advanced type of Zircaloy spacer grid with a greater DNB performance while retaining the reliability of a fuel and a relatively low pressure drop. For the greater DNB performance, MHI optimized geometrical shape of mixing vane to promote a fluid mixing performance. In this report, higher DNB performance provided by the advanced Zircaloy spacer grid is presented. The results of 3D simulation for the flow behavior in 5 x 5 partial assembly, a mixing test and a water DNB test were compared between the current and the advanced spacer grids. Consequently, it was confirmed that a crossover vane enhanced a fluid mixing and the advanced spacer grid could significantly improve DNB performance compared with the current design of spacer grids. (authors)

Seeing the Night in a New Light—VIIRS Day/Night Band Capabilities and Prospects for a Joint Suomi/JPSS-1 Observing System

Science.gov (United States)

Solbrig, J. E.; Miller, S. D.; Straka, W. C.; Seaman, C.; Combs, C.; Heidinger, A.; Walther, A.

2017-12-01

The Day/Night Band (DNB), a special sensor on board the Visible/Infrared Imaging Radiometer Suite (VIIRS) devoted to low-light visible imaging, has representated a kind of `disruptive technology' in terms of how we observe the nocturnal environment. Since its debut on the Suomi National Polar-orbiting Partnership (NPP), launched in Fall 2011, the DNB has solidified its claim to fame as the most novel addition to the National Oceanic and Atmospheric Administration's future polar-oribitng program, represented by the Joint Polar Satellite System (JPSS). The first member of which (JPSS-1) is scheduled to launch in Fall of 2017, joining Suomi in its 1330 local time ascending node orbit. JPSS-1 will be displaced by ½ orbit ahead of Suomi, providing roughly 50 min between overpasses. Importantly, JPSS-1 will provide a second DNB observation, enabling the first time-resolved measurements of low-light visible at low and mid-latitudes from this new sensor technology. The DNB provides unprecedented capability to leverage light emissions from natural and artificial nocturnal sources, ranging from moonlight and city lights, ships, fires, lightning flashes, and even atmospheric nightglow. The calibrated DNB observations enable use of moonlight in similar way to daytime visible, allowing for quantitative description of cloud and aerosol optical properties. This presentation updates the community on DNB-related research initiatives. Statistics based on a multi-year collection of data at Salar de Uyuni, Bolivia and White Sands, New Mexico lend confidence to the performance of a lunar irradiance model used to enable nighttime optical property retrievals. Selected examples of notable events, including the devastating Portugal wildfires, emergence of the massive rift in the Larsen C ice shelf, and examples from the growing compilation of atmospheric gravity waves in nightglow, will also be highlighted.

Quo vadis Inhaltserschließung der Deutschen Nationalbibliothek? Herausforderungen und Perspektiven

Directory of Open Access Journals (Sweden)

Ulrike Junger

2015-03-01

Full Text Available Seit der Ausweitung des Pflichtexemplarrechts auf sog. unkörperliche Medienwerke im Jahr 2006 steht die Deutsche Nationalbibliothek (DNB vor der Herausforderung, über neue Wege zur Erzeugung bibliografischer Metadaten nachzudenken. Die steigende Menge an Publikationen, die es zu bearbeiten gilt, aber auch deren zunehmende Diversität, machen dies erforderlich. Für die Inhaltserschließung der DNB bedeutet das, dass andere Verfahren neben die intellektuelle Erschließung treten müssen. Die Entwicklung und Einführung automatischer Verfahren für Klassifikation und Beschlagwortung gehören ebenso dazu wie die Nutzung von Fremddaten oder die Gewinnung inhaltserschließender Daten über Konkordanzen. Dies hat Auswirkungen auf Arbeitsabläufe, Erschließungsinstrumente und -standards sowie Datenformate. Die DNB strebt ein Konzept für die Inhaltserschließung an, das zum Ziel hat, so viele Publikationen als möglich angemessen zu erschließen, um thematische Recherchen erfolgreich zu unterstützen. Since the extension of the legal deposit mandate to so-called non-physical publications in 2006, the Deutsche Nationalbibliothek (DNB has to deal with the challenge of finding new ways to create bibliographic metadata. This is due to the growing amount of publications and their increasing diversity. Consequently, in the area of subject cataloguing in DNB, intellectual cataloguing must be complemented by additional methods. These include the development and implementation of automated procedures for classification and indexing as well as the use of third party data or the production of subject data via mappings. All this has effects on the workflows, tools and standards for subject cataloguing and data formats. DNB aims at a concept for subject cataloguing which will make it possible to catalogue as many publications as possible in an appropriate way in order to successfully support topical searches

Aurora Research: Earth/Space Data Fusion Powered by GIS and Python

Science.gov (United States)

Kalb, V. L.; Collado-Vega, Y. M.; MacDonald, E.; Kosar, B.

2017-12-01

The Aurora Borealis and Australis Borealis are visually spectacular, but are also an indicator of Sun-magnetosphere-ionosphere energy transfer during geomagnetic storms. The Saint Patrick's Day Storm of 2015 is a stellar example of this, and is the focus of our study that utilizes the Geographical Information Services of ArcGIS to bring together diverse and cross disciplinary data for analysis. This research leverages data from a polar-orbiting Earth science sensor band that is exquisitely sensitive to visible light, namely the Day/Night Band (DNB) of the VIIRS instrument onboard the Suomi NPP satellite. This Sun-synchronous data source can provide high temporal and spatial resolution observations of the aurorae, which is not possible with current space science instruments. This data can be compared with auroral model data, solar wind measurements, and citizen science data of aurora observations and tweets. While the proposed data sources are diverse in type and format, their common attribute is location. This is exploited by bringing all the data into ArcGIS for mapping and analysis. The Python programming language is used extensively to automate the data preprocessing, group the DNB and citizen science observations to temporal windows associated with an auroral model timestep, and print the data to a pdf mapbook for sharing with team members. There are several goals for this study: compare the auroral model predictions with DNB data, look for fine-grained structure of the aurora in the DNB data, compare citizen science data with DNB values, and correlate DNB intensity with solar wind data. This study demonstrates the benefits of using a GIS platform to bring together data that is diverse in type and format for scientific exploration, and shows how Python can be used to scale up to large datasets.

Prototype high voltage bushing: Configuration to its operational demonstration

Energy Technology Data Exchange (ETDEWEB)

Shah, Sejal, E-mail: sshah@iter-india.org [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar 382428 (India); Sharma, D. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India); Parmar, D.; Tyagi, H.; Joshi, K.; Shishangiya, H.; Bandyopadhyay, M.; Rotti, C.; Chakraborty, A. [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

2016-12-15

High Voltage Bushing (HVB) is the key component of Diagnostic Neutral Beam (DNB) system of ITER as it provides access to high voltage electrical, hydraulic, gas and diagnostic feedlines to the beam source with isolation from grounded vessel. HVB also provides primary vacuum confinement for the DNB system. Being Safety Important Class (SIC) component of ITER, it involves several configurational, technological and operational challenges. To ensure its operational performance & reliability, particularly electrostatic behavior, half scale down Prototype High Voltage Bushing (PHVB) is designed considering same design criteria of DNB HVB. Design optimization has been carried out followed by finite element (FE) analysis to obtain DNB HVB equivalent electric stress on different parts of PHVB, taking into account all design, manufacturing & space constraints. PHVB was tested up to 60 kV without breakdown, which validates its design for the envisaged operation of 50 kV DC. This paper presents the design of PHVB, FEA validation, manufacturing constraints, experimental layout with interfacing auxiliary systems and operational results related to functional performance.

Leveraging CubeSat Technology to Address Nighttime Imagery Requirements over the Arctic

Science.gov (United States)

Pereira, J. J.; Mamula, D.; Caulfield, M.; Gallagher, F. W., III; Spencer, D.; Petrescu, E. M.; Ostroy, J.; Pack, D. W.; LaRosa, A.

2017-12-01

The National Oceanic and Atmospheric Administration (NOAA) has begun planning for the future operational environmental satellite system by conducting the NOAA Satellite Observing System Architecture (NSOSA) study. In support of the NSOSA study, NOAA is exploring how CubeSat technology funded by NASA can be used to demonstrate the ability to measure three-dimensional profiles of global temperature and water vapor. These measurements are critical for the National Weather Service's (NWS) weather prediction mission. NOAA is conducting design studies on Earth Observing Nanosatellites (EON) for microwave (EON-MW) and infrared (EON-IR) soundings, with MIT Lincoln Laboratory and NASA JPL, respectively. The next step is to explore the technology required for a CubeSat mission to address NWS nighttime imagery requirements over the Arctic. The concept is called EON-Day/Night Band (DNB). The DNB is a 0.5-0.9 micron channel currently on the operational Visible Infrared Imaging Radiometer Suite (VIIRS) instrument, which is part of the Suomi-National Polar-orbiting Partnership and Joint Polar Satellite System satellites. NWS has found DNB very useful during the long periods of darkness that occur during the Alaskan cold season. The DNB enables nighttime imagery products of fog, clouds, and sea ice. EON-DNB will leverage experiments carried out by The Aerospace Corporation's CUbesat MULtispectral Observation System (CUMULOS) sensor and other related work. CUMULOS is a DoD-funded demonstration of COTS camera technology integrated as a secondary mission on the JPL Integrated Solar Array and Reflectarray Antenna mission. CUMULOS is demonstrating a staring visible Si CMOS camera. The EON-DNB project will leverage proven, advanced compact visible lens and focal plane camera technologies to meet NWS user needs for nighttime visible imagery. Expanding this technology to an operational demonstration carries several areas of risk that need to be addressed prior to an operational mission

Monitoring method for steam generator operation

International Nuclear Information System (INIS)

Tamaoki, Tetsuo

1991-01-01

In an LMFBR plant having an once-through steam generator, reduction of life of a heat transfer pipe caused by heat cycle fatigue is monitored by early finding for the occurrence of abnormality in the inside of the steam generator and by continuous monitoring for the position of departure from nucleate boiling (DNB), which are difficult with existent static characteristic analysis codes. That is, RMS values of fluctuations in temperature signals sent from thermocouples for measuring the fluid temperature in the vicinity of heat transfer pipe disposed along a primary channel of the once-through type steam generator. The abnormality in heat transfer performance is monitored by the distribution change of the RMS values. Subsequently, DNB point on the side of water and steam is determined by the distribution of the RMS value. Then, accumulated values of the product between the time in which the starting point stays in the DNB region and a life consumption amount per unit time given in accordance with the operation condition are monitored. Accordingly, thermal fatigue failure of the heat transfer pipe due to temperature fluctuation in the DNB region is monitored. (I.S.)

A glassy carbon electrode modified with β-cyclodextin, multiwalled carbon nanotubes and graphene oxide for sensitive determination of 1,3-dinitrobenzene

International Nuclear Information System (INIS)

Li, Junhua; Feng, Haibo; Liu, Jinlong; Liu, Youcai; Jiang, Jianbo; Feng, Yonglan; Qian, Dong

2014-01-01

We are presenting a host-guest electrochemical platform for sensing the pollutant 1,3-dinitrobenzene. The method is based on the use of a glassy carbon electrode (GCE) covered with a composite made from multiwalled carbon nanotubes and graphene oxide, and functionalized with β-cyclodextrin (β-CD). The resultant composite was characterized by scanning electron microscopy, Fourier transform infrared spectroscopy, thermogravimetric analysis and electrochemical techniques. The modified GCE was used for the sensitive detection of 1,3-dinitrobenzene (DNB) at working voltages of −355 mV and −483 mV. Due to the specific recognition property of β-CD and the excellent electronic properties of the carbon nanomaterials, the electrode exhibits outstanding supramolecular recognition and enhanced electrochemical response to DNB compared to more conventional electrodes. Under optimum conditions, the peak currents vary linearly with the DNB concentrations in the range from 0.02 to 30.0 μM, and the detection limit is 5.0 nM (at an S/N of 3). The electrode exhibits long-term stability and has been successfully applied to the determination of DNB in spiked soil and water samples. (author)

33 CFR Schedule II to Subpart A of... - Table of Speeds 1

Science.gov (United States)

2010-07-01

..., Lt. 137 Deer Island, Lt. 186 11.5 10.5. 9. Deer Island, Lt. 186 Bartlett Point, Lt. 227 8.5 upb; 10.5 dnb 8 upb; 10.5 dnb. 10. Bartlett Point, Lt. 227 Tibbetts Point 13 10.5. 11. Junction of Canadian.... All other canals 6 6. 1 Maximum speeds at which a vessel may travel in identified areas in both normal...

Validation of S-NPP VIIRS Day-Night Band and M Bands Performance Using Ground Reference Targets of Libya 4 and Dome C

Science.gov (United States)

Chen, Xuexia; Wu, Aisheng; Xiong, Xiaoxiong; Lei, Ning; Wang, Zhipeng; Chiang, Kwofu

2015-01-01

This paper provides methodologies developed and implemented by the NASA VIIRS Calibration Support Team (VCST) to validate the S-NPP VIIRS Day-Night band (DNB) and M bands calibration performance. The Sensor Data Records produced by the Interface Data Processing Segment (IDPS) and NASA Land Product Evaluation and Algorithm Testing Element (PEATE) are acquired nearly nadir overpass for Libya 4 desert and Dome C snow surfaces. In the past 3.5 years, the modulated relative spectral responses (RSR) change with time and lead to 3.8% increase on the DNB sensed solar irradiance and 0.1% or less increases on the M4-M7 bands. After excluding data before April 5th, 2013, IDPS DNB radiance and reflectance data are consistent with Land PEATE data with 0.6% or less difference for Libya 4 site and 2% or less difference for Dome C site. These difference are caused by inconsistent LUTs and algorithms used in calibration. In Libya 4 site, the SCIAMACHY spectral and modulated RSR derived top of atmosphere (TOA) reflectance are compared with Land PEATE TOA reflectance and they indicate a decrease of 1.2% and 1.3%, respectively. The radiance of Land PEATE DNB are compared with the simulated radiance from aggregated M bands (M4, M5, and M7). These data trends match well with 2% or less difference for Libya 4 site and 4% or less difference for Dome C. This study demonstrate the consistent quality of DNB and M bands calibration for Land PEATE products during operational period and for IDPS products after April 5th, 2013.

Estimation of Minimum DNBR Using Cascaded Fuzzy Neural Networks

International Nuclear Information System (INIS)

Kim, Dong Yeong; Yoo, Kwae Hwan; Na, Man Gyun

2015-01-01

This phenomenon of boiling crisis is called a departure from nucleate boiling (DNB). The DNB phenomena can influence the fuel cladding and fuel pellets. The DNB ratio (DNBR) is defined as the ratio of the expected DNB heat flux to the actual fuel rod heat flux. Since it is very important to monitor and predict the minimum DNBR in a reactor core to prevent the boiling crisis and clad melting, a number of researches have been conducted to predict DNBR values. The aim of this study is to estimate the minimum DNBR in a reactor core using the measured signals of the reactor coolant system (RCS) by applying cascaded fuzzy neural networks (CFNN) according to operating conditions. Reactor core monitoring and protection systems require minimum DNBR prediction. The CFNN can be used to optimize the minimum DNBR value through the process of adding fuzzy neural networks (FNN) repeatedly. The proposed algorithm is trained by using the data set prepared for training (development data) and verified by using another data set different (independent) from the development data. The developed CFNN models were applied to the first fuel cycle of OPR1000. The RMS errors are 0.23% and 0.12% for the positive and negative ASI, respectively

COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1994-03-01

The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)

Active beam spectroscopy for ITER

International Nuclear Information System (INIS)

Von Hellermann, M.; Giroud, C.; Jaspers, R.; Hawkes, N.C.; Mullane, M.O.; Zastrow, K.D.; Krasilnikov, A.; Tugarinov, S.; Lotte, P.; Malaquias, A.; Rachlew, E.

2003-01-01

The latest status of 'Active Beam' related spectroscopy aspects as part of the ITER diagnostic scenario is presented. A key issue of the proposed scheme is based on the concept that in order to achieve the ultimate goal of global data consistency, all particles involved, that is, intrinsic and seeded impurity ions as well as helium ash ions and bulk plasma ions and also the plasma background data (e.g. magnetic and electric fields, electron density and temperature profiles) need to be addressed. A further sensible step in this direction is the decision of exploiting both a dedicated low-energy, low-power diagnostic beam (DNB, 2.2 MW 100 keV/amu) as well as the high-power, high-energy heating beams (HNB, 17 MW 500 keV/amu) for maximum diagnostic information. The authors report some new aspects referring to the use of DNB for motional Stark effect (MSE) where the main idea is to treat both beams (HNB and DNB) as potential diagnostic tools with complementary roles. The equatorial ports for the DNB promise excellent spatial resolution, however, the angles are less favourable for a polarimetric MSE exploitation. HNB can be used as probe beam for diagnosing slowing-down fusion alpha with a birth energy of 3,5 MeV

Ab initio kinetics and thermal decomposition mechanism of mononitrobiuret and 1,5-dinitrobiuret

Energy Technology Data Exchange (ETDEWEB)

Sun, Hongyan, E-mail: hongyan.sun1@gmail.com, E-mail: ghanshyam.vaghjiani@us.af.mil; Vaghjiani, Ghanshyam L., E-mail: hongyan.sun1@gmail.com, E-mail: ghanshyam.vaghjiani@us.af.mil [Propellants Branch, Rocket Propulsion Division, Aerospace Systems Directorate, Air Force Research Laboratory, AFRL/RQRP, 10 E. Saturn Blvd., Edwards AFB, California 93524 (United States)

2015-05-28

Mononitrobiuret (MNB) and 1,5-dinitrobiuret (DNB) are tetrazole-free, nitrogen-rich, energetic compounds. For the first time, a comprehensive ab initio kinetics study on the thermal decomposition mechanisms of MNB and DNB is reported here. In particular, the intramolecular interactions of amine H-atom with electronegative nitro O-atom and carbonyl O-atom have been analyzed for biuret, MNB, and DNB at the M06-2X/aug-cc-pVTZ level of theory. The results show that the MNB and DNB molecules are stabilized through six-member-ring moieties via intramolecular H-bonding with interatomic distances between 1.8 and 2.0 Å, due to electrostatic as well as polarization and dispersion interactions. Furthermore, it was found that the stable molecules in the solid state have the smallest dipole moment amongst all the conformers in the nitrobiuret series of compounds, thus revealing a simple way for evaluating reactivity of fuel conformers. The potential energy surface for thermal decomposition of MNB was characterized by spin restricted coupled cluster theory at the RCCSD(T)/cc-pV∞ Z//M06-2X/aug-cc-pVTZ level. It was found that the thermal decomposition of MNB is initiated by the elimination of HNCO and HNN(O)OH intermediates. Intramolecular transfer of a H-atom, respectively, from the terminal NH{sub 2} group to the adjacent carbonyl O-atom via a six-member-ring transition state eliminates HNCO with an energy barrier of 35 kcal/mol and from the central NH group to the adjacent nitro O-atom eliminates HNN(O)OH with an energy barrier of 34 kcal/mol. Elimination of HNN(O)OH is also the primary process involved in the thermal decomposition of DNB, which processes C{sub 2v} symmetry. The rate coefficients for the primary decomposition channels for MNB and DNB were quantified as functions of temperature and pressure. In addition, the thermal decomposition of HNN(O)OH was analyzed via Rice–Ramsperger–Kassel–Marcus/multi-well master equation simulations, the results of which

International Benchmark based on Pressurised Water Reactor Sub-channel and Bundle Tests. Volume III: Departure from Nucleate Boiling

International Nuclear Information System (INIS)

Rubin, Adam; Avramova, Maria; Velazquez-Lozada, Alexander

2016-03-01

This report summarised the second phase of the Nuclear Energy Agency (NEA) and the Nuclear Regulatory Commission (NRC) Benchmark Based on NUPEC PWR Sub-channel and Bundle Tests (PSBT), which was intended to provide data for the verification of Departure from Nucleate Boiling (DNB) prediction in existing thermal-hydraulics codes and provide direction in the development of future methods. This phase was composed of three exercises; Exercise 1: fluid temperature benchmark, Exercise 2: steady-state rod bundle benchmark and Exercise 3: transient rod bundle benchmark. The experimental data provided to the participants of this benchmark is from a series of void measurement tests using full-size mock-up tests for both BWRs and PWRs. These tests were performed from 1987 to 1995 by the Nuclear Power Engineering Corporation (NUPEC) in Japan and made available by the Japan Nuclear Energy Safety Organisation (JNES) for the purposes of this benchmark, which was organised by Pennsylvania State University. Nine institutions from seven countries participated in this benchmark. Nine different computer codes were used in Exercise 1, 2 and 3. Among the computer codes were porous media, sub-channel and systems thermal-hydraulic code. The improvement between FLICA-OVAP (sub-channel) and FLICA (sub-channel) was noticeable. The main difference between the two was that FLICA-OVAP implicitly assigned flow regime based on drift flux, while FLICA assumes single phase flows. In Exercises 2 and 3, the codes were generally able to predict the Departure from Nucleate Boiling (DNB) power as well as the axial location of the onset of DNB (for the steady-state cases) and the time of DNB (for the transient cases). It was noted that the codes that used the Electric-Power-Research- Institute (EPRI) Critical-Heat-Flux (CHF) correlation had the lowest mean error in Exercise 2 for the predicted DNB power

COOLOD, Steady-State Thermal Hydraulics of Research Reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-01-01

1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. 2 - Method of solution: The 'Heat Transfer Package' is a subprogram for calculating heat transfer coefficients, ONB temperature, heat flux at onset of flow instability and DNB heat flux. The 'Heat transfer package' was especially developed for research reactors which are operated under low pressure and low temperature conditions using plate-type fuel, just like the JRR-3M. Heat transfer correlations adopted in the 'Heat Transfer Package' were obtained or estimated based on the heat transfer experiments in which thermal-hydraulic features of the upgraded JRR-3 core were properly reflected. The 'Heat Transfer Package' is applicable to upward and downward flow

Blowdown heat transfer experiment, (1)

International Nuclear Information System (INIS)

Soda, Kunihisa; Yamamoto, Nobuo; Osaki, Hideki; Shiba, Masayoshi

1976-09-01

Blowdown heat transfer experiment has been carried out with a transparent test section to observe phenomena in coolant behavior during blowdown process. Experimental parameters are discharge position, initial system pressure, initial coolant temperature, power supply to heater rods and number of heater rods. At initial pressure 7-12 ata and initial power 6-50 kw per one heater rod, the flow condition in the test section is a major factor in determining time of DNB occurrence and physical process to DNB during blowdown. (auth.)

Subchronic toxicity studies on 1,3,5-trinitrobenzene, 1,3-dinitrobenzene and tetryl in rats. Final report

Energy Technology Data Exchange (ETDEWEB)

Reddy, T.V.; Daniel, F.B.

1994-09-01

Toxic effects of 1,3-Dinitrobenzene (1,3-DNB) in male and female F344 rats were evaluated by feeding powdered certified laboratory chow diet supplemented with varied concentrations of 1,3-DNB (0, 2.5, 10, 25, 75 and 150 mg/kg diet) for fourteen days. The average daily 1 ,3-DNB doses consumed were 0.21, 0.87, 2.02, 6.28 and 11.82 mg/kg b.w. for females and 0.21, 0.80, 1.98, 5.77 and 10.56 for males. Food consumption was significantly decreased in high dose animals of both sexes. Final body weights were not altered but relative organ weights were significantly changed in the 150 and 75 mg dose groups involving the spleen (males and females) and testes (males). Hematology and clinical chemistry studies indicated significantly increased values in both sexes relating to reticulocytes and methemoglobin in the 150 and 75 mg/kg dose groups while the red blood cell count, hemoglobin level and % hematocrit were decreased in these same groups. In addition, the levels of bilirubin, protein and albumin were increased in high dose males, Histopathological evaluations suggested that the susceptible organs for 1,3-DNB toxicity were kidneys (hyaline droplets), spleen (erythroid cell hyperplasia), brain (malacia and microgliosis), testes (seminiferous tubular degeneration). These changes were noted mainly in the 150 and 75 mg/kg dose groups except those changes involving the brain (150 mg/kg group only).

Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code

International Nuclear Information System (INIS)

Mur, J.; Meignin, J.C.

1997-07-01

Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.)

Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code

Energy Technology Data Exchange (ETDEWEB)

Mur, J. [Electricite de France (EDF), 78 - Chatou (France); Meignin, J.C. [Electricite de France (EDF), 69 - Villeurbanne (France)

1997-07-01

Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.) 8 refs.

The effect of grid assembly mixing vanes on critical heat flux values and azimuthal location in fuel assemblies

International Nuclear Information System (INIS)

De Crecy, F.

1994-01-01

Critical heat flux (CHF) is one of the limiting phenomena for a PWR. It has been widely studied for years, but many facts are still not satisfactorily understood. This paper deals with the effect of the grid assembly mixing vanes on both the value of the CHF and the azimuthal location of the departure from nucleate boiling (DNB). A series of experimental studies was performed on electrically heated, 5x5 square pitched, vertical rod bundles. Two specific grid assembly designs were used: with and without mixing vanes. DNB was detected by eight thermocouples welded internally in each rod at the same level in order to determine the azimuthal location. The coolant was Freon-12 flowing upwards to simulate high pressure water (as defined by Stevens). Single-phase flow experiments were also conducted to measure the exit temperature field in order to obtain the mixing coefficients for subchannel analysis.The results show very clearly that the mixing vanes have a significant effect on both the DNB azimuthal location and the CHF value. - Without mixing vanes, DNB occurs mainly on the most central rod and preferentially at the azimuthal location facing the adjacent rod. - With mixing vanes, DNB can occur on any of the nine central rods and is distributed in an apparently random way around the rod. -The effect of the mixing vanes on CHF is dramatic and depends a great deal on the parameter range (pressure, local mass velocity and local quality). Generally speaking, CHF with mixing vanes is significantly higher than without mixing vanes, but this effect can be inverted in some cases.In order to understand this fact more clearly, it is necessary to perform detailed analysis of subchannel behavior. Indeed, the analyses show that the magnitude of this effect is closely related to the mixing coefficients used. These mixing coefficients, estimated from the single-phase flow experiments, are subject to large uncertainties in two-phase flow. ((orig.))

An Indian test facility to characterise diagnostic neutral beam for ITER

International Nuclear Information System (INIS)

Singh, M.J.; Bandyopadhyay, M.; Rotti, C.; Singh, N.P.; Shah, Sejal; Bansal, G.; Gahlaut, A.; Soni, J.; Lakdawala, H.; Waghela, Harshad; Ahmed, I.; Roopesh, G.; Baruah, U.K.; Chakraborty, A.K.

2011-01-01

The diagnostic neutral beam (DNB) line shall be used to diagnose the He ash content in the D-T phase of the ITER machine using the charge exchange recombination spectroscopy (CXRS). Implementation of a successful DNB at ITER requires several challenges related to the production, neutralization and transport of the neutral beam over path lengths of 20.665 m, to be overcome. The delivery is aided if the above effects are tested prior to onsite commissioning. As DNB is a procurement package for INDIA, an ITER approved Indian test facility, INTF, is under construction at Institute for Plasma Research (IPR), India and is envisaged to be operational in 2015. The timeline for this facility is synchronized with the RADI, ELISE (IPP, Garching), SPIDER (RFX, Padova) in a manner that best utilization of configurational inputs available from them are incorporated in the design. This paper describes the facility in detail and discusses the experiments planned to optimise the beam transmission and testing of the beam line components using various diagnostics.

Improving Nocturnal Fire Detection with the VIIRS Day-Night Band

Science.gov (United States)

Polivka, Thomas N.; Wang, Jun; Ellison, Luke T.; Hyer, Edward J.; Ichoku, Charles M.

2016-01-01

Building on existing techniques for satellite remote sensing of fires, this paper takes advantage of the day-night band (DNB) aboard the Visible Infrared Imaging Radiometer Suite (VIIRS) to develop the Firelight Detection Algorithm (FILDA), which characterizes fire pixels based on both visible-light and infrared (IR) signatures at night. By adjusting fire pixel selection criteria to include visible-light signatures, FILDA allows for significantly improved detection of pixels with smaller and/or cooler subpixel hotspots than the operational Interface Data Processing System (IDPS) algorithm. VIIRS scenes with near-coincident Advanced Spaceborne Thermal Emission and Reflection (ASTER) overpasses are examined after applying the operational VIIRS fire product algorithm and including a modified "candidate fire pixel selection" approach from FILDA that lowers the 4-µm brightness temperature (BT) threshold but includes a minimum DNB radiance. FILDA is shown to be effective in detecting gas flares and characterizing fire lines during large forest fires (such as the Rim Fire in California and High Park fire in Colorado). Compared with the operational VIIRS fire algorithm for the study period, FILDA shows a large increase (up to 90%) in the number of detected fire pixels that can be verified with the finer resolution ASTER data (90 m). Part (30%) of this increase is likely due to a combined use of DNB and lower 4-µm BT thresholds for fire detection in FILDA. Although further studies are needed, quantitative use of the DNB to improve fire detection could lead to reduced response times to wildfires and better estimate of fire characteristics (smoldering and flaming) at night.

Experimental validation of prototype high voltage bushing

Science.gov (United States)

Shah, Sejal; Tyagi, H.; Sharma, D.; Parmar, D.; M. N., Vishnudev; Joshi, K.; Patel, K.; Yadav, A.; Patel, R.; Bandyopadhyay, M.; Rotti, C.; Chakraborty, A.

2017-08-01

Prototype High voltage bushing (PHVB) is a scaled down configuration of DNB High Voltage Bushing (HVB) of ITER. It is designed for operation at 50 kV DC to ensure operational performance and thereby confirming the design configuration of DNB HVB. Two concentric insulators viz. Ceramic and Fiber reinforced polymer (FRP) rings are used as double layered vacuum boundary for 50 kV isolation between grounded and high voltage flanges. Stress shields are designed for smooth electric field distribution. During ceramic to Kovar brazing, spilling cannot be controlled which may lead to high localized electrostatic stress. To understand spilling phenomenon and precise stress calculation, quantitative analysis was performed using Scanning Electron Microscopy (SEM) of brazed sample and similar configuration modeled while performing the Finite Element (FE) analysis. FE analysis of PHVB is performed to find out electrical stresses on different areas of PHVB and are maintained similar to DNB HV Bushing. With this configuration, the experiment is performed considering ITER like vacuum and electrical parameters. Initial HV test is performed by temporary vacuum sealing arrangements using gaskets/O-rings at both ends in order to achieve desired vacuum and keep the system maintainable. During validation test, 50 kV voltage withstand is performed for one hour. Voltage withstand test for 60 kV DC (20% higher rated voltage) have also been performed without any breakdown. Successful operation of PHVB confirms the design of DNB HV Bushing. In this paper, configuration of PHVB with experimental validation data is presented.

Energy metabolism and biotransformation as endpoints to pre-screen hepatotoxicity using a liver spheroid model

International Nuclear Information System (INIS)

Xu Jinsheng; Purcell, Wendy M.

2006-01-01

The current study investigated liver spheroid culture as an in vitro model to evaluate the endpoints relevant to the status of energy metabolism and biotransformation after exposure to test toxicants. Mature rat liver spheroids were exposed to diclofenac, galactosamine, isoniazid, paracetamol, m-dinitrobenzene (m-DNB) and 3-nitroaniline (3-NA) for 24 h. Pyruvate uptake, galactose biotransformation, lactate release and glucose secretion were evaluated after exposure. The results showed that pyruvate uptake and lactate release by mature liver spheroids in culture were maintained at a relatively stable level. These endpoints, together with glucose secretion and galactose biotransformation, were related to and could reflect the status of energy metabolism and biotransformation in hepatocytes. After exposure, all of the test agents significantly reduced glucose secretion, which was shown to be the most sensitive endpoint of those evaluated. Diclofenac, isoniazid, paracetamol and galactosamine reduced lactate release (P < 0.01), but m-DNB increased lactate release (P < 0.01). Diclofenac, isoniazid and paracetamol also reduced pyruvate uptake (P < 0.01), while galactosamine had little discernible effect. Diclofenac, galactosamine, paracetamol and m-DNB also reduced galactose biotransformation (P < 0.01), by contrast, isoniazid did not. The metabolite of m-DNB, 3-NA, which served as a negative control, did not cause significant changes in lactate release, pyruvate uptake or galactose biotransformation. It is concluded that pyruvate uptake, galactose biotransformation, lactate release and glucose secretion can be used as endpoints for evaluating the status of energy metabolism and biotransformation after exposure to test agents using the liver spheroid model to pre-screen hepatotoxicity

Pemetaan daerah perikanan lampu (light fishing menggunakan data viirs day-night band di perairan Pandeglang Provinsi Banten

Directory of Open Access Journals (Sweden)

Adi Susanto

2015-08-01

Full Text Available Abstract. Light fishing in Pandeglang Banten has significantly developed in number of fishing fleet and using of lamp technology. The fishing ground of light fishing fleet dispersed from Labuan until Taman Nasional Ujung Kulon. The aim of this research is to map the fishing ground of light fishing using VIIRS-DNB data at August to November 2014. This research use descriptive analysis with case study on fishing ground of ligt fishing using VIIRS-DNB data in Pandeglang waters. The results show the fishing ground of light fishing at August to November spread from Lada Bay, Lesung Cape, Sumur District, selamat Datang Bay, until Panaitan Strait. In November, the numbers of fishing fleet have significantly decreased. The loft ofwind velocity in Sunda Strait caused big waves in Pandeglang waters. Moreover, light fishing fleet also move from coastal water to the sea to find the ideal depth.The fluctuation of fishing ground at August to November 2014 influenced by monsoon circulation that effectto current and water masses circulation there. It cause the changing of surface water fertility and influent to spreading of pelagic fish fishing ground as a main target of light fishing fleets in Pandeglang waters. Keyword: fishing ground; light fishing; Pandeglang; VIIRS-DNB Abstrak. Perikanan lampu di perairan Pandeglang Banten telah mengalami perkembangan yang signifikan baik dalam jumlah armada maupun teknologi lampu yang digunakan. Daerah penangkapan armada perikanan lampu tersebar mulai dari perairan Labuan hingga Taman Nasional Ujung Kulon. Penelitian ini bertujuan untuk memetakan daerah perikanan lampu di perairan Pandeglang menggunakan data VIIRS-DNB yang mampu mendeteksi radiasi yang dihasilkan oleh lampu yang digunakan untuk menarik perhatian ikan. Metode penelitian yang digunakan adalah deskriptif dengan studi kasus berupa sebaran daerah perikanan lampu di perairan Pandeglang menggunakan data VIIRS-DNB. Hasil analisis menunjukkan bahwa pada Bulan

OECD/NRC Benchmark Based on NUPEC PWR Sub-channel and Bundle Test (PSBT). Volume I: Experimental Database and Final Problem Specifications

International Nuclear Information System (INIS)

Rubin, A.; Schoedel, A.; Avramova, M.; Utsuno, H.; Bajorek, S.; Velazquez-Lozada, A.

2012-01-01

The need to refine models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear applications. The needs arising in this respect should not be limited to the currently available macroscopic methods but should be extended to next-generation analysis techniques that focus on more microscopic processes. One of the most valuable databases identified for the thermal-hydraulics modelling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan, which includes sub-channel void fraction and departure from nucleate boiling (DNB) measurements in a representative Pressurised Water Reactor (PWR) fuel assembly. Part of this database has been made available for this international benchmark activity entitled 'NUPEC PWR Sub-channel and Bundle Tests (PSBT) benchmark'. This international project has been officially approved by the Japanese Ministry of Economy, Trade, and Industry (METI), the US Nuclear Regulatory Commission (NRC) and endorsed by the OECD/NEA. The benchmark team has been organised based on the collaboration between Japan and the USA. A large number of international experts have agreed to participate in this programme. The fine-mesh high-quality sub-channel void fraction and departure from nucleate boiling data encourages advancement in understanding and modelling complex flow behaviour in real bundles. Considering that the present theoretical approach is relatively immature, the benchmark specification is designed so that it will systematically assess and compare the participants' analytical models on the prediction of detailed void distributions and DNB. The development of truly mechanistic models for DNB prediction is currently underway. The benchmark problem includes both macroscopic and microscopic measurement data. In this context, the sub-channel grade void fraction data are regarded as the macroscopic data and the digitised computer graphic images are the

Long-term impact of earthquakes on sleep quality.

Science.gov (United States)

Tempesta, Daniela; Curcio, Giuseppe; De Gennaro, Luigi; Ferrara, Michele

2013-01-01

We investigated the impact of the 6.3 magnitude 2009 L'Aquila (Italy) earthquake on standardized self-report measures of sleep quality (Pittsburgh Sleep Quality Index, PSQI) and frequency of disruptive nocturnal behaviours (Pittsburgh Sleep Quality Index-Addendum, PSQI-A) two years after the natural disaster. Self-reported sleep quality was assessed in 665 L'Aquila citizens exposed to the earthquake compared with a different sample (n = 754) of L'Aquila citizens tested 24 months before the earthquake. In addition, sleep quality and disruptive nocturnal behaviours (DNB) of people exposed to the traumatic experience were compared with people that in the same period lived in different areas ranging between 40 and 115 km from the earthquake epicenter (n = 3574). The comparison between L'Aquila citizens before and after the earthquake showed a significant deterioration of sleep quality after the exposure to the trauma. In addition, two years after the earthquake L'Aquila citizens showed the highest PSQI scores and the highest incidence of DNB compared to subjects living in the surroundings. Interestingly, above-the-threshold PSQI scores were found in the participants living within 70 km from the epicenter, while trauma-related DNBs were found in people living in a range of 40 km. Multiple regressions confirmed that proximity to the epicenter is predictive of sleep disturbances and DNB, also suggesting a possible mediating effect of depression on PSQI scores. The psychological effects of an earthquake may be much more pervasive and long-lasting of its building destruction, lasting for years and involving a much larger population. A reduced sleep quality and an increased frequency of DNB after two years may be a risk factor for the development of depression and posttraumatic stress disorder.

The 2015 Academic College of Emergency Experts in India's INDO-US Joint Working Group White Paper on Establishing an Academic Department and Training Pediatric Emergency Medicine Specialists in India

Science.gov (United States)

Mahajan, Prashant; Batra, Prerna; Shah, Binita R; Saha, Abhijeet; Galwankar, Sagar; Aggrawal, Praveen; Hassoun, Ameer; Batra, Bipin; Bhoi, Sanjeev; Kalra, Om Prakash; Shah, Dheeraj

2015-01-01

The concept of pediatric emergency medicine (PEM) is virtually nonexistent in India. Suboptimally, organized prehospital services substantially hinder the evaluation, management, and subsequent transport of the acutely ill and/or injured child to an appropriate facility. Furthermore, the management of the ill child at the hospital level is often provided by overburdened providers who, by virtue of their training, lack experience in the skills required to effectively manage pediatric emergencies. Finally, the care of the traumatized child often requires the involvement of providers trained in different specialities, which further impedes timely access to appropriate care. The recent recognition of Doctor of Medicine (MD) in Emergency Medicine (EM) as an approved discipline of study as per the Indian Medical Council Act provides an unprecedented opportunity to introduce PEM as a formal academic program in India. PEM has to be developed as a 3-year superspeciality course (in PEM) after completion of MD/Diplomate of National Board (DNB) Pediatrics or MD/DNB in EM. The National Board of Examinations (NBE) that accredits and administers postgraduate and postdoctoral programs in India also needs to develop an academic program – DNB in PEM. The goals of such a program would be to impart theoretical knowledge, training in the appropriate skills and procedures, development of communication and counseling techniques, and research. In this paper, the Joint Working Group of the Academic College of Emergency Experts in India (JWG-ACEE-India) gives its recommendations for starting 3-year DM/DNB in PEM, including the curriculum, infrastructure, staffing, and training in India. This is an attempt to provide an uniform framework and a set of guiding principles to start PEM as a structured superspeciality to enhance emergency care for Indian children. PMID:26807394

New Departure from Nucleate Boiling model relying on first principle energy balance at the boiling surface

Science.gov (United States)

Demarly, Etienne; Baglietto, Emilio

2017-11-01

Predictions of Departure from Nucleate Boiling have been a longstanding challenge when designing heat exchangers such as boilers or nuclear reactors. Many mechanistic models have been postulated over more than 50 years in order to explain this phenomenon but none is able to predict accurately the conditions which trigger the sudden change of heat transfer mode. This work aims at demonstrating the pertinence of a new approach for detecting DNB by leveraging recent experimental insights. The new model proposed departs from all the previous models by making the DNB inception come from an energy balance instability at the heating surface rather than a hydrodynamic instability of the bubbly layer above the surface (Zuber, 1959). The main idea is to modulate the amount of heat flux being exchanged via the nucleate boiling mechanism by the wetted area fraction on the surface, thus allowing a completely automatic trigger of DNB that doesn't require any parameter prescription. This approach is implemented as a surrogate model in MATLAB in order to validate the principles of the model in a simple and controlled geometry. Good agreement is found with the experimental data leveraged from the MIT Flow Boiling at various flow regimes. Consortium for Advanced Simulation of Light Water Reactors (CASL).

Supporting Disaster Assessment and Response with the VIIRS Day-Night Band

Science.gov (United States)

Schultz, Lori A.; Cole, Tony; Molthan, Andrew L.

2015-01-01

When meteorological or man-made disasters occur, first responders often focus on impacts to the affected population and other human activities. Often, these disasters result in significant impacts to local infrastructure and power, resulting in widespread power outages. For minor events, these power outages are often short-lived, but major disasters often include long-term outages that have a significant impact on wellness, safety, and recovery efforts within the affected areas. Staff at NASA's Short-term Prediction Research and Transition (SPoRT) Center have been investigating the use of the VIIRS day-night band for monitoring power outages that result from significant disasters, and developing techniques to identify damaged areas in near real-time following events. In addition to immediate assessment, the VIIRS DNB can be used to monitor and assess ongoing recovery efforts. In this presentation, we will highlight previous applications of the VIIRS DNB following Superstorm Sandy in 2012, and other applications of the VIIRS DNB to more recent disaster events, including detection of outages following the Moore, Oklahoma tornado of May 2013 and the Chilean earthquake of April 2014. Examples of current products will be shown, along with future work and other goals for supporting disaster assessment and response with VIIRS capabilities.

Flow transients experiments with refrigerant-12

International Nuclear Information System (INIS)

Celata, G.P.; D'Annibale, F.; Farello, G.E.; Setaro, T.

1986-01-01

Flow transients have been investigated in a wide range of thermal-hydraulics situations with Refrigerannt-12. Six pressures (including the reference to PWR and BWR characteristic liquid to vapour densities ratios), several periods of the flowrate transients coastdown during the simulated flow decays, and different specific mass flowrate have been studied emploiyng a circular duct test section (Dsub(i)=7,5 mm). Two heated lengths of the test section have been considered (L = 2300 and 1180 mm). Experimental data have shown the complete inadequacy of steady-state critical heat flux correlations in predicting the onset of boiling crisis during fast flow transients (half-flow decay time, tsub(h)lt5.0-6.0 s). The flow transient does not show dependence, in terms of DNB conditions ,upon the length of the test section: the ratio between transient and steady-state critical mass flowrate is not dependent on the tested geometry. The time interval from the start of the flowrate transient to the onset of DNB (time to crisis), has been experimentally determined for all the runs. Data analysis for a better theoretical prediction of the phenomenon has been accomplished, and a design correlation for DNB conditons and time to crisis prediction has been proposed

Development of an optimization technique of CETOP-D inlet flow factor for reactor core thermal margin improvement

International Nuclear Information System (INIS)

Hong, Sung Duk; Im, Jong Sun; Yoo, Yun Jong; Kwon, Jung Taek; Park, Jong Ryool

1995-01-01

The recent ABB/CE(Asea Brown Boveri Combustion Engineering) type pressurized water reactors have the on-line monitoring system, i.e., the COLSS(core operating limit supervisory system), to prevent the specified acceptable fuel design limits from being violated during normal operation and anticipated operational occurrences. One of the main functions of COLSS is the on-line monitoring of the DNB(departure from nucleate boiling) overpower margin by calculating the MDNBR(minimum DNB ratio) for the measured operating condition at every second. The CETOP-D model, used in the MDNBR calculation of COLSS, is benchmarked conservatively against the TORC model using an inlet flow factor of hot assembly in CETOP-D as an adjustment factor for TORC. In this study, a technique to optimize the CETOP-D inlet flow factor has been developed by eliminating the excessive conservatism in the ABB/CE's. A correlation is introduced to account for the actual variation of the CETOP-D inlet flow factor within the core operating limits. This technique was applied to the core operating range of the Yonggwang Units 3 and 4 Cycle 1, which results in the increase of 2% in the DNB overpower margin at the normal operating condition, compared with that from the ABB/CE method. 7 figs., 2 tabs., 10 refs. (Author)

Characterization of pitches by liquid chromatography using cellulose 3,5-dinitrobenzoate as the packing material

Energy Technology Data Exchange (ETDEWEB)

Arai, K.; Judo, R.; Ota, E. [Gunma University, Gunma (Japan). Dept. of Chemistry

1997-08-01

Characterization of coal tar, petroleum and PVC pitches by a liquid chromatography using cellulose 3,5-dinitrobenzoate (DNB-cellulose) as the packing material was investigated. Separation mechanism based on charge-transfer interaction between the dinitrobenzoyl group and polyaromatic compounds was expected to be useful for separation of the constituents of the pitches. First, 26 model polyaromatic compounds were tested to examine the characteristic feature of the packing material by liquid chromatography. The compounds were found to be classified roughly into four groups with different retention volume, principally according to the number of condensed rings. The nonplanar structure and aliphatic side chain of the polyaromatic compounds also affected the separation behavior. Both benzene soluble-hexane soluble and benzene soluble-hexane insoluble fractions of the three pitches were separated on DNB-cellulose. It was found that coal tar pitch contains relatively large amounts of some highly condensed polyaromatic compounds with condensed rings of 4 to 5; petroleum pitch has small amounts of such specific highly condensed polyaromatic compounds, while PVC pitch has large amounts of less condensed polyaromatic compounds and there is no significant amount of highly condensed compound in it. Thus DNB-cellulose was useful as the convenient packing material for liquid chromatography to characterize pitches.

Some leaders, despite reputations, offer Latvians hope through crisis / Arta Ankrava

Index Scriptorium Estoniae

Ankrava, Arta

2009-01-01

DnB Nord Latvia Barometer küsitlusest selgus, et kõige enam usaldatakse ekspresident Vaira Vike-Freiberga, Riia linnapea Nils Usakovsi ja Ventspilsi endise linnapea Aivars Lembergsi majandusprognoose

Study on the thermohydraulic characteristic of marine nuclear reactors, 6

International Nuclear Information System (INIS)

Kurosawa, Akira; Otsuji, Tomoo; Iwahori, Koji

1981-01-01

The objective of this research was to observe the bubble behaviour at the location of boiling crisis, middle stream, upper stream in low pressure, subcooled flow boiling with high speed motion and still photography. The observations included measurements from photographs of bubbles and photographic records of the flow structure before, during, and after DNB for Freon-113, 1 ton/hr 45 0 C subcooling at 3 kg/cm 2 . The experimental conditions covered the gravity acceleration, flow modulation and steady state condition. The dominant flow mechanisms at DNB in each case were compared for conditions tested on the basis of the photographic information. The phenomena of CHF decrease by gravity acceleration condition were observed and judged. The necessity for still more useful quantitative informations was cleared. (author)

Stapedotomy and its effect on hearing – our experience with 54 cases.

African Journals Online (AJOL)

Nigeria, and clinical attaché Indorewala ENT Hospital, DNB Institution and Research, behind Mahamarg Bus Stand ..... financial support from any individual, group or organiza- tion. ... Comparative result in the Chinese Population in Taiwan.

Loss-of-coolant accident test series TC-1 experiment operating specifications

International Nuclear Information System (INIS)

Yackle, T.R.

1979-09-01

The purpose of this document is to specify the experiment operating procedure for the test series TC-1. The effects of externally mounted cladding thermocouples on the fuel rod thermal behavior during LOCA blowdown and reflood cycles will be investigated in the test. Potential thermocouple effects include: (a) delayed DNB, (b) momentary cladding rewets following DNB, (c) premature cladding rewet during a blowdown two-phase slug period, and (d) early cladding rewet during reflood. The two-phase slug period will be controlled by momentarily opening the hot leg valve. The slug will consist of lower plenum liquid that is sent through the flow shrouds and will be designed to quench the fuel rods at a rate that is similar to the slug experienced early in the LOFT L2-2 and L2-3 tests

COOLOD-N: a computer code, for the analyses of steady-state thermal-hydraulics in plate-type research reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1990-02-01

The COOLOD-N code provides a capability for the analysis of the steady-state thermal-hydraulics of research reactors in which plate-type fuel is employed. This code is revised version of the COOLOD code, and is applicable not only to a forced convection cooling mode, but also to a natural convection cooling mode. In the code, a function to calculate flow rate under a natural convection, and a heat transfer package which was a subroutine program to calculate heat transfer coefficient, ONB temperature and DNB heat flux, and was especially developed for the upgraded JRR-3, have been newly added to the COOLOD code. The COOLOD-N code also has a capability of calculating the heat flux at onset of flow instability as well as DNB heat flux. (author)

Talks on nuclear plant break down

Index Scriptorium Estoniae

2007-01-01

Kasvavad probleemid Leedu tuumaelektrijaama projektiga. Läbirääkimistest Leedu valitsuse ja NDX Energija vahel. DnB Nord Bankas nimetas projekti liiga riskantseks ja kulukaks. Lisa: Eesti Energia board member suggests plant for Estonia

The capacitor banks for the text diagnostic neutral beam and electron cyclotron heating experiments

International Nuclear Information System (INIS)

Nelin, K.; Jagger, J.; Baker, M.; Ourou, A.; De Turk, P.

1986-01-01

The Texas Experimental Tokamak (TEXT) has been operational since November of 1980. Since that time, many experimental systems have been added to the machine. Currently, two major experiments are being added to compliment the diagnostics already online. These systems, the Diagnostic Neutral Beam (DNB) and the Electron Cyclotron Heating (ECH) experiments are described in separate papers. A set of five modular, bipolar capacitor banks are used to power both the DNB and the ECH. The total capacitance of the banks is 92μF. The stored energy is about 500kJ at+or-100kV. The banks are built as five identical, interchangeable modules. One module is adequate to run the DNB. Up to four banks are used to power the ECH. The banks are portable so that they can be moved to the open end of the laboratory for maintenance. This gives much better access for repair work and allows the experiments to continue to run with the remaining banks. Due to budgetary constraints, these banks were constructed in the most economical manner possible consistent with worker safety and long term reliability. The capacitors themselves are on loan from Los Alamos National Labs. They are rated at 1.85μF at 60kV. Our application requires that they be used in a series/parallel configuration with a peak voltage of 50kV each. This paper describes the electrical, mechanical and control design considerations required to achieve a working set of banks

Synergistic Use of Nighttime Satellite Data, Electric Utility Infrastructure, and Ambient Population to Improve Power Outage Detections in Urban Areas

Directory of Open Access Journals (Sweden)

Tony A. Cole

2017-03-01

Full Text Available Natural and anthropogenic hazards are frequently responsible for disaster events, leading to damaged physical infrastructure, which can result in loss of electrical power for affected locations. Remotely-sensed, nighttime satellite imagery from the Suomi National Polar-orbiting Partnership (Suomi-NPP Visible Infrared Imaging Radiometer Suite (VIIRS Day/Night Band (DNB can monitor power outages in disaster-affected areas through the identification of missing city lights. When combined with locally-relevant geospatial information, these observations can be used to estimate power outages, defined as geographic locations requiring manual intervention to restore power. In this study, we produced a power outage product based on Suomi-NPP VIIRS DNB observations to estimate power outages following Hurricane Sandy in 2012. This product, combined with known power outage data and ambient population estimates, was then used to predict power outages in a layered, feedforward neural network model. We believe this is the first attempt to synergistically combine such data sources to quantitatively estimate power outages. The VIIRS DNB power outage product was able to identify initial loss of light following Hurricane Sandy, as well as the gradual restoration of electrical power. The neural network model predicted power outages with reasonable spatial accuracy, achieving Pearson coefficients (r between 0.48 and 0.58 across all folds. Our results show promise for producing a continental United States (CONUS- or global-scale power outage monitoring network using satellite imagery and locally-relevant geospatial data.

Simulation of fuel rod behaviour during various break LOCAs in PWRs

International Nuclear Information System (INIS)

Gadalla, A.A.; El-Fawal, M.M.

1996-01-01

During loss of coolant accident (LOCAs) course of events, attention focuses on fuel rod cladding temperature behaviour. In this study, the DRUFAN analytical model and LOBI-MOD2 experimental modeling scheme for fuel rod temperature behaviour during C L-Break LOCA in PWRs, are described and discussed. These models are applied for the investigation of fuel rod cladding temperature behaviour during LOCA blowdown phase. A spectrum of selected values representing small, intermediate and large CL- Break sizes are considered in the predictions. The results of the predictions demonstrated that calculated heater rod temperature at steady state as well as the transient period up to 1000 sec are going in good agreement with the measured values. However above 1000 sec the calculated temperatures are higher than the measured values. This indicates that code predictions in this period are conservative. The results indicated also that, in case of small CL-break LOCA (0.01 A and 0.01 and 0.03 A), the heater rod cladding temperature don't rise above saturation temperature. However, on the top of the heater rod, DNB is occurred in case of 0.03 A CL break, while for 0.01 A break, DNB didn't occur. In case of intermediate and large CL-break; (0.05 A, 0.10 A and 1 A), the results showed that, the heater rod cladding temperature exceeded the saturation temperature and DNB prevailed in upper and intermediate sections of the core. 15 figs., 2 tabs

Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

International Nuclear Information System (INIS)

Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon

2006-02-01

The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation

On-line thermal margin estimation of a PWR core using a neural network approach

International Nuclear Information System (INIS)

Park, Soon Ok; Kim, Hyun Koon; Lee, Seung Hynk; Chang, Soon Heung

1992-01-01

A new approach for on-line thermal margin monitoring of a PWR Core is proposed in this paper, where a neural network model is introduced to predict the DNBR values at the given reactor operating conditions. The neural network is learned by the Back Propagation algorithm with the optimized random training data and is tested to investigate the generalized performance for the steady state operating region as well as for the transient situations where DNB is of the primary concern. The test results show that the high level of accuracy in predicting the DNBR can be achieved by the neural network model compared to the detailed code results. An insight has been gained from this study that the neural network model for estimating DNB performance can be a viable tool for on-line thermal margin monitoring of a nuclear power plant

Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2006-02-15

The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation.

Financial Literacy, Retirement Planning, and Household Portfolio Behavior: Four Empirical Contributions

NARCIS (Netherlands)

van Rooij, M.C.J.

2009-01-01

This thesis provides empirical evidence on financial skills and the relation with household financial decision-making based upon specially designed questions for the DNB Household Survey (DHS). The majority of the respondents has some grasp of concepts such as interest compounding, inflation, and

Ultrasonic intensification of electrochemical destruction of 1,3-dinitrobenzene and 2,4-dinitrotoluene with ozone and electrocoagulation of azo-dyes

Energy Technology Data Exchange (ETDEWEB)

Abramov, V.O.; Abramov, O.V.; Kuznetsov, V.M. [Russian Academy of Sciences, Lab. of Ultrasonics, Moscow (Russian Federation). Kumakov Inst. of General and Inorganic Chemistry

2002-07-01

For the detoxification of waste and sewage, oxidation of toxic components using strong and environmentally-friendly oxidants such as hydrogen peroxide or ozone in combination with additional physicochemical processes such as ultraviolet radiation, electric discharge and ultrasonic irradiation (advanced oxidation processes) is considered to be promising. The presence of the electron-withdrawing nitro group substantially reduces the reactivity of nitroaromatics in oxidation reactions. Therefore, even when using ozone, an acceptable rate of destruction of some stable compounds such as 1,3-dinitrobenzene (DNB), 2,4-dinitrotoluene (DNT) or TNT, cannot be achieved. We have previously found that the oxidation of organic compounds by ozone or a combination of ozone with hydrogen peroxide in an ultrasonic field is enhanced in a low electric field. The objective of the present work is to study the possibility of the oxidation of DNB and DNT by ozone in an electrochemical cell under ultrasonic irradiation. (orig.)

Protection set-points lines for the reactor core and considerations about power distribution and peak factors

International Nuclear Information System (INIS)

Furieri, E.B.

1981-01-01

In order to assure the reactor core integrity during the slow operational transients (power excursion above the nominal value and the high coolant temperature), the formation of a steam film (DNB-Departure from Nucleate Boiling) in the control rods must be avoided. The protection set points lines presents the points where DNBR (relation between critical heat flux-q sub(DNB) and the local heat flux-q' sub(local) is equal to 1.30, corrected by peak factors and uncertainty in function of ΔTr and T sub(R), respectively coolant elevation and medium coolant temperature in reactor pressure vessel. The curve set-points were determined using a new version of COBRA-IIIF (CUPRO) computer code, implemented with new subroutines and linearized convergence scheme. Pratical results for Angra-1 core were obtained and its were compared with the results from the fabricator. (E.G.) [pt

Reference Computational Meshing Strategy for Computational Fluid Dynamics Simulation of Departure from Nucleate BoilingReference Computational Meshing Strategy for Computational Fluid Dynamics Simulation of Departure from Nucleate Boiling

Energy Technology Data Exchange (ETDEWEB)

Pointer, William David [ORNL

2017-08-01

The objective of this effort is to establish a strategy and process for generation of suitable computational mesh for computational fluid dynamics simulations of departure from nucleate boiling in a 5 by 5 fuel rod assembly held in place by PWR mixing vane spacer grids. This mesh generation process will support ongoing efforts to develop, demonstrate and validate advanced multi-phase computational fluid dynamics methods that enable more robust identification of dryout conditions and DNB occurrence.Building upon prior efforts and experience, multiple computational meshes were developed using the native mesh generation capabilities of the commercial CFD code STAR-CCM+. These meshes were used to simulate two test cases from the Westinghouse 5 by 5 rod bundle facility. The sensitivity of predicted quantities of interest to the mesh resolution was then established using two evaluation methods, the Grid Convergence Index method and the Least Squares method. This evaluation suggests that the Least Squares method can reliably establish the uncertainty associated with local parameters such as vector velocity components at a point in the domain or surface averaged quantities such as outlet velocity magnitude. However, neither method is suitable for characterization of uncertainty in global extrema such as peak fuel surface temperature, primarily because such parameters are not necessarily associated with a fixed point in space. This shortcoming is significant because the current generation algorithm for identification of DNB event conditions relies on identification of such global extrema. Ongoing efforts to identify DNB based on local surface conditions will address this challenge

Coastal flood risk

CSIR Research Space (South Africa)

Luck-Vogel, Melanie

2017-07-01

Full Text Available ! Unless… People @ Coasts https://eoimages.gsfc.nasa.gov/images/imagerecords/79000/79765/dnb_land_ocean_ice.2012.3600x1800.jpg People & Coasts • About 40% of the world’s population is situated within 100km of the coastline (Millennium Ecosystem Assessment...

Financial literacy and stock market participation

NARCIS (Netherlands)

van Rooij, Maarten; Lusardi, Annamaria; Alessie, Rob

We have devised two special modules for De Nederlandsche Bank (DNB) Household Survey to measure financial literacy and study its relationship to stock market participation. We find that the majority of respondents display basic financial knowledge and have some grasp of concepts such as interest

75 FR 33445 - U.S. Citizenship and Immigration Services Fee Schedule

Science.gov (United States)

2010-06-11

... comprehensive fee study and refined its cost accounting process, and determined that current fees do not recover... State. DNB--Dun and Bradstreet. EAD--Employment Authorization Document. FASAB--Federal Accounting... naturalization benefits and ensures the collection, safeguarding, and accounting of fees by USCIS. INA section...

Supervision on takeover bids : a comparison of regulatory arrangements

NARCIS (Netherlands)

Haan-Kamminga, Avelien

2006-01-01

Toezicht is een thema dat de laatste jaren veel aandacht heeft gekregen en niet zonder reden. Een aantal jaren geleden ging een grote verzekeringsmaatschappij failliet (Vie d'Or), terwijl zij onder toezicht stond van de Verzekeringskamer (nu DNB) had kunnen voorkomen dat de maatschappij failliet

PICTORIAL ESSAY Ultrasound diagnosis of ulnar nerve dislocation ...

African Journals Online (AJOL)

J Patil, MD, DNB. Department of Radiology, Apple Hospital, Kolhapur, Maharashtra, India. Corresponding author: V ... suspected nerve entrapment is magnetic resonance imaging (MRI). Alternatively, high-resolution ultrasound ... The site of the common origin of the flexor muscles of the forearm was identified at the apex of ...

Synthesis, spectroscopic characterization and structural investigations of a new charge transfer complex of 2,6-diaminopyridine with 3,5-dinitrobenzoic acid: DNA binding and antimicrobial studies

Science.gov (United States)

Khan, Ishaat M.; Ahmad, Afaq; Kumar, Sarvendra

2013-03-01

A new charge transfer (CT) complex [(DAPH)+(DNB)-] consisting of 2,6-diaminopyridine (DAP) as donor and 3,5-dinitrobenzoic acid (DNB-H) as acceptor, was synthesized and characterized by FTIR, 1H and 13C NMR, ESI mass spectroscopic and X-ray crystallographic techniques. The hydrogen bonding (N+-H⋯O-) plays an important role to consolidate the cation and anion together. CT complex shows a considerable interaction with Calf thymus DNA. The CT complex was also tested for its antibacterial activity against two Gram-positive bacteria Staphylococcus aureus and Bacillus subtilis and two Gram-negative bacteria Escherichia coli and Pseudomonas aeruginosa strains by using Tetracycline as standard, and antifungal property against Aspergillus niger, Candida albicans, and Penicillium sp. by using Nystatin as standard. The results were compared with standard drugs and significant conclusions were obtained. A polymeric net work through H-bonding interactions between neighboring moieties was observed. This has been attributed to the formation of 1:1 type CT complex.

Assessment of core protection and monitoring systems for an advanced reactor SMART

International Nuclear Information System (INIS)

In, Wang Kee; Hwang, Dae Hyun; Yoo, Yeon Jong; Zee, Sung Qunn

2002-01-01

Analogue and digital core protection/monitoring systems were assessed for the implementation in an advanced reactor. The core thermal margins to nuclear fuel design limits (departure from nucleate boiling and fuel centerline melting) were estimated using the design data for a commercial pressurized water reactor and an advanced reactor. The digital protection system resulted in a greater power margin to the fuel centerline melting by at least 30% of rated power for both commercial and advanced reactors. The DNB margin with the digital system is also higher than that for the analogue system by 8 and 12.1% of rated power for commercial and advanced reactors, respectively. The margin gain with the digital system is largely due to the on-line calculations of DNB ratio and peak local power density from the live sensor signals. The digital core protection and monitoring systems are, therefore, believed to be more appropriate for the advanced reactor

Uncertainty analysis for hot channel

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2006-01-01

The fulfillment of the safety analysis acceptance criteria is usually evaluated by separate hot channel calculations using the results of neutronic or/and thermo hydraulic system calculations. In case of an ATWS event (inadvertent withdrawal of control assembly), according to the analysis, a number of fuel rods are experiencing DNB for a longer time and must be regarded as failed. Their number must be determined for a further evaluation of the radiological consequences. In the deterministic approach, the global power history must be multiplied by different hot channel factors (kx) taking into account the radial power peaking factors for each fuel pin. If DNB occurs it is necessary to perform a few number of hot channel calculations to determine the limiting kx leading just to DNB and fuel failure (the conservative DNBR limit is 1.33). Knowing the pin power distribution from the core design calculation, the number of failed fuel pins can be calculated. The above procedure can be performed by conservative assumptions (e.g. conservative input parameters in the hot channel calculations), as well. In case of hot channel uncertainty analysis, the relevant input parameters (k x, mass flow, inlet temperature of the coolant, pin average burnup, initial gap size, selection of power history influencing the gap conductance value) of hot channel calculations and the DNBR limit are varied considering the respective uncertainties. An uncertainty analysis methodology was elaborated combining the response surface method with the one sided tolerance limit method of Wilks. The results of deterministic and uncertainty hot channel calculations are compared regarding to the number of failed fuel rods, max. temperature of the clad surface and max. temperature of the fuel (Authors)

Detorsion night-time bracing for the treatment of early onset idiopathic scoliosis.

Science.gov (United States)

Moreau, S; Lonjon, G; Mazda, K; Ilharreborde, B

2014-12-01

Management for early onset scoliosis has recently changed, with the development of new surgical procedures. However, multiple surgeries are often required and high complication rates are still reported. Conservative management remains an alternative, serial casting achieving excellent results in young children. Better compliance and improvement over natural history have been reported with night-time bracing in adolescent idiopathic scoliosis (AIS), but this treatment has never been reported in early onset idiopathic scoliosis (EIOS). All patients treated for progressive EOIS by detorsion night-time bracing (DNB), and meeting the Scoliosis Research Society (SRS) criteria for brace studies were reviewed. Recommendations were given to wear the DNB 8h/night and no restriction was given regarding sports activities. Radiological parameters were compared between referral and latest follow-up. Based on the SRS criteria defined for AIS, a similar classification was used as follows to analyze the course of the curves: success group: patients with a progression of 5° or less; unsuccess group (progression or failure): patients with a progression>5°, patients with curves exceeding 45° at maturity, or who have had recommendation for/undergone surgery, or patients who changed orthopaedic treatment, or who were lost to follow-up. Thirty-three patients were included (21 girls and 12 boys), with a median Cobb angle of 31° (Q1-Q3: 22-40). Age at brace initiation averaged 50months (Q1-Q3: 25-60). Median follow-up was 102-months (Q1-Q3: 63-125). Fifteen patients (45.5%) had reached skeletal maturity at last follow-up. The success rate was 67% (22 patients), with a median Cobb angle reduction of 15° (P<0.001). Four patients stopped DNB due to an important regression. Eleven patients were in the unsuccessful group (33%). Only one had surgery. All patients remained balanced in the frontal plane and normokyphotic. Initial curve magnitude and age at brace initiation appeared to be

Evolution of central banking? De Nederlandsche Bank 1814-1852

NARCIS (Netherlands)

Uittenbogaard, R.A.

2014-01-01

Nowadays the role of central bank is unquestioned and nearly ubiquitous. But was this always the case? This thesis analyses how De Nederlandsche Bank (DNB) developed into a central bank during the first four decades of its existence. Its establishment in 1814 was the result of a combination of both

78 FR 33445 - Office of Small Credit Unions (OSCUI) Grant Program Access For Credit Unions

Science.gov (United States)

2013-06-04

... hardware and software necessary to convert to computerized operations. The maximum award amount for this...,000 per credit union for financial education projects that improve financial capability in the... found on D&B's Web site at http://fedgov.dnb.com/webform or by calling D&B, toll-free, at 1-866-705-5711...

Critical heat flux experiments for high conversion light water reactor, (3)

International Nuclear Information System (INIS)

Iwamura, Takamichi; Okubo, Tsutomu; Suemura, Takayuki; Hiraga, Fujio; Murao, Yoshio

1990-03-01

As a part of the thermal-hydraulic feasibility study of a high conversion light water reactor (HCLWR), critical heat flux (CHF) experiments were performed using triangular array rod bundles under steady-state and flow reduction transient conditions. The geometries of test sections were: rod outer diameter 9.5 mm, number of rods 4∼7, heated length 0.5∼1.0 m, and pitch to diameter ratio (P/D) 1.126∼1.2. The simulated fuel rod was a stainless steel tube and uniformly heated electrically with direct current. In the steady-state tests, pressures ranged: 1.0∼3.9 Mpa, mass velocities: 460∼4270 kg/s·m 2 , and exit qualities: 0.02∼0.35. In the transient tests, the times to CHF detection ranged from 0.5 to 25.4 s. The steady-state CHF's for the 4-rod test sections were higher than those for the 7-rod test sections with respect to the bundle averaged flow conditions. The measured CHF's increased with decreasing the heated length and decreased with decreasing the P/D. Based on the local flow conditions obtained with the subchannel analysis code COBRA-IV-I, KfK correlation agreed with the CHF data within 20 %, while WSC-2, EPRI-B and W, EPRI-Columbia and Kattor correlations failed to give satisfactory agreements. Under flow reduction rates less than 6 %/s, no significant difference in the onset conditions of DNB (departure from nucleate boiling) was recognized between the steady-state and transient conditions. At flow reduction rates higher than 6 %/s, on the other hand, the DNB occurred earlier than the DNB time predicted with the steady-state experiments. (author)

Method for measuring the public's appreciation and knowledge of bank notes

Science.gov (United States)

de Heij, Hans A. M.

2002-04-01

No matter how sophisticated a banknotes' security features are, they are only effective if the public uses them. Surveys conducted by the De Nederlandsche Bank (the Dutch central bank, hereinafter: DNB) in the period 1989-1999 have shown that: the more people like a banknote, the more they know about it, including its security features; there is a positive correlation between the appreciation of a banknote (beautiful or ugly) and the knowledge of its security features, its picture and text elements; hardly anybody from the general public knows more than 4 security features by heart, which is why the number of security features for the public should be confined to a maximum of 4; the average number of security features known to a Dutchman was about 1.7 in 1999; over the years, the awareness of banknote security features gradually increased from 1.03 to 1983 to 1.7 in 1999, as a result of new banknote design and information campaigns. In 1999, DNB conducted its last opinion poll on NLG-notes. After the introduction of the euro banknotes on 1 January 2002, a new era of measurements will start. It is DNB's intention to apply the same method for the euro notes as it is used to for the NLG-notes, as this will permit: A comparison of the results of surveys on Dutch banknotes with those of surveys on the new euro notes (NLG) x (EUR); a comparison between the results of similar surveys conducted in other euro countries: (EUR1)x(EUR2). Furthermore, it will enable third parties to compare their banknote model XXX with the euro: (XXX)x(EUR). This article deals with the survey and the results regarding the NLG- notes and is, moreover, intended as an invitation to use the survey method described.

Measurements with vertically viewing charge exchange analyzers during ion cyclotron range of frequencies heating in TFTR

International Nuclear Information System (INIS)

Kaita, R.; Hammett, G.W.; Gammel, G.; Goldston, R.J.; Medley, S.S.; Scott, S.D.; Young, K.M.

1988-01-01

The utility of charge exchange neutral particle analyzers for studying energetic ion distributions in high-temperature plasmas has been demonstrated in a variety of tokamak experiments. Power deposition profiles have been estimated in the Princeton large torus (PLT) from particle measurements as a function of energy and angle during heating in the ion cyclotron range of frequencies (ICRF) and extensive studies of this heating mode are planned for the upcoming operational period in the tokamak fusion test reactor (TFTR). Unlike the horizontally scanning analyzer on PLT, the TFTR system consists of vertical sightlines intersecting a poloidal cross section of the plasma. A bounce-averaged Fokker--Planck program, which includes a quasilinear operator to calculate ICRF-generated energetic ions, is used to simulate the charge exchange flux expected during fundamental hydrogen heating. These sightlines also cross the trajectory of a diagnostic neutral beam (DNB), and it may be possible to observe the fast ion tail during 3 He minority heating, if the DNB is operated in helium for double charge exchange neutralization

Core heat transfer experiment for JRR-3 to be upgraded at 20 MWt, 2

International Nuclear Information System (INIS)

Sudo, Yukio; Miyata, Keiichi; Ikawa, Hiromasa; Ohgawara, Masami; Kaminaga, Masanori

1985-09-01

Experiments were carried out to investigate the condition of onset of nucleate boiling (ONB) and the departure from nucleate boiling (DNB) heat flux under forced convection in a vertical rectangular channel, both of which take important roles in the core thermal-hydraulic design of the upgraded JRR-3. This report presents the validity and applicability of the correlations proposed for ONB condition and DNB heat flux, based on the analysis of the experimental results. The upgraded JRR-3 is a low-pressure, low-temperature research reactor and the core heat generation is removed by two cooling modes, one is natural circulation under upflow up to 200 kW and the other is forced circulation under downflow up to 20 MW. Therefore, the difference in heat transfer characteristics between upflow and downflow were investigated in the experiments, which were carried out by using a heated channel properly simulating a subchannel of fuel element because the heat transfer characteristics are considered to be strongly dependent on the configuration of flow channel. (author)

Contribution to the multidimensional modelling of convective high pressure boiling flows for pressurised water reactors

International Nuclear Information System (INIS)

Gueguen, J.

2013-01-01

This study is a contribution to the modelling of multidimensional high pressure boiling flows relative to PWR. Numerical simulation of such two-phase flows is considered to be an interesting way for the DNB understanding. The first part of this study exposes a two-dimensional steady state two-phase flows model able to predict velocity and temperature profiles in tube. The mixture balanced equations are used with the eddy diffusivity concept to close the turbulent transport terms. The second part is devoted to the development of the model in the general two dimensional case. Contrary to the steady state model, this model is independent of experimental data and implies the use of an original local homogeneous relaxation model (HRM). The results obtained from the comparison with the data bank DEBORA reveals that in a mixture approach two sub models are sufficient to obtain a physical good description of turbulent boiling flows. Some limitations appear at conditions close to DNB conditions. The turbulent closures and the relaxation time in the HRM model have been clearly identified as the most important and sensitive parameters in the model. (author) [fr

LOFT fuel rod surface temperature measurement testing

International Nuclear Information System (INIS)

Eaton, A.M.; Tolman, E.L.; Solbrig, C.W.

1978-01-01

Testing of the LOFT fuel rod cladding surface thermocouples has been performed to evaluate how accurately the LOFT thermocouples measure the cladding surface temperature during a loss-of-coolant accident (LOCA) sequence and what effect, if any, the thermocouple would have on core performance. Extensive testing has been done to characterize the thermocouple design. Thermal cycling and corrosion testing of the thermocouple weld design have provided an expected lifetime of 6000 hours when exposed to reactor coolant conditions of 620 K and 15.9 MPa and to sixteen thermal cycles with an initial temperature of 480 K and peak temperatures ranging from 870 to 1200K. Departure from nucleate boiling (DNB) tests have indicated a DNB penalty (5 to 28% lower) during steady state operation and negligible effects during LOCA blowdown caused by the LOFT fuel rod surface thermocouple arrangement. Experience with the thermocouple design in Power Burst Facility (PBF) and LOFT nonnuclear blowdown testing has been quite satisfactory. Tests discussed here were conducted using both stainless steel and zircaloy-clad electrically heated rod in the LOFT Test Support Facility (LTSF) blowdown simulation loop

Experiment on transient heat transfer in closed narrow channel

International Nuclear Information System (INIS)

Ochiai, Masaaki

1985-01-01

Heat transfer coefficients and transient pressures in closed narrow channels were obtained experimentally, in order to assess the gap heat transfer models in the computer code WTRLGD which were devised to analyze the internal pressure behavior of waterlogged fuel rods. Gap widths of channels are 0.1--0.5mm to simulate the gap region of waterlogged fuel rods, and test fluids are water (7--89.2 0 C) and Freon-113 (9.2 0 C). The results show that the heater temperature and the pressure measured in the experiments without the DNB occurrence are simulated fairly well by the calculational model of WTRLGD where the heat transfer in a closed narrow channel is evaluated with one-dimensional transient thermal conduction equation and Jens and Lottes' correlation for nucleate boiling. Consequently, it is also suggested that the above equations are available for evaluation of heat flux from fuel to internal water of waterlogged fuel rods. The film boiling heat transfer coefficient was in the same order of that evaluated by Bromley's correlation and the DNB heat flux was smaller than that obtained in quasi-steady experiments with ordinary systems, although the experimental data for them were not enough. (author)

Single-phase CFD applicability for estimating fluid hot-spot locations in a 5 x 5 fuel rod bundle

International Nuclear Information System (INIS)

Ikeda, Kazuo; Makino, Yasushi; Hoshi, Masaya

2006-01-01

High-thermal performance PWR spacer grids require both of low pressure loss and high critical heat flux (CHF) properties. Therefore, a numerical study using computational fluid dynamics (CFD) was carried out to estimate pressure loss in strap and mixing vane structures. Moreover, a CFD simulation under single-phase flow condition was conducted for one specific condition in a water departure from nucleate boiling (DNB) test to examine the applicability of the CFD model for predicting the CHF rod position. Energy flux around the rod surface in a water DNB test is the sum of the intrinsic energy flux from a rod and the extrinsic energy flux from other rods, and increments of the enthalpy and decrements of flow velocity near the rod surface are assumed to affect CHF performance. CFD makes it possible to model the complicated flow field consisting of a spacer grid and a rod bundle and evaluate the local velocity and enthalpy distribution around the rod surface, which are assumed to determine the initial conditions for the two-phase structure. The results of this study indicate that single-phase CFD can play a significant role in designing PWR spacer grids for improved CHF performance

The optimal ecological factors and the denitrification populationof a denitrifying process for sulfate reducing bacteriainhibition

Science.gov (United States)

Li, Chunying

2018-02-01

SRB have great negative impacts on the oil production in Daqing Oil field. A continuous-flow anaerobic baffled reactors (ABR) are applied to investigate the feasibility and optimal ecological factors for the inhibition of SRB by denitrifying bacteria (DNB). The results showed that the SO42- to NO3- concentration ratio (SO42-/NO3-) are the most important ecological factor. The input of NO3- and lower COD can enhance the inhibition of S2-production effectively. The effective time of sulfate reduction is 6 h. Complete inhibition of SRB is obtained when the influent COD concentration is 600 mg/L, the SO42-/NO3- is 1/1 (600 mg/L for each), N is added simultaneously in the 2# and the 5# ABR chambers. By extracting the total DNA of wastewater from the effective chamber, 16SrDNA clones of a bacterium had been constructed. It is showed that the Proteobacteria accounted for eighty- four percent of the total clones. The dominant species was the Neisseria. Sixteen percent of the total clones were the Bacilli of Frimicutes. It indicated that DNB was effective and feasible for SRB inhibition.

Ignition of Ionic Liquids. Volume 2

Science.gov (United States)

2010-09-01

TOFMS time-of-flight-mass-spectrometry TS transition state VUV vacuum ultraviolet ZPE zero-point energy Approved for public...energies ( ZPEs ) were scaled by a factor of 0.9613 and 0.9804, respectively, and when necessary intrinsic reaction coordinate (IRC) calculations were...oscillations in the PE reflect the vibration of the DNB molecule, including ZPE . The trajectory shows three dissociation steps, eliminating NO2 followed

Assembly process of the ITER neutral beam injectors

Energy Technology Data Exchange (ETDEWEB)

Graceffa, J., E-mail: joseph.graceffa@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul lez Durance (France); Boilson, D.; Hemsworth, R.; Petrov, V.; Schunke, B.; Urbani, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul lez Durance (France); Pilard, V. [Fusion for Energy, C/ Josep Pla, n°2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

2013-10-15

The ITER neutral beam (NB) injectors are used for heating and diagnostics operations. There are 4 injectors in total, 3 heating neutral beam injectors (HNBs) and one diagnostic neutral beam injector (DNB). Two HNBs and the DNB will start injection into ITER during the hydrogen/helium phase of ITER operations. A third HNB is considered as an upgrade to the ITER heating systems, and the impact of the later installation and use of that injector have to be taken into account when considering the installation and assembly of the whole NB system. It is assumed that if a third HNB is to be installed, it will be installed before the nuclear phase of the ITER project. The total weight of one injector is around 1200 t and it is composed of 18 main components and 36 sets of shielding plates. The overall dimensions are length 20 m, height 10 m and width 5 m. Assembly of the first two HNBs and the DNB will start before the first plasma is produced in ITER, but as the time required to assemble one injector is estimated at around 1.5 year, the assembly will be divided into 2 steps, one prior to first plasma, and the second during the machine second assembly phase. To comply with this challenging schedule the assembly sequence has been defined to allow assembly of three first injectors in parallel. Due to the similar design between the DNB and HNBs it has been decided to use the same tools, which will be designed to accommodate the differences between the two sets of components. This reduces the global cost of the assembly and the overall assembly time for the injector system. The alignment and positioning of the injectors is a major consideration for the injector assembly as the alignment of the beamline components and the beam source are critical if good injector performance is to be achieved. The theoretical axes of the beams are defined relative to the duct liners which are installed in the NB ports. The concept adopted to achieve the required alignment accuracy is to use the

Radical pair formation in γ-irradiated 2-methyltetrahydrofuran rigid solutions of polynitrobenzenes

International Nuclear Information System (INIS)

Konishi, S.; Hoshino, M.; Imamura, M.

1981-01-01

The γ-irradiated MTHF (2-methyltetrahydrofuran) rigid solutions of mDNB (m-dinitrobenzene) and sTNB (s-trinitrobenzene) showed at 77 K ESR spectra characteristic of triplet species in addition to the spectra of doublet species, whereas no triplet ESR spectra were observed for the mononitrobenzene and o- and p-di-nitrobenzene solutions. The distances of the unpaired spins evaluated from the observed fine structure constants by using a point-dipole approximation are 4.3 and 4.6 A for the mDNB solution and 3.9 and 4.7 A for the sTNB solution. The detection of only the solute anion radicals by the optical absorption spectra of the irradiated solutions and the difference of the rate of formation for the triplet species and the solute anion strongly suggest that the triplet species are ascribed to the solute anion-solvent radical pairs. Such radical pairs are most likely to be formed through the migration of a MTHF cation radical, i.e., so-called hole migration, to a specific site between the two nitro groups on the meta positions of a solute anion followed by the production of a stable solvent radical, which is paired with the solute anion

Thyc, a 3D thermal-hydraulic code for rod bundles. Recent developments and validation tests

International Nuclear Information System (INIS)

Caremoli, C.; Rascle, P.; Aubry, S.; Olive, J.

1993-09-01

PWR or LMFBR cores or fuel assemblies, PWR steam generators, condensers, tubular heat exchangers, are basic components of a nuclear power plant involving two-phase flows in tube or rod bundles. A deep knowledge of the detailed flow patterns on the shell side is necessary to evaluate DNB margins in reactor cores, singularity effects (grids, wire spacers, support plates, baffles), corrosion on steam generator tube sheet, bypass effects and vibration risks. For that purpose, Electricite de France has developed, since 1986, a general purpose code named THYC (Thermal HYdraulic Code) designed to study three-dimensional single and two phase flows in rod or tube bundles (pressurized water reactor cores, steam generators, condensers, heat exchangers). It considers the three-dimensional domain to contain two kinds of components: fluid and solids. The THYC model is obtained by space-time averaging of the instantaneous equations (mass, momentum and energy) of each phase over control volumes including fluid and solids. This paper briefly presents the physical model and the numerical method used in THYC. Then, validation tests (comparison with experiments) and applications (coupling with three-dimensional neutronics code and DNB predictions) are presented. They emphasize the last developments and new capabilities of the code. (authors). 10 figs., 3 tabs., 21 refs

Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1976

Energy Technology Data Exchange (ETDEWEB)

Ferguson, J. B. [ed.

1977-04-01

Light water reactor safety research performed October through December 1976 is discussed. An analysis to determine the effect of emergency core coolant (ECC) injection location and pump speed on system response characteristics was performed. An analysis to evaluate the capability of commonly used critical heat flux (CHF) correlations to calculate the time of the first CHF in the Semiscale core during a loss-of-coolant experiment (LOCE) was performed. A test program and study to determine the effect thermocouples mounted on the outside fuel rod surfaces would have on the departure from nucleate boiling (DNB) phenomena in the LOFT core during steady state operation were completed. A correlation for use in predicting DNB heat fluxes in the LOFT core was developed. Tests of an experimental transit time flowmeter were completed. A nuclear test was performed to obtain fuel rod behavior data from four PWR-type rods during film boiling operation representative of PWR conditions. Preliminary results from the postirradiation examination of Test IE-1 fuel rods are given. Results of Irradiation Effects Tests IE-2 and IE-3 are given. Gap Conductance Test GC 2-1 was performed to evaluate the effects of fuel density, initial gap width, and fill gas composition on the pellet-cladding gap conductance.

Selective Hydrogenation of m-Dinitrobenzene to m-Nitroaniline over Ru-SnOx/Al2O3 Catalyst

Directory of Open Access Journals (Sweden)

Haiyang Cheng

2014-07-01

Full Text Available Series catalysts of Ru-SnOx/Al2O3 with varying SnOx loading of 0–3 wt% were prepared, and their catalytic activity and selectivity have been discussed and compared for the selective hydrogenation of m-dinitrobenzene (m-DNB to m-nitroaniline (m-NAN. The Ru-SnOx/Al2O3 catalysts were characterized by X-ray powder diffraction (XRD, X-ray photoelectron spectroscopy (XPS, transmission electron microscopy (TEM and hydrogen temperature-programmed reduction (H2-TPR and desorption (H2-TPD. Under the modification of SnOx, the reaction activity increased obviously, and the best selectivity to m-NAN reached above 97% at the complete conversion of m-DNB. With the increasing of the SnOx loading, the amount of active hydrogen adsorption on the surface of the catalyst increased according to the H2-TPD analysis, and the electron transferred from Ru to SnOx species, as determined by XPS, inducing an electron-deficient Ru, which is a benefit for the absorption of the nitro group. Therefore, the reaction rate and product selectivity were greatly enhanced. Moreover, the Ru-SnOx/Al2O3 catalyst presented high stability: it could be recycled four times without any loss in activity and selectivity.

Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1976

International Nuclear Information System (INIS)

Ferguson, J.B.

1977-04-01

Light water reactor safety research performed October through December 1976 is discussed. An analysis to determine the effect of emergency core coolant (ECC) injection location and pump speed on system response characteristics was performed. An analysis to evaluate the capability of commonly used critical heat flux (CHF) correlations to calculate the time of the first CHF in the Semiscale core during a loss-of-coolant experiment (LOCE) was performed. A test program and study to determine the effect thermocouples mounted on the outside fuel rod surfaces would have on the departure from nucleate boiling (DNB) phenomena in the LOFT core during steady state operation were completed. A correlation for use in predicting DNB heat fluxes in the LOFT core was developed. Tests of an experimental transit time flowmeter were completed. A nuclear test was performed to obtain fuel rod behavior data from four PWR-type rods during film boiling operation representative of PWR conditions. Preliminary results from the postirradiation examination of Test IE-1 fuel rods are given. Results of Irradiation Effects Tests IE-2 and IE-3 are given. Gap Conductance Test GC 2-1 was performed to evaluate the effects of fuel density, initial gap width, and fill gas composition on the pellet-cladding gap conductance

Power-cooling mismatch test series. Test PCM-2A; test results report

International Nuclear Information System (INIS)

Cawood, G.W.; Holman, G.W.; Martinson, Z.R.; Legrand, B.L.

1976-09-01

The report describes the results of an in-pile experimental investigation of pre- and postcritical heat flux (CHF) behavior of a single 36-inch-long, pressurized water reactor (PWR) type, UO 2 -fueled, zircaloy-clad fuel rod. The nominal coolant conditions for pressure and temperature were representative of those found in a commercial PWR. Nine separate departures from nucleate boiling (DNB) cycles were performed by either of two different methods: (a) decreasing the coolant flow rate while the fuel rod power was held constant, or (b) increasing the fuel rod power while the coolant flow rate was held constant. DNB was obtained during eight of the nine cycles performed. For the final flow reduction, the mass flux was decreased to 6.1 x 10 5 lb/hr-ft 2 at a constant peak linear heat generation rate of 17.8 kW/ft. The fuel rod was allowed to remain in film boiling for about 210 seconds during this final flow reduction. The fuel rod remained intact during the test. Results of on-line measurements of the fuel rod behavior are presented together with discussion of instrument performance. A comparison of the data with Fuel Rod Analysis Program-Transient 2 (FRAP-T2) computer code calculations is included

Statistical core design

International Nuclear Information System (INIS)

Oelkers, E.; Heller, A.S.; Farnsworth, D.A.; Kearfott, K.J.

1978-01-01

The report describes the statistical analysis of DNBR thermal-hydraulic margin of a 3800 MWt, 205-FA core under design overpower conditions. The analysis used LYNX-generated data at predetermined values of the input variables whose uncertainties were to be statistically combined. LYNX data were used to construct an efficient response surface model in the region of interest; the statistical analysis was accomplished through the evaluation of core reliability; utilizing propagation of the uncertainty distributions of the inputs. The response surface model was implemented in both the analytical error propagation and Monte Carlo Techniques. The basic structural units relating to the acceptance criteria are fuel pins. Therefore, the statistical population of pins with minimum DNBR values smaller than specified values is determined. The specified values are designated relative to the most probable and maximum design DNBR values on the power limiting pin used in present design analysis, so that gains over the present design criteria could be assessed for specified probabilistic acceptance criteria. The results are equivalent to gains ranging from 1.2 to 4.8 percent of rated power dependent on the acceptance criterion. The corresponding acceptance criteria range from 95 percent confidence that no pin will be in DNB to 99.9 percent of the pins, which are expected to avoid DNB

Where Angels Fear to Trade : The Role of Religion in Household Finance

OpenAIRE

Renneboog, L.D.R.; Spaenjers, C.

2009-01-01

Although the relationship between religion and economic development on the macro-level has been investigated, it is less clear how religious background influences economic attitudes and financial decision-making on the level of the individual or household, the micro-level. We use panel data from the extensive DNB Household Survey, covering the period from 1995 to 2008, to investigate whether – and through which channel – religious denomination affects household finance in the Netherlands. We ...

Die DDC auf neuen Wegen - verbale Sucheinstiege für klassifikatorisch erschlossene Titel

OpenAIRE

Christiane Maibach

2014-01-01

Die Dewey-Dezimalklassifikation (DDC) ist ein mächtiges Instrument der klassifikatorischen Inhaltserschließung. Immer mehr Bibliotheken im In- und Ausland erkennen den Nutzen der international am weitesten verbreiteten Universalklassifikation. Die Deutsche Nationalbibliothek (DNB) setzt die DDC seit 2006 für die inhaltliche Erschließung der Bibliografiereihen A, B und H ein. Seitdem wurden rund 800.000 Publikationen mit DDC-Notationen versehen. Der Nutzen von Klassifikationen ist unter Biblio...

Virginia Power thermal-hydraulics methods

International Nuclear Information System (INIS)

Anderson, R.C.; Basehore, K.L.; Harrell, J.R.

1987-01-01

Virginia Power's nuclear safety analysis group is responsible for the safety analysis of reload cores for the Surry and North Anna power stations, including the area of core thermal-hydraulics. Postulated accidents are evaluated for potential departure from nucleate boiling violations. In support of these tasks, Virginia Power has employed the COBRA code and the W-3 and WRB-1 DNB correlations. A statistical DNBR methodology has also been developed. The code, correlations and statistical methodology are discussed

Modelling of subcooled boiling and DNB-type boiling crisis in forced convection

International Nuclear Information System (INIS)

Bricard, Patrick

1995-01-01

This research thesis aims at being a contribution to the modelling of two phenomena occurring during a forced convection: the axial evolution of the vacuum rate, and the boiling crisis. Thus, the first part of this thesis addresses the prediction of the vacuum rate, and reports the development of a modelling of under-saturated convection in forced convection. The author reports the development and assessment of two-fluid one-dimensional model, the development of a finer analysis based on an averaging of local equations of right cross-sections in different areas. The second part of this thesis addresses the prediction of initiation of a boiling crisis. The author presents generalities and motivations for this study, reports a bibliographical study and a detailed analysis of mechanistic models present in this literature. A mechanism of boiling crisis is retained, and then further developed in a numerical modelling which is used to assess some underlying hypotheses [fr

Anti-D Antibodies in Pregnant D Variant Antigen Carriers Initially Typed as RhD.

Science.gov (United States)

Lukacevic Krstic, Jelena; Dajak, Slavica; Bingulac-Popovic, Jasna; Dogic, Vesna; Mratinovic-Mikulandra, Jela

2016-11-01

To evaluate the incidence, the consequences, and the prevention strategy of anti-D alloimmunizations of D variant carriers in the obstetric population of Split-Dalmatia County, Croatia. RhD immunization events were evaluated retrospectively for the period between 1993 and 2012. Women were tested for RhD antigen and irregular antibodies. Those with anti-D antibody who were not serologically D- were genotyped for RHD. They were evaluated for their obstetric and transfusion history and their titer of anti-D. The neonates were evaluated for RhD status, direct antiglobulin test (DAT), hemoglobin and bilirubin levels, transfusion therapy as well as phototherapy and outcome. Out of 104,884 live births 102,982 women were tested for RhD antigen. Anti-D immunization occurred in 184 women which accounts for 0.9% of individuals at risk of anti-D formation. 181 cases occurred in women serologically typed as D-. Three women were partial D carriers (DVa n = 2, DNB n = 1), initially typed RhD+, and recognized as D variant carriers after the immunization occurred. Anti-D titer varied from 1:1 to 1:16. Six children were RhD+, four had positive DAT, and two underwent phototherapy. Anti-D immunization occurred in pregnant partial D carriers (DVa, DNB). RhD+ children had serologic markers of hemolytic disease of the fetus and newborn (HDFN), with no cases of severe HDFN.

Uncertainties assessment for safety margins evaluation in MTR reactors core thermal-hydraulic design

International Nuclear Information System (INIS)

Gimenez, M.; Schlamp, M.; Vertullo, A.

2002-01-01

This report contains a bibliographic review and a critical analysis of different methodologies used for uncertainty evaluation in research reactors core safety related parameters. Different parameters where uncertainties are considered are also presented and discussed, as well as their intrinsic nature regarding the way their uncertainty combination must be done. Finally a combined statistical method with direct propagation of uncertainties and a set of basic parameters as wall and DNB temperatures, CHF, PRD and their respective ratios where uncertainties should be considered is proposed. (author)

External Shocks and Macroeconomic Policy: Simulations with EUROMON

OpenAIRE

M. Demertzis; L. de Haan

2001-01-01

We carry out a number of policy simulations with DNB's multicountry model, EUROMON. With these simulations we aim to analyse the effectiveness of monetary and fiscal expansion in light of the current global downturn in the US and the Euro area. We thus run two types of simulations in which we examine first, the real and nominal effects of the interest rate reductions implemented over the past year (2002) in both the US and the EU and second compare, the effects of similar macro policies appli...

Public feedback for better banknote design

Science.gov (United States)

de Heij, Hans A. M.

2006-02-01

The euro banknotes recently celebrated their 4th year anniversary (2002 - 2006)! DNB has monitored the Dutch public's acceptance of the euro banknotes through surveys in 2002, 2003 and 2005, contrasting its findings with those for the former guilder notes as well as the results of other recent consumer surveys conducted by the US Treasury (2002), Bank of Canada (2003) and several others. A unique time line is presented of the public's knowledge of security features in the Netherlands over the years 1983 - 2005.

Analysis of Rod Withdrawal at Power (RWAP) Accident using ATHLET Mod 2.2 Cycle A and RELAP5/mod 3.3 Codes

International Nuclear Information System (INIS)

Bencik, V.; Cavlina, N.; Grgic, D.

2012-01-01

The system code ATHLET is being developed at Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS) in Germany. In 1996, the NPP Krsko (NEK) input deck for ATHLET Mod 1.1 Cycle C has been developed at Faculty of Electrical Engineering (FER), University of Zagreb. The input deck was tested by analyzing the realistic plant event 'Main Steam Isolation Valve Closure' and the results were assessed against the measured data. The input deck was established before plant modernization that took place in 2000 and included the power uprate and SG replacement. The released ATHLET version (Mod 2.2 Cycle A) is now being available at FER Zagreb. Accordingly, the NEK input deck for ATHLET Mod 2.2 Cycle A has been developed. A completely new input deck has been created taking into account the large number of changes due to power uprate and SG replacement as well as taking advantage of developmental work on NEK data base performed at FER. The new NEK input deck for ATHLET code has been tested by analyzing the Rod Withdrawal Power (RWAP) accident and the results were assessed against the analysis performed by RELAP5/mod 3.3 code. The RWAP accident can be either Departure from Nucleate Boiling (DNB) ratio or overpower limiting accident depending on initial power and reactivity insertion rate. Since the automatic rod control system is assumed unavailable, the only negative reactivity is due to Doppler and moderator feedback. Consequently, the nuclear power and the transferred heat in the steam generators (SGs) increase. Since the steam flow to the turbine and the extracted power from the SGs remain constant, the SG secondary pressure and the temperatures on the primary side increase. Unless terminated by manual or automatic action, the power mismatch between primary and secondary side and the resultant coolant temperature rise could eventually result in DNB ratio and/or fuel centreline melt. In order to avoid core damage, the reactor protection system is designed to automatically

An investigation of transition boiling mechanisms of subcooled water under forced convective conditions

Energy Technology Data Exchange (ETDEWEB)

Kwang-Won, Lee; Sang-Yong, Lee

1995-09-01

A mechanistic model for forced convective transition boiling has been developed to investigate transition boiling mechanisms and to predict transition boiling heat flux realistically. This model is based on a postulated multi-stage boiling process occurring during the passage time of the elongated vapor blanket specified at a critical heat flux (CHF) condition. Between the departure from nucleate boiling (DNB) and the departure from film boiling (DFB) points, the boiling heat transfer is established through three boiling stages, namely, the macrolayer evaporation and dryout governed by nucleate boiling in a thin liquid film and the unstable film boiling characterized by the frequent touches of the interface and the heated wall. The total heat transfer rates after the DNB is weighted by the time fractions of each stage, which are defined as the ratio of each stage duration to the vapor blanket passage time. The model predictions are compared with some available experimental transition boiling data. The parametric effects of pressure, mass flux, inlet subcooling on the transition boiling heat transfer are also investigated. From these comparisons, it can be seen that this model can identify the crucial mechanisms of forced convective transition boiling, and that the transition boiling heat fluxes including the maximum heat flux and the minimum film boiling heat flux are well predicted at low qualities/high pressures near 10 bar. In future, this model will be improved in the unstable film boiling stage and generalized for high quality and low pressure situations.

Lizenzierungsservice Vergriffene Werke (VW-LiS – ein neuer Dienst der Deutschen Nationalbibliothek

Directory of Open Access Journals (Sweden)

Reinhard Altenhöner

2015-12-01

Full Text Available Vor allem die bestehende urheberrechtliche Situation verhindert die Digitalisierung von Beständen des 20. Jahrhunderts. Basierend auf einem EU-Memorandum hat der deutsche Gesetzgeber durch eine Neuregelung für die Materialart Buch eine Möglichkeit geschaffen: Nach dem Erwerb einer entsprechenden Lizenz bewegt sich die digitalisierende Einrichtung bei der öffentlichen Bereitstellung digitalisierter Bücher in einem rechtssicheren Raum. In Kooperation mit dem Deutschen Bibliotheksverband (dbv, der VG Wort und VG Bild-Kunst sowie dem Deutschen Patent- und Markenamt (DPMA hat die Deutsche Nationalbibliothek (DNB bis Juli 2015 einen Dienst entwickelt, der die Ermittlung vergriffener Bücher sowie ihre Lizenzierung im Rahmen der gesetzlichen Vorgabe wesentlich erleichtert. Dieser Dienst wird hier vorgestellt und eingeordnet. The existing copyright situation prevents the digitization of collections of the 20th century. Based on a European memorandum, the German legislature has created a new way for books: After the acquisition of the appropriate licence, the digitization becomes possible; the public provision of digitized books takes place in a legally safeguarded manner. In cooperation with the German Library Association (dbv, the copyright collecting agencies Wort and VG Bild-Kunst and the DPMA, the German National Library (DNB has until July 2015 developed a service which significantly facilitates the identification of out-of-commerce-books as well as their licensing under the new regulation. The new service will be presented and explained.

Utilization of a statistical procedure for DNBR calculation and in the survey of reactor protection limits

International Nuclear Information System (INIS)

Pontedeiro, A.C.; Camargo, C.T.M.; Galetti, M.R. da Silva.

1987-01-01

A new procedure is applied to Angra 1 NPP, which is related to DNBR calculations, considering the design parameters statistically: Improved Thermal Design Procedure (ITDP). The ITDP application leads to the determination of uncertainties in the input parameters, the sensitivity factors on DNBR. The DNBR limit and new reactor protection limits. This was done to Angra 1 with the subchannel code COBRA-IIIP. The analysis of limiting accident in terms of DNB confirmed a gain in DNBR margin, and greater operation flexibility of the plant, decreasing unnecessary trips of the reactor. (author) [pt

Summary report on the aerobic degradation of diesel fuel and the degradation of toluene under aerobic, denitrifying and sulfate reducing conditions

International Nuclear Information System (INIS)

Coyne, P.; Smith, G.

1995-01-01

This report contains a number of studies that were performed to better understand the technology of the biodegradation of petroleum hydrocarbons. Topics of investigation include the following: diesel fuel degradation by Rhodococcus erythropolis; BTEX degradation by soil isolates; aerobic degradation of diesel fuel-respirometry; aerobic degradation of diesel fuel-shake culture; aerobic toluene degradation by A3; effect of HEPES, B1, and myo-inositol addition on the growth of A3; aerobic and anaerobic toluene degradation by contaminated soils; denitrifying bacteria MPNs; sulfate-reducing bacteria MPNs; and aerobic, DNB and SRB enrichments

Verification of bubble tracking method and DNS examinations of single- and two-phase turbulent channel flows

Energy Technology Data Exchange (ETDEWEB)

Tryggvason, Gretar [Univ. of Notre Dame, IN (United States); Bolotnov, Igor [North Carolina State Univ., Raleigh, NC (United States); Fang, Jun [North Carolina State Univ., Raleigh, NC (United States); Lu, Jiacai [Univ. of Notre Dame, IN (United States)

2017-03-30

Direct numerical simulation (DNS) has been regarded as a reliable data source for the development and validation of turbulence models along with experiments. The realization of DNS usually involves a very fine mesh that should be able to resolve all relevant turbulence scales down to Kolmogorov scale [1]. As the most computationally expensive approach compared to other CFD techniques, DNS applications used to be limited to flow studies at very low Reynolds numbers. Thanks to the tremendous growth of computing power over the past decades, the simulation capability of DNS has now started overlapping with some of the most challenging engineering problems. One of those examples in nuclear engineering is the turbulent coolant flow inside reactor cores. Coupled with interface tracking methods (ITM), the simulation capability of DNS can be extended to more complicated two-phase flow regimes. Departure from nucleate boiling (DNB) is the limiting critical heat flux phenomena for the majority of accidents that are postulated to occur in pressurized water reactors (PWR) [2]. As one of the major modeling and simulation (M&S) challenges pursued by CASL, the prediction capability is being developed for the onset of DNB utilizing multiphase-CFD (M-CFD) approach. DNS (coupled with ITM) can be employed to provide closure law information for the multiphase flow modeling at CFD scale. In the presented work, research groups at NCSU and UND will focus on applying different ITM to different geometries. Higher void fraction flow analysis at reactor prototypical conditions will be performed, and novel analysis methods will be developed, implemented and verified for the challenging flow conditions.

A Generalized turbulent dispersion model for bubbly flow numerical simulation in NEPTUNE-CFD

Energy Technology Data Exchange (ETDEWEB)

Laviéville, Jérôme, E-mail: Jerome-marcel.lavieville@edf.fr; Mérigoux, Nicolas, E-mail: nicolas.merigoux@edf.fr; Guingo, Mathieu, E-mail: mathieu.guingo@edf.fr; Baudry, Cyril, E-mail: Cyril.baudry@edf.fr; Mimouni, Stéphane, E-mail: stephane.mimouni@edf.fr

2017-02-15

The NEPTUNE-CFD code, based upon an Eulerian multi-fluid model, is developed within the framework of the NEPTUNE project, financially supported by EDF (Electricité de France), CEA (Commissariat à l’Energie Atomique et aux Energies Alternatives), IRSN (Institut de Radioprotection et de Sûreté Nucléaire) and AREVA-NP. NEPTUNE-CFD is mainly focused on Nuclear Safety applications involving two-phase water-steam flows, like two-phase Pressurized Shock (PTS) and Departure from Nucleate Boiling (DNB). Many of these applications involve bubbly flows, particularly, for application to flows in PWR fuel assemblies, including studies related to DNB. Considering a very usual model for interfacial forces acting on bubbles, including drag, virtual mass and lift forces, the turbulent dispersion force is often added to moderate the lift effect in orthogonal directions to the main flow and get the right dispersion shape. This paper presents a formal derivation of this force, considering on the one hand, the fluctuating part of drag and virtual mass, and on the other hand, Turbulent Pressure derivation obtained by comparison between Lagrangian and Eulerian description of bubbles motion. An extension of the Tchen’s theory is used to express the turbulent kinetic energy of bubbles and the two-fluid turbulent covariance tensor in terms of liquid turbulent velocities and time scale. The model obtained by this way, called Generalized Turbulent Dispersion Model (GTD), does not require any user parameter. The model is validated against Liu & Bankoff air-water experiment, Arizona State University (ASU) experiment, DEBORA experiment and Texas A&M University (TAMU) boiling flow experiments.

Reliability analysis of PWR thermohydraulic design by the Monte Carlo method

International Nuclear Information System (INIS)

Silva Junior, H.C. da; Berthoud, J.S.; Carajilescov, P.

1977-01-01

The operating power level of a PWR is limited by the occurence of DNB. Without affecting the safety and performance of the reactor, it is possible to admit failure of a certain number of core channels. The thermohydraulic design, however, is affect by a great number of uncertainties of deterministic or statistical nature. In the present work, the Monte Carlo method is applied to yield the probability that a number F of channels submitted to boiling crises will not exceed a number F* previously given. This probability is obtained as function of the reactor power level. (Author) [pt

Manufacturing of large size RF based -ve ion source with 8 drivers-challenges and learnings

International Nuclear Information System (INIS)

Joshi, Jaydeep; Patel, Hitesh; Singh, Mahendrajit; Bandyopadhyay, Mainak; Chakraborty, Arun

2017-01-01

Radio Frequency (RF) Ion Source for ITER Diagnostic Neutral Beam (DNB) system, is an 8 driver based ion source, where the desired plasma density is produced by inductive coupling of RF power. The present paper describes the experience of developing a manufacturing design to meet the above mentioned requirements, feasibility assessment, prototyping carried out, parallel experiments in support of manufacturing and realization of sub-components along with their quality inspections activities performed. Additionally, paper also presents to the observations in terms of deviations and non-conformities encountered, as a part of learning for the future components

Powerloads on the front end components and the duct of the heating and diagnostic neutral beam lines at ITER

Energy Technology Data Exchange (ETDEWEB)

Singh, M. J.; Boilson, D.; Hemsworth, R. S.; Geli, F.; Graceffa, J.; Urbani, M.; Schunke, B.; Chareyre, J. [ITER Organisation, 13607 St. Paul-Lez-Durance Cedex (France); Dlougach, E.; Krylov, A. [RRC Kurchatov institute, 1, Kurchatov Sq, Moscow, 123182 (Russian Federation)

2015-04-08

The heating and current drive beam lines (HNB) at ITER are expected to deliver ∼16.7 MW power per beam line for H beams at 870 keV and D beams at 1 MeV during the H-He and the DD/DT phases of ITER operation respectively. On the other hand the diagnostic neutral beam (DNB) line shall deliver ∼2 MW power for H beams at 100 keV during both the phases. The path lengths over which the beams from the HNB and DNB beam lines need to be transported are 25.6 m and 20.7 m respectively. The transport of the beams over these path lengths results in beam losses, mainly by the direct interception of the beam with the beam line components and reionisation. The lost power is deposited on the surfaces of the various components of the beam line. In order to ensure the survival of these components over the operational life time of ITER, it is important to determine to the best possible extent the operational power loads and power densities on the various surfaces which are impacted by the beam in one way or the other during its transport. The main factors contributing to these are the divergence of the beamlets and the halo fraction in the beam, the beam aiming, the horizontal and vertical misalignment of the beam, and the gas profile along the beam path, which determines the re-ionisation loss, and the re-ionisation cross sections. The estimations have been made using a combination of the modified version of the Monte Carlo Gas Flow code (MCGF) and the BTR code. The MCGF is used to determine the gas profile in the beam line and takes into account the active gas feed into the ion source and neutraliser, the HNB-DNB cross over, the gas entering the beamline from the ITER machine, the additional gas atoms generated in the beam line due to impacting ions and the pumping speed of the cryopumps. The BTR code has been used to obtain the power loads and the power densities on the various surfaces of the front end components and the duct modules for different scenarios of ITER

Application of the SCANAIR code for VVER RIA conditions - Boron dilution accident

International Nuclear Information System (INIS)

Arffman, A.; Cazalis, B.

2010-01-01

This paper consists of two parts. In part A, RIA pulse tests conducted at the Russian BIGR reactor are being analysed at IRSN with SCANAIR V6 fuel performance code as a part of the code validation for VVER fuel. Recently a new version of the SCANAIR code was made available to VTT Technical Research Centre of Finland, and part B of the paper covers the introduction of the code version at VTT by a calculation of a hypothetical boron dilution accident in a VVER-440 power reactor. Concerning part A, it appears that the SCANAIR V6 version, including a BIGR/NSRR heat transfer model, validated by Japanese NSRR experiments, and a Norton viscoplastic clad mechanical behaviour, is able to simulate the rod thermal behaviour in BIGR tests. Concerning the clad mechanics, it has been seen that a pellet swelling model is able to simulate the average rod deformation. Nonetheless, the current clad creep model associated with the free volume equilibrium assumption is not suited to predict the maximum clad deformation and the possible post DNB rod failure because they do not simulate local balloons. Furthermore, it has been shown that the clad deformation is strongly dependent on transient gas transfer. Concerning part B, a boron dilution accident previously calculated with SCANAIR V2 was recalculated with SCANAIR V6. A limited amount of result parameters were compared with the results of VTT's neutronics code TRAB. Divergence problems encountered previously when reaching the DNB limit were not present anymore. Fuel and cladding temperatures produced by SCANAIR were in good agreement with those calculated with TRAB

Amphibole Fractional Crystallization and Delamination in Arc Roots: Implications for the `Missing' Nb Reservoir in the Earth

Science.gov (United States)

Galster, F.; Chatterjee, R. N.; Stockli, D. F.

2017-12-01

Most geologic processes should not fractionate Nb from Ta but Earth's major silicate reservoirs have subchondritic Nb/Ta values. Nb/Ta of >10000 basalts and basaltic andesites from different tectonic settings (GEOROC) cluster around 16, indistinguishable from upper mantle values. In contrast, Nb/Ta in more evolved arc volcanics have progressively lower values, reaching continental crust estimates, and correlate negatively with SiO2 (see figure) and positively with TiO2 and MgO. This global trend suggests that differentiation processes in magmatic arcs could explain bulk crustal Nb/Ta estimates. Understanding processes that govern fractionation of Nb from Ta in arcs can provide key insights on continental crust formation and help identify Earth's `missing' Nb reservoir. Ti-rich phases (rutile, titanite and ilmenite) have DNb/DTa values in the evolved liquid. Lack of correlation between Nb/Ta and K2O in global volcanic rocks implies that biotite plays a minor role in fractionating Nb from Ta during differentiation. Experimental petrology and evidence from exposed arc sections indicate that amphibole fractionation and delamination of island arc roots can explain the andesitic composition of bulk continental crust. Experimental studies have shown that amphibole Mg# correlate with DNb/DTa and amphibole could effectively fractionate Nb from Ta. Preliminary data from lower to middle crustal amphiboles from preserved arcs show sub- to super-chondritic Nb/Ta up to >60. This suggests that delamination of amphibole-rich cumulates can be a viable mechanism for the preferential removal of Nb from the continental crust. Future examination of Nb/Ta ratios in lower crustal amphiboles from various preserved arcs will provide improved constraints on the Nb-Ta paradox of the silicate Earth.

FY 1993 progress report on the ANS thermal-hydraulic test loop operation and results

International Nuclear Information System (INIS)

Siman-Tov, M.; Felde, D.K.; Farquharson, G.

1994-07-01

The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). Highly subcooled heavy-water coolant flows vertically upward at a very high mass flux of almost 27 MG/m 2 -s. In a parallel fuel plate configuration as in the ANSR, the flow is subject to a potential excursive static-flow instability that can very rapidly lead to flow starvation and departure from nucleate boiling (DNB) in the ''hot channel''. The current correlations and experimental data bases for flow excursion (FE) and critical heat flux (CHF) seldom evaluate the specific combination of ANSR operating parameters. The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 17 MW/m 2 , a mass flux range of 8 to 28 Mg/m 2 -s, an exit pressure range of 1.4 to 2.1 MPa, and an inlet temperature range of 40 to 50 C. FE experiments were also conducted using as ''soft'' a system as possible to secure a true FE phenomena (actual secondary burnout). True DNB experiments under similar conditions were also conducted. To the author's knowledge, no other FE data have been reported in the literature to date that dover such a combination of conditions of high mass flux, high heat flux, and moderately high pressure

Variación espacial, temporal y espectral de la contaminación lumínica y sus fuentes: Metodología y resultados

OpenAIRE

Sánchez de Miguel, Alejandro

2016-01-01

A partir de las imágenes nocturnas en banda pancromática obtenidas por los satélites de observación de la tierra (DMSP/OLS y SNPP/VIIRS/DNB) se ha estudiado la evolución temporal (a nivel global y regional) del gasto en alumbrado público. Para ello se ha diseñado un método de intercalibración de los datos de diferentes misiones que ha proporcionado, como primer resultado y usando los datos del satélite DMSP/OLS, la reconstrucción de la evolución del gasto en alumbrado público en España entre ...

Pseudo-cubic thin-plate type Spline method for analyzing experimental data

Energy Technology Data Exchange (ETDEWEB)

Crecy, F de

1994-12-31

A mathematical tool, using pseudo-cubic thin-plate type Spline, has been developed for analysis of experimental data points. The main purpose is to obtain, without any a priori given model, a mathematical predictor with related uncertainties, usable at any point in the multidimensional parameter space. The smoothing parameter is determined by a generalized cross validation method. The residual standard deviation obtained is significantly smaller than that of a least square regression. An example of use is given with critical heat flux data, showing a significant decrease of the conception criterion (minimum allowable value of the DNB ratio). (author) 4 figs., 1 tab., 7 refs.

Pseudo-cubic thin-plate type Spline method for analyzing experimental data

International Nuclear Information System (INIS)

Crecy, F. de.

1993-01-01

A mathematical tool, using pseudo-cubic thin-plate type Spline, has been developed for analysis of experimental data points. The main purpose is to obtain, without any a priori given model, a mathematical predictor with related uncertainties, usable at any point in the multidimensional parameter space. The smoothing parameter is determined by a generalized cross validation method. The residual standard deviation obtained is significantly smaller than that of a least square regression. An example of use is given with critical heat flux data, showing a significant decrease of the conception criterion (minimum allowable value of the DNB ratio). (author) 4 figs., 1 tab., 7 refs

Thermohydraulic characteristics analysis of natural convective cooling mode on the steady state condition of upgraded JRR-3 core, using COOLOD-N code

International Nuclear Information System (INIS)

Kaminaga, Masanori; Watanabe, Shukichi; Ando, Hiroei; Sudo, Yukio; Ikawa, Hiromasa.

1987-03-01

This report describes the results of the steady state thermohydraulic analysis of upgraded JRR-3 core under natural convective cooling mode, using COOLOD-N code. In the code, function to calculate flow-rate under natural convective cooling mode, and a heat transfer package have been newly added to the COOLOD code which has been developed in JAERI. And this report describes outline of the COOLOD-N code. The results of analysis show that the thermohydraulics of upgraded JRR-3 core, under natural convective cooling mode have enough margine to ONB temperature, DNB heat flux and occurance of blisters in fuel meats, which are design criterion of upgraded JRR-3. (author)

Heat and mass transfer and hydrodynamics in two-phase flows in nuclear power plants

International Nuclear Information System (INIS)

Styrikovich, M.A.; Polonskii, V.S.; Tsiklauri, G.V.

1986-01-01

This book examines nuclear power plant equipment from the point of view of heat and mass transfer and the behavior of impurities contained in water and in steam, with reference to real water regimes of nuclear power plants. The transfer processes of equipment are considered. Heat and mass transfer are analyzed in the pre-crisis regions of steam-generating passages with non-permeable surfaces, and in capillary-porous structures. Attention is given to forced convection boiling crises and top post-DNB heat transfer. Data on two-phase hydrodynamics in straight and curved channels are correlated and safety aspects of nuclear power plants are discussed

Application of a new methodology to evaluate Dnb limits based on statistical propagation of uncertainties

International Nuclear Information System (INIS)

Machado, Marcio Dornellas

1998-09-01

One of the most important thermalhydraulics safety parameters is the DNBR (Departure from Nucleate Boiling Ratio). The current methodology in use at Eletronuclear to determine DNBR is extremely conservative and may result in penalties to the reactor power due to an increase plugging level of steam generator tubes. This work uses a new methodology to evaluate DNBR, named mini-RTDP. The standard methodology (STDP) currently in use establishes a limit design value which cannot be surpassed. This limit value is determined taking into account the uncertainties of the empirical correlation used in COBRA IIC/MIT code, modified to Angra 1 conditions. The correlation used is the Westinghouse's W-3 and the minimum DNBR (MDBR) value cannot be less than 1.3. The new methodology reduces the excessive level of conservatism associated with the parameters used in the DNBR calculation, which take most unfavorable values in the STDP methodology, by using their best estimate values. The final goal is to obtain a new DNBR design limit which will provide a margin gain due to more realistic parameters values used in the methodology. (author)

Automated JPSS VIIRS GEO code change testing by using Chain Run Scripts

Science.gov (United States)

Chen, W.; Wang, W.; Zhao, Q.; Das, B.; Mikles, V. J.; Sprietzer, K.; Tsidulko, M.; Zhao, Y.; Dharmawardane, V.; Wolf, W.

2015-12-01

The Joint Polar Satellite System (JPSS) is the next generation polar-orbiting operational environmental satellite system. The first satellite in the JPSS series of satellites, J-1, is scheduled to launch in early 2017. J1 will carry similar versions of the instruments that are on board of Suomi National Polar-Orbiting Partnership (S-NPP) satellite which was launched on October 28, 2011. The center for Satellite Applications and Research Algorithm Integration Team (STAR AIT) uses the Algorithm Development Library (ADL) to run S-NPP and pre-J1 algorithms in a development and test mode. The ADL is an offline test system developed by Raytheon to mimic the operational system while enabling a development environment for plug and play algorithms. The Perl Chain Run Scripts have been developed by STAR AIT to automate the staging and processing of multiple JPSS Sensor Data Record (SDR) and Environmental Data Record (EDR) products. JPSS J1 VIIRS Day Night Band (DNB) has anomalous non-linear response at high scan angles based on prelaunch testing. The flight project has proposed multiple mitigation options through onboard aggregation, and the Option 21 has been suggested by the VIIRS SDR team as the baseline aggregation mode. VIIRS GEOlocation (GEO) code analysis results show that J1 DNB GEO product cannot be generated correctly without the software update. The modified code will support both Op21, Op21/26 and is backward compatible with SNPP. J1 GEO code change version 0 delivery package is under development for the current change request. In this presentation, we will discuss how to use the Chain Run Script to verify the code change and Lookup Tables (LUTs) update in ADL Block2.

FY 1993 progress report on the ANS thermal-hydraulic test loop operation and results

Energy Technology Data Exchange (ETDEWEB)

Siman-Tov, M.; Felde, D.K.; Farquharson, G. [and others

1994-07-01

The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). Highly subcooled heavy-water coolant flows vertically upward at a very high mass flux of almost 27 MG/m{sup 2}-s. In a parallel fuel plate configuration as in the ANSR, the flow is subject to a potential excursive static-flow instability that can very rapidly lead to flow starvation and departure from nucleate boiling (DNB) in the ``hot channel``. The current correlations and experimental data bases for flow excursion (FE) and critical heat flux (CHF) seldom evaluate the specific combination of ANSR operating parameters. The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 17 MW/m{sup 2}, a mass flux range of 8 to 28 Mg/m{sup 2}-s, an exit pressure range of 1.4 to 2.1 MPa, and an inlet temperature range of 40 to 50 C. FE experiments were also conducted using as ``soft`` a system as possible to secure a true FE phenomena (actual secondary burnout). True DNB experiments under similar conditions were also conducted. To the author`s knowledge, no other FE data have been reported in the literature to date that dover such a combination of conditions of high mass flux, high heat flux, and moderately high pressure.

Engagement of National Board of Examinations in strengthening public health education in India: present landscape, opportunities and future directions.

Science.gov (United States)

Sharma, Anjali; Zodpey, Sanjay; Batra, Bipin

2014-01-01

A trained and adequate heath workforce forms the crux in designing, implementing and monitoring health programs and delivering quality health services. Education is recognized as a critical instrument for creating such trained health professionals who can effectively address the 21 st century health challenges. At present, the Public Health Education in India is offered through medical colleges and also outside the corridors of medical colleges which was not the scenario earlier. Traditionally, Public Health Education has been a domain of medical colleges and was open for medical graduates only. In order to standardize the Postgraduate Medical Education in India, the National Board of Examinations (NBE) was set up as an independent autonomous body of its kind in the country in the field of medical sciences with the prime objective of improving the quality of the medical education. NBE has also played a significant role in enhancing Public Health Education in India through its Diplomat of National Board (DNB) Programs in Social and Preventive Medicine, Health and Hospital Administration, Maternal and Child Health, Family Medicine and Field Epidemiology. It envisions creating a cadre of skilled and motivated public health professionals and also developing a roadmap for postgraduate career pathways. However, there still exists gamut of opportunities for it to engage in expanding the scope of Public Health Education. It can play a key role in accreditation of public health programs and institutions which can transform the present landscape of education of health professionals. It also needs to revisit and re-initiate programs like DNB in Tropical Medicine and Occupational Health which were discontinued. The time is imperative for NBE to seize these opportunities and take necessary actions in strengthening and expanding the scope of Public Health Education in India.

Genetic correlations between conformation traits and radiographic findings in the limbs of German Warmblood riding horses

Directory of Open Access Journals (Sweden)

Distl Ottmar

2006-11-01

Full Text Available Abstract Studbook inspection (SBI data of 20 768 German Warmblood mares and radiography results (RR data of 5102 Hanoverian Warmblood horses were used for genetic correlation analyses. The scores on a scale from 0 to 10 were given for conformation and basic quality of gaits, resulting in 14 SBI traits which were used for the correlation analyses. The radiographic findings considered included osseous fragments in fetlock (OFF and hock joints (OFH, deforming arthropathy in hock joints (DAH and distinct radiographic findings in the navicular bones (DNB which were analyzed as binary traits, and radiographic appearance of the navicular bones (RNB which was analyzed as a quasi-linear trait. Genetic parameters were estimated multivariately in linear animal models with REML using information on 24 448 horses with SBI and/or RR records. The ranges of heritability estimates were h2 = 0.14–0.34 for the RR traits and h2 = 0.09–0.50 for the SBI traits. Negative additive genetic correlations of rg = -0.19 to -0.56 were estimated between OFF and conformation of front and hind limbs and walk at hand, and between DNB and hind limb conformation. There were indications of negative additive genetic correlations between DAH and all SBI traits, but because of low prevalence and low heritability of DAH, these results require further scrutiny. Positive additive genetic correlations of rg = 0.37–0.52 were estimated between OFF and withers height and between OFH and withers height, indicating that selection for taller horses will increase disposition to develop OFF and OFH. Selection of broodmares with regards to functional conformation will assist, but cannot replace possible selection against radiographic findings in the limbs of young Warmblood riding horses, particularly with regards to OFF.

Status of the R&D activities to the design of an ITER core CXRS diagnostic system

Energy Technology Data Exchange (ETDEWEB)

Mertens, Philippe, E-mail: ph.mertens@fz-juelich.de [Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich (FZJ), Trilateral Euregio Cluster, D-52425 Jülich (Germany); Castaño Bardawil, David A. [Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich (FZJ), Trilateral Euregio Cluster, D-52425 Jülich (Germany); Baross, Tétény [Wigner Research Centre for Physics (Wigner RCP), HU-1121 Budapest (Hungary); Biel, Wolfgang; Friese, Sebastian [Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich (FZJ), Trilateral Euregio Cluster, D-52425 Jülich (Germany); Hawkes, Nick [Culham Centre for Fusion Energy (CCFE), Culham OX14 3DB (United Kingdom); Jaspers, Roger J.E. [Eindhoven University of Technology (TU/e), PO Box 513, NL-5600 MB Eindhoven (Netherlands); Kotov, Vladislav; Krasikov, Yury; Krimmer, Andreas; Litnovsky, Andrey; Marchuk, Oleksander; Neubauer, Olaf [Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich (FZJ), Trilateral Euregio Cluster, D-52425 Jülich (Germany); Offermanns, Guido [Zentralinstitut für Engineering, Elektronik und Analytik ZEA-1 (Engineering and Technology), FZJ, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Panin, Anatoly [Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich (FZJ), Trilateral Euregio Cluster, D-52425 Jülich (Germany); and others

2015-10-15

Highlights: • The CXRS diagnostic for the core plasma of ITER will provide observation of the dedicated diagnostic beam (DNB) over a wide radial range, roughly r/a = 0.7 to 0. • A high performance (étendue × transmission, dynamic range) is expected for the port plug system since the beam attenuation is large and the background light omnipresent. • The design is particularly challenging in view of the ITER environment, especially with respect to the first mirror which faces the plasma. • The current status of development is presented by detailing several sub-systems before a four years design phase under an FPA between F4E and the ITER core CXRS Consortium (IC3). - Abstract: The CXRS (Charge-eXchange Recombination Spectroscopy) diagnostic for the core plasma of ITER will be designed to provide observation of the dedicated diagnostic beam (DNB) over a wide radial range, roughly from a normalised radius r/a = 0.7 to close to the plasma axis. The collected light will be transported through the Upper Port Plug #3 (UPP3) to a bundle of fibres and ultimately to a set of remote spectrometers. The design is particularly challenging in view of the ITER environment of particle, heat and neutron fluxes, temperature cycles, electromagnetic loads, vibrations, expected material degradation and fatigue, constraints against tritium penetration, integration in the plug and limited opportunities for maintenance. Moreover, a high performance (étendue × transmission, dynamic range) is expected for the port plug system since the beam attenuation is large and the background light omnipresent, especially in terms of bremsstrahlung, line radiation and reflections. The present contribution will give an overview of the current status and activities which deal with the core CXRS system, summarising the investigations which have taken place before entering the actual development and design phase.

Investigation of the properties of indium tin oxide-organic contacts for optoelectronic applications

Energy Technology Data Exchange (ETDEWEB)

Stanculescu, A. [National Institute of Materials Physics, 105 bis Atomistilor Street, P.O. Box MG-7, Bucharest-Magurele 077125 (Romania)], E-mail: sanca@infim.ro; Stanculescu, F. [University of Bucharest, Faculty of Physics, 405 Atomistilor Street, P.O. Box MG-11, Bucharest-Magurele 077125 (Romania)

2007-10-15

This paper presents some investigations on the electrical transport properties of ITO/single (double) layer organic semiconductor (m-DNB, benzil, PTCDA, Alq3) contacts in SIS-like (ITO/organic/Si) and MIS-like (ITO/organic/metal) heterostructures. The I-V characteristics have emphasised the injection properties of different contacts and the effect of space charge limited currents in correlation with the type and preparation conditions of the contacts. We have studied the influence of the type of contact (In/ITO; In/Al) on the electrical conduction in Alq3/PTCDA/Si/In heterostructure. In a planar grid contact configuration for In/Al/PTCDA/Al/In structure we have observed the effect of the low electric field on the shape of the I-V characteristic.

Assembly and gap management strategy for the ITER NBI vessel passive magnetic shield

Energy Technology Data Exchange (ETDEWEB)

Ríos, Luis, E-mail: luis.rios@ciemat.es [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Ahedo, Begoña; Alonso, Javier; Barrera, Germán; Cabrera, Santiago; Rincón, Esther; Ramos, Francisco [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); El-Ouazzani, Anass; Graceffa, Joseph; Urbani, Marc; Shah, Darshan [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3 – 07/08, 08019 Barcelona (Spain)

2015-10-15

The neutral beam system for ITER consists of two heating and current drive neutral ion beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. The HNB Passive Magnetic Shield (PMS) works in conjunction with the active compensation/correction coils to limit the magnetic field inside the Beam Line Vessel (BLV), Beam Source Vessel (BSV), High Voltage Bushing (HVB) and Transmission Line (TL) elbow to acceptable levels that do not interfere with the operation of the HNB components. This paper describes the current design of the PMS, having had only minor modifications since the preliminary design review (PDR) held in IO in April 2013, and the assembly strategy for the vessel PMS.

Curative effects of sodium fusidate on the development of dinitrobenzenesulfonic acid-induced colitis in rats

DEFF Research Database (Denmark)

Di Marco, Roberto; Mangano, Katia; Quattrocchi, Cinzia

2003-01-01

Fusidic acid and sodium fusidate (fusidin) are antibiotics with low toxicity and powerful immunomodulatory activities in vitro and in vivo. In this study we have evaluated the effect of fusidin on the development of dinitrobenzenesulfonic acid (DNB)-induced colitis in rats that serves....... These entailed a significant reduction in body weight loss, smaller increase in colon weights, milder macroscopic damage, and lower histological scores. In addition, when sacrificed at the end of the study, fusidin-treated rats had significantly lower blood levels of tumor necrosis factor alpha and interferon......-gamma compared with untreated controls. The present findings concur with the beneficial actions of fusidin in a pilot study conducted in patients with Crohn's disease and warrant controlled studies in humans with IBD....

Theoretical and experimental studies on critical heat flux in subcooled boiling and vertical flow geometry; Badania teoretyczne i eksperymentalne kryzysu wrzenia w warunkach wrzenia przechlodzonego w przeplywie w kanale pionowym

Energy Technology Data Exchange (ETDEWEB)

Staron, E. [Institute of Atomic Energy, Otwock-Swierk (Poland)

1996-12-31

Critical Heat Flux is a very important subject of interest due to design, operation and safety analysis of nuclear power plants. Every new design of the core must be thoroughly checked. Experimental studies have been performed using freon as a working fluid. The possibility of transferring of results into water equivalents has been proved. The experimental study covers vertical flow, annular geometry over a wide range of pressure, mass flow and temperature at inlet of test section. Theoretical models of Critical Heat Flux have been presented but only those which cover DNB. Computer programs allowing for numerical calculations using theoretical models have been developed. A validation of the theoretical models has been performed in accordance with experimental results. (author). 83 refs, 32 figs, 4 tabs.

Theoretical and experimental studies on critical heat flux in subcooled boiling and vertical flow geometry; Badania teoretyczne i eksperymentalne kryzysu wrzenia w warunkach wrzenia przechlodzonego w przeplywie w kanale pionowym

Energy Technology Data Exchange (ETDEWEB)

Staron, E [Institute of Atomic Energy, Otwock-Swierk (Poland)

1997-12-31

Critical Heat Flux is a very important subject of interest due to design, operation and safety analysis of nuclear power plants. Every new design of the core must be thoroughly checked. Experimental studies have been performed using freon as a working fluid. The possibility of transferring of results into water equivalents has been proved. The experimental study covers vertical flow, annular geometry over a wide range of pressure, mass flow and temperature at inlet of test section. Theoretical models of Critical Heat Flux have been presented but only those which cover DNB. Computer programs allowing for numerical calculations using theoretical models have been developed. A validation of the theoretical models has been performed in accordance with experimental results. (author). 83 refs, 32 figs, 4 tabs.

Theoretical and experimental studies on critical heat flux in subcooled boiling and vertical flow geometry

International Nuclear Information System (INIS)

Staron, E.

1996-01-01

Critical Heat Flux is a very important subject of interest due to design, operation and safety analysis of nuclear power plants. Every new design of the core must be thoroughly checked. Experimental studies have been performed using freon as a working fluid. The possibility of transferring of results into water equivalents has been proved. The experimental study covers vertical flow, annular geometry over a wide range of pressure, mass flow and temperature at inlet of test section. Theoretical models of Critical Heat Flux have been presented but only those which cover DNB. Computer programs allowing for numerical calculations using theoretical models have been developed. A validation of the theoretical models has been performed in accordance with experimental results. (author). 83 refs, 32 figs, 4 tabs

Efficacious of estimatives of thermal-hydraulic conditions of the PWR core by measured parameters

International Nuclear Information System (INIS)

Camargo, C.T.M.; Pontedeiro, A.C.

1985-01-01

Using ALMOD 3W2 and COBRA IIIP computer codes an evaluation of usual methods of estimatives of heat transfer conditions in the PWR core was made, using variables of the monitored processes. It was done a parametric study in conditions of the permanent regim to verify the influence of variables such as, pressure, temperature and power in the value of critical heat flux. Parameters to prevent the DNB phenomenon in KWU power plants and Westinghouse were calculated and implemented in the ALMOD 3W2 program to estimate the DNBR evolution. It was identified a common origin to both methods and comparing with detailed calculations of the COBRA IIIP code, it was settled limitations in the application of parameters. (M.C.K.) [pt

Improving the peak power density estimation for the DNBR trip signal

International Nuclear Information System (INIS)

Moreira, Joao M. L.; Souza, Rose Mary G.P.

2002-01-01

The departure from nucleate boiling (DNB) core protection in PWR reactors is usually carried out through the over temperature trip or the instantaneous minimum DNB ratio (DNBR) trip. The protection is obtained through specialized correlations or fast digital computer simulators that infer the core power level, and local coolant thermal and flow conditions out of process variables furnished by the instrumentation. The power density distribution information is usually expressed in terms of F q , the power peak factor, and its location. F q , in its turn, can be determined through the control rod position or, more often, through the power axial offset (AO) F q =f (AO, control rod positions). The AO, defined as the difference between upper and lower long ion chambers signals, is supplied for each channel by separate sets of out-of-core detectors positioned 90 or 120 degrees apart in plan. The AO is given by AO=(S t -S b )/(S t +S b ) where S t and S b are the out-of-core signals from the top and the bottom sections, respectively. In current PWRs a large penalty is imposed to the result of the first equation, because of the difficult of inferring with good accuracy the peak factor from the AO obtained from the out-of-core instrumentation. This ends up reducing the plant capacity factor. In this work, the f function in the first equation, which correlates the power peak factor with the axial offset yielded by out-of-core detectors and control rod positions, is obtained through a combination of specific experiments in the IPEN/MB-01 zero-power reactor and calculation results. For improving the peak factor estimation, it is necessary to consider accurately the response of the out-of-core detectors to different power density distribution in the core. This task is not easily accomplished through calculation due to the difficulties involved in the necessary neutron transport treatment for the out-of-core detector responses

A formal approach for the prediction of the critical heat flux in subcooled water

Energy Technology Data Exchange (ETDEWEB)

Lombardi, C. [Polytechnic of Milan (Italy)

1995-09-01

The critical heat flux (CHF) in subcooled water at high mass fluxes are not yet satisfactory correlated. For this scope a formal approach is here followed, which is based on an extension of the parameters and the correlation used for the dryout prediction for medium high quality mixtures. The obtained correlation, in spite of its simplicity and its explicit form, yields satisfactory predictions, also when applied to more conventional CHF data at low-medium mass fluxes and high pressures. Further improvements are possible, if a more complete data bank will be available. The main and general open item is the definition of a criterion, depending only on independent parameters, such as mass flux, pressure, inlet subcooling and geometry, to predict whether the heat transfer crisis will result as a DNB or a dryout phenomenon.

Comparison of thermal margin for W-3 R frid and WRB-1 correlations, for STDP and ITDP, RTDP method

Energy Technology Data Exchange (ETDEWEB)

Song, D. S. [KEPRI, Taejon (Korea, Republic of)

1999-05-01

DNBR sensitivity studies ware performed and Design Limit DNBRs were calculated by W-3 R grid and WRB-1 DNB correlations using ITDP(Improved Thermal Design Procedure) for 16 x 16 standard fuel assembly. The results of ITDP design limits using W-3 R grid and WRB-1 correlation were found to be 1.541(typical)/1.464(thimble) and 1.37(typical)/1.36(thimble) respectively. For thermal margin comparison, between W-3 R grid and WRB-1 correlation, minimum DNBRs for several cases were calculated and compared with Design Limit DNBR. It is found that around 7.8 % of thermal margin can be increase by correlation change W3-R to WRB-1. The additional thermal margin of 12-58 % can be obtained by adoption the RTDP or ITDP instead of STDP procedure.

Comparison of thermal margin for W-3 R frid and WRB-1 correlations, for STDP and ITDP, RTDP method

International Nuclear Information System (INIS)

Song, D. S.

1999-01-01

DNBR sensitivity studies ware performed and Design Limit DNBRs were calculated by W-3 R grid and WRB-1 DNB correlations using ITDP(Improved Thermal Design Procedure) for 16 x 16 standard fuel assembly. The results of ITDP design limits using W-3 R grid and WRB-1 correlation were found to be 1.541(typical)/1.464(thimble) and 1.37(typical)/1.36(thimble) respectively. For thermal margin comparison, between W-3 R grid and WRB-1 correlation, minimum DNBRs for several cases were calculated and compared with Design Limit DNBR. It is found that around 7.8 % of thermal margin can be increase by correlation change W3-R to WRB-1. The additional thermal margin of 12-58 % can be obtained by adoption the RTDP or ITDP instead of STDP procedure

Research and development on next generation reactor (phase I)

International Nuclear Information System (INIS)

Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong

1994-10-01

The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author)

Research and development on next generation reactor (phase I)

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); and others

1994-10-01

The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author).

A formal approach for the prediction of the critical heat flux in subcooled water

International Nuclear Information System (INIS)

Lombardi, C.

1995-01-01

The critical heat flux (CHF) in subcooled water at high mass fluxes are not yet satisfactory correlated. For this scope a formal approach is here followed, which is based on an extension of the parameters and the correlation used for the dryout prediction for medium high quality mixtures. The obtained correlation, in spite of its simplicity and its explicit form, yields satisfactory predictions, also when applied to more conventional CHF data at low-medium mass fluxes and high pressures. Further improvements are possible, if a more complete data bank will be available. The main and general open item is the definition of a criterion, depending only on independent parameters, such as mass flux, pressure, inlet subcooling and geometry, to predict whether the heat transfer crisis will result as a DNB or a dryout phenomenon

Safety analysis of the IAEA reference research reactor during loss of flow accident using the code MERSAT

International Nuclear Information System (INIS)

Hainoun, A.; Ghazi, N.; Abdul-Moaiz, B. Mansour

2010-01-01

Using the thermal hydraulic code MERSAT detailed model including primary and secondary loop was developed for the IAEA's reference research reactor MTR 10 MW. The developed model enables the simulation of expected neutronic and thermal hydraulic phenomena during normal operation, reactivity and loss of flow accidents. Two different loss of flow accident (LOFA) have been simulated using slow and fast decrease time of core mass flow. In both cases the expected flow reversal from downward forced to upward natural circulation has been successfully simulated. The results indicate that in both accidents the limit of onset of subcooled boiling was not arrived and consequently no exceed of design limits in term of thermal hydraulic instability or DNB is observed. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermal hydraulic system codes.

Perspectives on the neuroscience of alcohol from the National Institute on Alcohol Abuse and Alcoholism.

Science.gov (United States)

Reilly, Matthew T; Noronha, Antonio; Warren, Kenneth

2014-01-01

Mounting evidence over the last 40 years clearly indicates that alcoholism (alcohol dependence) is a disorder of the brain. The National Institute on Alcohol Abuse and Alcoholism (NIAAA) has taken significant steps to advance research into the neuroscience of alcohol. The Division of Neuroscience and Behavior (DNB) was formed within NIAAA in 2002 to oversee, fund, and direct all research areas that examine the effects of alcohol on the brain, the genetic underpinnings of alcohol dependence, the neuroadaptations resulting from excessive alcohol consumption, advanced behavioral models of the various stages of the addiction cycle, and preclinical medications development. This research portfolio has produced important discoveries in the etiology, treatment, and prevention of alcohol abuse and dependence. Several of these salient discoveries are highlighted and future areas of neuroscience research on alcohol are presented. © 2014 Elsevier B.V. All rights reserved.

Recent advances in modeling and validation of nuclear thermal-hydraulics applications with NEPTUNE CFD - 15471

International Nuclear Information System (INIS)

Guingo, M.; Baudry, C.; Hassanaly, M.; Lavieville, J.; Mechitouna, N.; Merigoux, N.; Mimouni, S.; Bestion, D.; Coste, P.; Morel, C.

2015-01-01

NEPTUNE CFD is a Computational Multi-(Fluid) Dynamics code dedicated to the simulation of multiphase flows, primarily targeting nuclear thermo-hydraulics applications, such as the departure from nuclear boiling (DNB) or the two-phase Pressurized Thermal Shock (PTS). It is co-developed within the joint research/development project NEPTUNE (AREVA, CEA, EDF, IRSN) since 2001. Over the years, to address the aforementioned applications, dedicated physical models and numerical methods have been developed and implemented in the code, including specific sets of models for turbulent boiling flows and two-phase non-adiabatic stratified flows. This paper aims at summarizing the current main modeling capabilities of the code, and gives an overview of the associated validation database. A brief summary of emerging applications of the code, such as containment simulation during a potential severe accident or in-vessel retention, is also provided. (authors)

Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

Science.gov (United States)

2013-06-01

approach is left to future investigators, in addition to analysis of the impact of axial power on DNB calculations. 2.5 Plutonium Vector The...H23 F06 H40 H75 G42 07 1 G13 H49 G84 H13 G66 H14 G29 H67 G21 H38 G57 H57 G74 H08 F08 08 1 F26 H80 H25 F25 H47 F11 H63...J52 H58 H73 H22 J74 H39 J24 H38 06 1 H16 J68 H15 H51 H81 G31 J48 G15 J06 G23 J23 G06 J40 J75 H42 07 1 H13 J49 H84 J13 H66

Standardization of the methodology used for fuel pressure drop evaluation to improve hydraulic calculation of heterogeneous cores

International Nuclear Information System (INIS)

Le Borgne, E.; Mattei, A.

1994-01-01

Continuous searching for safer and more efficient fuel, and diversification of fuel supply have as a consequence a possible change in the characteristics of the fuel assemblies used in nuclear reactors. By partially refueling cores with new assemblies, nuclear power plant operators are confronted with the problem of heterogeneous cores. The complexity of the problem increases as products diversify in isotopic concentration, types of alloy, size and shape of structure components. This document will focus strictly on the differences in hydraulic resistance related to the modifications in grid structures having no effect on DNB correlations. Although this is an extremely simplified approach to the problem, establishing data to evaluate the hydraulic compatibility between two different assemblies can be difficult, and if not controlled closely, can lead to false conclusions that may affect the operation and safety of the reactor. (authors). 2 figs

Critical heat flux and transition boiling characteristics for a sodium-heated steam generator tube for LMFBR applications

Energy Technology Data Exchange (ETDEWEB)

Wolf, S.; Holmes, D.H.

1977-04-01

An experimental program was conducted to characterize critical heat flux (CHF) in a sodium-heated steam generator tube model at a proposed PLBR steam generator design pressure of 7.2 MPa. Water was circulated vertically upward in the tube and the heating sodium was flowing counter-current downward. The experimental ranges were: mass flux, 110 to 1490 kg/s.m/sup 2/ (0.08 to 1.10 10/sup 6/ lbm/h.ft/sup 2/); critical heat flux, 0.16 to 1.86 MW/m/sup 2/ (0.05 to 0.59 10/sup 6/ Btu/h.ft/sup 2/); and critical quality, 0.48 to 1.0. The CHF phenomenon for the experimental conditions is determined to be dryout as opposed to departure from nucleate boiling (DNB). The data are divided into high- and low-mass flux regions.

Uncertainty analysis for the BEACON-COLSS core monitoring system application

International Nuclear Information System (INIS)

Morita, T.; Boyd, W.A.; Seong, K.B.

2005-01-01

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)

International Companies Withdrawal from Lithuania: Problematics and Alternative Solutions

Directory of Open Access Journals (Sweden)

Viktorija Tauraitė

2017-06-01

Full Text Available The main attention in this article is focused on the problematic of international companies’ withdrawal from Lithuania and presentation of alternative solutions of this problem. The macro(Sweden, Austria, Latvia, Lithuania, Estonia, Poland level analysis and micro (“Coca-Cola”, “Nordea” and DNB, “Orkla” level analysis showed that competitiveness, business conditions, employment relations, institutional environment and innovation should be improved and the corruption should be reduced in Lithuania. It is advisable that current Lithuanian Labour Code should be revised in order to increase the efficiency of labour relations. It is found out that the significance of “Coca-Cola”company is the highest in the context of the withdrawing companies from Lithuania. It is assumed that the most rational solution for each company is to move from Lithuania to another country.

Increase in VVER type reactor critical heat fluxes due to placing the mixing grids

International Nuclear Information System (INIS)

Bezrukov, Y.; Lisenkov, E.; Vasilchenko, I.

2011-01-01

The report deals with the results of studies of critical heat fluxes (CHF) on the models of VVER type reactor fuel assembly models equipped with the 'Vihr' intensifiers-grids. The models are the seven-rod bundles with the uniform and non-uniform axial power that correspond to two periods of FA operation i.e. beginning of cycle and end of cycle. The experiments performed showed that the mixing grids of this type are capable of increasing the FA burnout power. The power ascension rate depends on both coolant pressure and steam quality value in the CHF point. Placing the mixing grids in the bundle upper spans results in shifting the point of DNB occurrence downward along the FA height. The experimental data obtained will be used to develop the correlations for determining the CHF in the FA equipped with the mixing grids. (authors)

Design of Data Acquisition and Control System for Indian Test Facility of Diagnostics Neutral Beam

International Nuclear Information System (INIS)

Soni, Jignesh; Tyagi, Himanshu; Yadav, Ratnakar; Rotti, Chandramouli; Bandyopadhyay, Mainak; Bansal, Gourab; Gahluat, Agrajit; Sudhir, Dass; Joshi, Jaydeep; Prasad, Rambilas; Pandya, Kaushal; Shah, Sejal; Parmar, Deepak; Chakraborty, Arun

2015-01-01

Highlights: • More than 900 channels Data Acquisition and Control System. • INTF DACS has been designed based on ITER-PCDH guidelines. • Separate Interlock and Safety system designed based on IEC 61508 standard. • Hardware selected from ITER slow controller and fast controller catalog. • Software framework based on ITER CODAC Core System and LabVIEW software. - Abstract: The Indian Test Facility (INTF) – a negative hydrogen ion based 100 kV, 60 A, 5 Hz modulated NBI system having 3 s ON/20 s OFF duty cycle. Prime objective of the facility is to install a full-scale test bed for the qualification of all Diagnostic Neutral Beam (DNB) parameters, prior to installation in ITER. The automated and safe operation of the INTF will require a reliable and rugged instrumentation and control system which provide control, data acquisition (DAQ), interlock and safety functions, referred as INTF-DACS. The INTF-DACS has been decided to be design based on the ITER CODAC architecture and ITER-PCDH guidelines since the technical understanding of CODAC technology gained from this will later be helpful in development of plant system I&C for DNB. For complete operation of the INTF, approximately 900 numbers of signals are required to be superintending by the DACS. In INTF conventional control loop time required is within the range of 5–100 ms and for DAQ except high-end diagnostics, required sampling rates in range of 5 sample per second (Sps) to 10 kSps; to fulfill these requirements hardware components have been selected from the ITER slow and fast controller catalogs. For high-end diagnostics required sampling rates up to 100 MSps normally in case of certain events, therefore event and burst based DAQ hardware has been finalized. Combined use of CODAC core software (CCS) and NI-LabVIEW has been finalized due to the fact that full required DAQ support is not available in present version of CCS. Interlock system for investment protection of facility and Safety system for

Effects of Bubble-Mediated Processes on Nitrous Oxide Dynamics in Denitrifying Bioreactors

Science.gov (United States)

McGuire, P. M.; Falk, L. M.; Reid, M. C.

2017-12-01

To mitigate groundwater and surface water impacts of reactive nitrogen (N), agricultural and stormwater management practices can employ denitrifying bioreactors (DNBs) as low-cost solutions for enhancing N removal. Due to the variable nature of hydrologic events, DNBs experience dynamic flows which can impact physical and biological processes within the reactors and affect performance. A particular concern is incomplete denitrification, which can release the potent greenhouse gas nitrous oxide (N2O) to the atmosphere. This study aims to provide insight into the effects of varying hydrologic conditions upon the operation of DNBs by disentangling abiotic and biotic controls on denitrification and N2O dynamics within a laboratory-scale bioreactor. We hypothesize that under transient hydrologic flows, rising water levels lead to air entrapment and bubble formation within the DNB porous media. Mass transfer of oxygen (O2) between trapped gas and liquid phases creates aerobic microenvironments that can inhibit N2O reductase (NosZ) enzymes and lead to N2O accumulation. These bubbles also retard N2O transport and make N2O unavailable for biological reduction, further enhancing atmospheric fluxes when water levels fall. The laboratory-scale DNB permits measurements of longitudinal and vertical profiles of dissolved constituents as well as trace gas concentrations in the reactor headspace. We describe a set of experiments quantifying denitrification pathway biokinetics under steady-state and transient hydrologic conditions and evaluate the role of bubble-mediated processes in enhancing N2O accumulation and fluxes. We use sulfur hexafluoride and helium as dissolved gas tracers to examine the impact of bubble entrapment upon retarded gas transport and enhanced trace gas fluxes. A planar optode sensor within the bioreactor provides near-continuous 2-D profiles of dissolved O2 within the bioreactor and allows for identification of aerobic microenvironments. We use qPCR to

Design of Data Acquisition and Control System for Indian Test Facility of Diagnostics Neutral Beam

Energy Technology Data Exchange (ETDEWEB)

Soni, Jignesh, E-mail: jsoni@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar 382 428, Gujarat (India); Tyagi, Himanshu; Yadav, Ratnakar; Rotti, Chandramouli; Bandyopadhyay, Mainak [ITER-India, Institute for Plasma Research, Gandhinagar 380 025, Gujarat (India); Bansal, Gourab; Gahluat, Agrajit [Institute for Plasma Research, Bhat, Gandhinagar 382 428, Gujarat (India); Sudhir, Dass; Joshi, Jaydeep; Prasad, Rambilas [ITER-India, Institute for Plasma Research, Gandhinagar 380 025, Gujarat (India); Pandya, Kaushal [Institute for Plasma Research, Bhat, Gandhinagar 382 428, Gujarat (India); Shah, Sejal; Parmar, Deepak [ITER-India, Institute for Plasma Research, Gandhinagar 380 025, Gujarat (India); Chakraborty, Arun [Institute for Plasma Research, Bhat, Gandhinagar 382 428, Gujarat (India)

2015-10-15

Highlights: • More than 900 channels Data Acquisition and Control System. • INTF DACS has been designed based on ITER-PCDH guidelines. • Separate Interlock and Safety system designed based on IEC 61508 standard. • Hardware selected from ITER slow controller and fast controller catalog. • Software framework based on ITER CODAC Core System and LabVIEW software. - Abstract: The Indian Test Facility (INTF) – a negative hydrogen ion based 100 kV, 60 A, 5 Hz modulated NBI system having 3 s ON/20 s OFF duty cycle. Prime objective of the facility is to install a full-scale test bed for the qualification of all Diagnostic Neutral Beam (DNB) parameters, prior to installation in ITER. The automated and safe operation of the INTF will require a reliable and rugged instrumentation and control system which provide control, data acquisition (DAQ), interlock and safety functions, referred as INTF-DACS. The INTF-DACS has been decided to be design based on the ITER CODAC architecture and ITER-PCDH guidelines since the technical understanding of CODAC technology gained from this will later be helpful in development of plant system I&C for DNB. For complete operation of the INTF, approximately 900 numbers of signals are required to be superintending by the DACS. In INTF conventional control loop time required is within the range of 5–100 ms and for DAQ except high-end diagnostics, required sampling rates in range of 5 sample per second (Sps) to 10 kSps; to fulfill these requirements hardware components have been selected from the ITER slow and fast controller catalogs. For high-end diagnostics required sampling rates up to 100 MSps normally in case of certain events, therefore event and burst based DAQ hardware has been finalized. Combined use of CODAC core software (CCS) and NI-LabVIEW has been finalized due to the fact that full required DAQ support is not available in present version of CCS. Interlock system for investment protection of facility and Safety system for

Die DDC auf neuen Wegen - verbale Sucheinstiege für klassifikatorisch erschlossene Titel

Directory of Open Access Journals (Sweden)

Christiane Maibach

2014-12-01

Full Text Available Die Dewey-Dezimalklassifikation (DDC ist ein mächtiges Instrument der klassifikatorischen Inhaltserschließung. Immer mehr Bibliotheken im In- und Ausland erkennen den Nutzen der international am weitesten verbreiteten Universalklassifikation. Die Deutsche Nationalbibliothek (DNB setzt die DDC seit 2006 für die inhaltliche Erschließung der Bibliografiereihen A, B und H ein. Seitdem wurden rund 800.000 Publikationen mit DDC-Notationen versehen. Der Nutzen von Klassifikationen ist unter Bibliothekar/innen unumstritten. Die Bibliotheksnutzer/innen hingegen wissen oft nicht, dass der Katalog ihrer Bibliothek auch eine systematische Recherche ermöglicht. Ihre Recherchestrategie ist meist verbal. Durch Internetsuchmaschinen wird diese Strategie noch bestärkt. Daher liegt es nahe, auch für die klassifikatorisch erschlossenen Titel einen verbalen Sucheinstieg zu schaffen. Die DDC enthält nicht nur die in vielen Klassifikationssystemen üblichen Registereinträge und Klassenbenennungen, sondern auch die im Rahmen des Projekts CrissCross in großem Umfang erstellten Verknüpfungen zu Schlagwörtern der Gemeinsamen Normdatei (GND, die für diesen Zweck ausgewertet werden können. The Dewey Decimal Classification (DDC is a powerful indexing tool. A rising number of libraries worldwide recognize the benefit of this universal classification. Since 2006 the German National Library (DNB has used the DDC to index titles belonging to series A, B and H of the German National Bibliography. In 2012 100,000 publications were indexed with DDC notations. The benefits of classifications for indexing is widely accepted among librarians. However, library users take a different point of view. Mostly, they are not even aware that their library's catalogue offers a systematic search. Most library users prefer verbal search strategies. This is enforced by internet search engines, which have changed the search habits of library users. Therefore, the obvious solution is

Delineating Spatial Patterns in Human Settlements Using VIIRS Nighttime Light Data: A Watershed-Based Partition Approach

Directory of Open Access Journals (Sweden)

Ting Ma

2018-03-01

Full Text Available As an informative proxy measure for a range of urbanization and socioeconomic variables, satellite-derived nighttime light data have been widely used to investigate diverse anthropogenic activities in human settlements over time and space from the regional to the national scale. With a higher spatial resolution and fewer over-glow and saturation effects, nighttime light data derived from the Visible Infrared Imaging Radiometer Suite (VIIRS instrument with day/night band (DNB, which is on the Suomi National Polar-Orbiting Partnership satellite (Suomi-NPP, may further improve our understanding of spatiotemporal dynamics and socioeconomic activities, particularly at the local scale. Capturing and identifying spatial patterns in human settlements from VIIRS images, however, is still challenging due to the lack of spatially explicit texture characteristics, which are usually crucial for general image classification methods. In this study, we propose a watershed-based partition approach by combining a second order exponential decay model for the spatial delineation of human settlements with VIIRS-derived nighttime light images. Our method spatially partitions the human settlement into five different types of sub-regions: high, medium-high, medium, medium-low and low lighting areas with different degrees of human activity. This is primarily based on the local coverage of locally maximum radiance signals (watershed-based and the rank and magnitude of the nocturnal radiance signal across the whole region, as well as remotely sensed building density data and social media-derived human activity information. The comparison results for the relationship between sub-regions with various density nighttime brightness levels and human activities, as well as the densities of different types of interest points (POIs, show that our method can distinctly identify various degrees of human activity based on artificial nighttime radiance and ancillary data. Furthermore

Margin benefit assessment of the YGN 3 cycle 1 fxy error files for COLSS and CPC overall uncertainty analyses

International Nuclear Information System (INIS)

Yoon, Rae Young; In, Wang Kee; Auh, Geun Sun; Kim, Hee Cheol; Lee, Sang Keun

1994-01-01

Margin benefits are quantitatively assessed for the Yonggwang Unit 3 (YGN 3) Cycle 1 planar radial peaking factor (Fxy) error files for each time-in-life, i.e., BOC, IOC, MOC and EOC. The generic Fxy error file (FXYMEQO) is presently used for Yonggwang Unit 3 Cycle 1 COLSS (Core Operating Limit Supervisory System) and CPC (Core Protection Calculator) Overall Uncertainty Analyses (OUA). However, because this file is more conservative than the plant/cycle specific Fxy error files, COLSS and CPC thermal margins (DNB-OPM) for the generic Fxy error file are less than those of the plant/cycle specific Fxy error file. Therefore, the YGN 3 Cycle 1 Fxy error files were generated and analyzed by the modified codes for Yonggwang Plants. The YGN 3 Cycle 1 Fxy error files gave the increased thermal margin by about 1% for COLSS and CPC, respectively

CASL Verification and Validation Plan

Energy Technology Data Exchange (ETDEWEB)

Mousseau, Vincent Andrew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Dinh, Nam [North Carolina State Univ., Raleigh, NC (United States)

2016-06-30

This report documents the Consortium for Advanced Simulation of LWRs (CASL) verification and validation plan. The document builds upon input from CASL subject matter experts, most notably the CASL Challenge Problem Product Integrators, CASL Focus Area leaders, and CASL code development and assessment teams. This document will be a living document that will track progress on CASL to do verification and validation for both the CASL codes (including MPACT, CTF, BISON, MAMBA) and for the CASL challenge problems (CIPS, PCI, DNB). The CASL codes and the CASL challenge problems are at differing levels of maturity with respect to validation and verification. The gap analysis will summarize additional work that needs to be done. Additional VVUQ work will be done as resources permit. This report is prepared for the Department of Energy’s (DOE’s) CASL program in support of milestone CASL.P13.02.

Study of Two-Phase Heat Transfer in Nano-fluids for Nuclear Applications

International Nuclear Information System (INIS)

Kim, S.J.; Truong, B.; Buongiorno, J.; Hu, L.W.; Bang, I.C.

2006-01-01

Nano-fluids are engineered colloidal suspensions of nano-particles in a base fluid. We are investigating the two-phase heat transfer behavior of water-based nano-fluids, to evaluate their potential use in nuclear applications, including the PWR primary coolant and PWR and BWR safety systems. A simple pool boiling wire experiment shows that a significant increase in Critical Heat Flux (CHF) can be achieved at modest nano-particle concentrations. For example, the CHF increases by 50% in nano-fluids with alumina nano-particles at 0.001%v concentration. The CHF enhancement appears to correlate with the presence of a layer of nano-particles that builds up on the heated surface during nucleate boiling. A review of the prevalent Departure from Nucleate Boiling (DNB) theories suggests that an alteration of the nucleation site density (brought about by the nano-particle layer) could plausibly explain the CHF enhancement. (authors)

Monitoring Disaster-Related Power Outages Using NASA Black Marble Nighttime Light Product

Science.gov (United States)

Wang, Z.; Román, M. O.; Sun, Q.; Molthan, A. L.; Schultz, L. A.; Kalb, V. L.

2018-04-01

Timely and accurate monitoring of disruptions to the electricity grid, including the magnitude, spatial extent, timing, and duration of net power losses, is needed to improve situational awareness of disaster response and long-term recovery efforts. Satellite-derived Nighttime Lights (NTL) provide an indication of human activity patterns and have been successfully used to monitor disaster-related power outages. The global 500 m spatial resolution National Aeronautics and Space Administration (NASA) Black Marble NTL daily standard product suite (VNP46) is generated from Visible Infrared Imaging Radiometer Suite (VIIRS) Day/Night Band (DNB) onboard the NASA/National Oceanic and Atmospheric Administration (NOAA) Suomi National Polar-orbiting Partnership (Suomi- NPP) satellite, which began operations on 19 January 2012. With its improvements in product accuracy (including critical atmospheric and BRDF correction routines), the VIIRS daily Black Mable product enables systematic monitoring of outage conditions across all stages of the disaster management cycle.

From OLS to VIIRS, an overview of nighttime satellite aerosol retrievals using artificial light sources

Science.gov (United States)

Zhang, J.; Miller, S. D.; Reid, J. S.; Hyer, E. J.; McHardy, T. M.

2015-12-01

Compared to abundant daytime satellite-based observations of atmospheric aerosol, observations at night are relatively scarce. In particular, conventional satellite passive imaging radiometers, which offer expansive swaths of spatial coverage compared to non-scanning lidar systems, lack sensitivity to most aerosol types via the available thermal infrared bands available at night. In this talk, we make the fundamental case for the importance of nighttime aerosol information in forecast models, and the need to mitigate the existing nocturnal gap. We review early attempts at estimating nighttime aerosol optical properties using the modulation of stable artificial surface lights. Initial algorithm development using DMSP Operational Linescan System (OLS) has graduated to refined techniques based on the Suomi-NPP Visible Infrared Imaging Radiometer Suite (VIIRS) Day/Night Band (DNB). We present examples of these retrievals for selected cases and compare the results to available surface-based point-source validation data.

Characteristic of The RSG-Gas Oxide Fuel Element Temperature Under Forced Convection And Natural Convection Mode

International Nuclear Information System (INIS)

Sudarmono

2000-01-01

One of the methods used for fuel element plate temperature measurement in RSG-Gas is a direct measurement. Evaluation on the measurement results were done by using HEATHYDE and NATCON code, which was then compared to the safety margin criteria. Results of thermalhydraulic measurement on transitional core both under forced and natural convection were compared with the results of calculations using the two codes. Measurement result for maximum fuel element plate temperature at typical working core of 30 MW, was 121 o C. The deviation between calculation and measurement result was under 9.75 %. Under normal operation, safety margin on DNB and OFI are 3.56 and 2.60, respectively. Natcon calculation result showed that the typical working core under the natural circulation mode, an onset of nucleate boiling (ONB)occurred at a core power level of 826 kW (2.8% of the nominal power)

Usefulness of charge-transfer complexation for the assessment of sympathomimetic drugs: Spectroscopic properties of drug ephedrine hydrochloride complexed with some π-acceptors

Science.gov (United States)

Refat, Moamen S.; Ibrahim, Omar B.; Saad, Hosam A.; Adam, Abdel Majid A.

2014-05-01

Recently, ephedrine (Eph) assessment in food products, pharmaceutical formulations, human fluids of athletes and detection of drug toxicity and abuse, has gained a growing interest. To provide basic data that can be used to assessment of Eph quantitatively based on charge-transfer (CT) complexation, the CT complexes of Eph with 7‧,8,8‧-tetracyanoquinodimethane (TCNQ), dichlorodicyanobenzoquinone (DDQ), 1,3-dinitrobenzene (DNB) or tetrabromothiophene (TBT) were synthesized and spectroscopically investigated. The newly synthesized complexes have been characterized via elemental analysis, IR, Raman, 1H NMR, and UV-visible spectroscopy. The formation constant (KCT), molar extinction coefficient (εCT) and other spectroscopic data have been determined using the Benesi-Hildebrand method and its modifications. The sharp, well-defined Bragg reflections at specific 2θ angles have been identified from the powder X-ray diffraction patterns. Thermal decomposition behavior of these complexes was also studied, and their kinetic thermodynamic parameters were calculated with Coats-Redfern and Horowitz-Metzger equations.

Thermal-hydraulic effects of transition to improved System 80TM fuel

International Nuclear Information System (INIS)

Rodack, T.; Joffre, P.F.; Kapoor, R.K.

2004-01-01

ABB CE's improved System 80 TM PWR fuel design includes GUARDIAN debris-resistant features and laser-welded Zircaloy grids. The GUARDIAN features include an Inconel grid with debris-filtering features located just above the Lower End Fitting, and a solid fuel rod bottom end cap that extends above the filtering features. Tests and analyses were done to establish the impact of these design improvements on fuel assembly hydraulic performance. Further analysis was done to determine the mixed core thermal-hydraulic performance as the transition is made over two fuel cycles to a full core of the improved System 80 TM fuel. Results confirm that the Thermal-Hydraulic (T-H) effects of the reduction in hydraulic resistance between the improved and resident fuel due to the laser-welded Zircaloy grids offsets the effects of the increased resistance GUARDIAN grid. Therefore, the mechanically improved System 80 TM fuel can be implemented with no net impact on Departure from Nucleate Boiling (DNB) margin in transition cores. (author)

Adsorptive removal of hydrophobic organic compounds by carbonaceous adsorbents: a comparative study of waste-polymer-based, coal-based activated carbon, and carbon nanotubes.

Science.gov (United States)

Lian, Fei; Chang, Chun; Du, Yang; Zhu, Lingyan; Xing, Baoshan; Liu, Chang

2012-01-01

Adsorption of the hydrophobic organic compounds (HOCs) trichloroethylene (TCE), 1,3-dichlorobenzene (DCB), 1,3-dinitrobenzene (DNB) and gamma-hexachlorocyclohexane (HCH) on five different carbonaceous materials was compared. The adsorbents included three polymer-based activated carbons, one coal-based activated carbon (F400) and multiwalled carbon nanotubes (MWNT). The polymer-based activated carbons were prepared using KOH activation from waste polymers: polyvinyl chloride (PVC), polyethyleneterephthalate (PET) and tire rubber (TR). Compared with F400 and MWNT, activated carbons derived from PVC and PET exhibited fast adsorption kinetics and high adsorption capacity toward the HOCs, attributed to their extremely large hydrophobic surface area (2700 m2/g) and highly mesoporous structures. Adsorption of small-sized TCE was stronger on the tire-rubber-based carbon and F400 resulting from the pore-filling effect. In contrast, due to the molecular sieving effect, their adsorption on HCH was lower. MWNT exhibited the lowest adsorption capacity toward HOCs because of its low surface area and characteristic of aggregating in aqueous solution.

The motional stark effect with laser-induced fluorescence diagnostic

Science.gov (United States)

Foley, E. L.; Levinton, F. M.

2010-05-01

The motional Stark effect (MSE) diagnostic is the worldwide standard technique for internal magnetic field pitch angle measurements in magnetized plasmas. Traditionally, it is based on using polarimetry to measure the polarization direction of light emitted from a hydrogenic species in a neutral beam. As the beam passes through the magnetized plasma at a high velocity, in its rest frame it perceives a Lorentz electric field. This field causes the H-alpha emission to be split and polarized. A new technique under development adds laser-induced fluorescence (LIF) to a diagnostic neutral beam (DNB) for an MSE measurement that will enable radially resolved magnetic field magnitude as well as pitch angle measurements in even low-field (experiments. An MSE-LIF system will be installed on the National Spherical Torus Experiment (NSTX) at the Princeton Plasma Physics Laboratory. It will enable reconstructions of the plasma pressure, q-profile and current as well as, in conjunction with the existing MSE system, measurements of radial electric fields.

Simulation of a two phase boiling flow in Poseidon geometry with Astrid steam-water software

International Nuclear Information System (INIS)

Larrauri, D.

1997-01-01

After different validation test runs in tube an annular geometries, the simulation of a subcooled boiling flow in a rod bundle geometry has been achieved with ASTRID Steam-Water software. The experiment we have simulated is the Poseidon experiment. It is a three heating tube geometry. The thermohydraulic conditions of the simulated flow are closed to the DNB conditions. The simulation results are analysed and compared against the available measurements of liquid and wall temperatures. ASTRID Steam-Water behaviour in such a geometry brings satisfaction. The wall and the liquid temperatures are well predicted in the different parts of the flow. The void fraction reaches 40 % in the vicinity of the heating rods. Besides, the evolution of the different calculated variables shows that a three-dimensional simulation gives capital information for the analyse of the physical phenomena involved in this kind of flow. The good results obtained in Poseidon geometry lead us to think about simulating and analyzing rod bundle flows with ASTRID Steam-Water code. (author)

MONITORING DISASTER-RELATED POWER OUTAGES USING NASA BLACK MARBLE NIGHTTIME LIGHT PRODUCT

Directory of Open Access Journals (Sweden)

Z. Wang

2018-04-01

Full Text Available Timely and accurate monitoring of disruptions to the electricity grid, including the magnitude, spatial extent, timing, and duration of net power losses, is needed to improve situational awareness of disaster response and long-term recovery efforts. Satellite-derived Nighttime Lights (NTL provide an indication of human activity patterns and have been successfully used to monitor disaster-related power outages. The global 500 m spatial resolution National Aeronautics and Space Administration (NASA Black Marble NTL daily standard product suite (VNP46 is generated from Visible Infrared Imaging Radiometer Suite (VIIRS Day/Night Band (DNB onboard the NASA/National Oceanic and Atmospheric Administration (NOAA Suomi National Polar-orbiting Partnership (Suomi- NPP satellite, which began operations on 19 January 2012. With its improvements in product accuracy (including critical atmospheric and BRDF correction routines, the VIIRS daily Black Mable product enables systematic monitoring of outage conditions across all stages of the disaster management cycle.

A deep belief network approach using VDRAS data for nowcasting

Science.gov (United States)

Han, Lei; Dai, Jie; Zhang, Wei; Zhang, Changjiang; Feng, Hanlei

2018-04-01

Nowcasting or very short-term forecasting convective storms is still a challenging problem due to the high nonlinearity and insufficient observation of convective weather. As the understanding of the physical mechanism of convective weather is also insufficient, the numerical weather model cannot predict convective storms well. Machine learning approaches provide a potential way to nowcast convective storms using various meteorological data. In this study, a deep belief network (DBN) is proposed to nowcast convective storms using the real-time re-analysis meteorological data. The nowcasting problem is formulated as a classification problem. The 3D meteorological variables are fed directly to the DBN with dimension of input layer 6*6*80. Three hidden layers are used in the DBN and the dimension of output layer is two. A box-moving method is presented to provide the input features containing the temporal and spatial information. The results show that the DNB can generate reasonable prediction results of the movement and growth of convective storms.

Power-Cooling-Mismatch Test Series. Test PCM-2: postirradiation examination

International Nuclear Information System (INIS)

Seiffert, S.L.

1977-03-01

The report describes the results of the postirradiation examination of four 0.91-m long pressurized water reactor (PWR) type, UO 2 -fueled zircaloy-clad fuel rods tested in an in-pile experimental investigation of pre- and post-departure from nucleate boiling (DNB) behavior of previously unirradiated fuel rods. None of the four fuel rods tested failed during testing or during reactor shut down. Visual and metallographic examination of the individual fuel rods indicated that the length of the high temperature zone of film boiling varied from rod to rod. Two of the four fuel rods showed evidence of damage by film boiling, characterized by oxide spalling and cladding collapse. Metallographic examination of these fuel rods showed internal cladding oxidation resulting from fuel-cladding reaction. Cladding embrittlement by oxidation is evaluated. A comparison of the cladding surface temperatures estimated metallographically for the separate fuel rods with cladding surface temperatures measured during testing and calculated from the Fuel Rod Analysis Program-Transient (FRAP-T) computer code is included

Adsorptive removal of hydrophobic organic compounds by carbonaceous adsorbents: A comparative study of waste-polymer-based,coal-based activated carbon, and carbon nanotubes

Institute of Scientific and Technical Information of China (English)

Fei Lian; Chun Chang; Yang Du; Lingyan Zhu; Baoshan Xing; Chang Liu

2012-01-01

Adsorption of the hydrophobic organic compounds (HOCs) trichloroethylene (TCE),1,3-dichlorobenzene (DCB),1,3-dinitrobenzene (DNB) and γ-hexachlorocyclohexane (HCH) on five different carbonaceous materials was compared.The adsorbents included three polymer-based activated carbons,one coal-based activated carbon (F400) and multiwalled carbon nanotubes (MWNT).The polymerbased activated carbons were prepared using KOH activation from waste polymers:polyvinyl chloride (PVC),polyethyleneterephthalate (PET) and tire rubber (TR).Compared with F400 and MWNT,activated carbons derived from PVC and PET exhibited fast adsorption kinetics and high adsorption capacity toward the HOCs,attributed to their extremely large hydrophobic surface area (2700 m2/g) and highly mesoporous structures.Adsorption of small-sized TCE was stronger on the tire-rubber-based carbon and F400 resulting from the pore-filling effect.In contrast,due to the molecular sieving effect,their adsorption on HCH was lower.MWNT exhibited the lowest adsorption capacity toward HOCs because of its low surface area and characteristic of aggregating in aqueous solution.

Development of Mitsubishi high thermal performance grid 1 - CFD applicability for thermal hydraulic design

International Nuclear Information System (INIS)

Ikeda, K.; Hoshi, M.

2001-01-01

Mitsubishi applied the Computational Fluid Dynamics (CFD) evaluation method for designing of the new lower pressure loss and higher DNB performance grid spacer. Reduction of pressure loss of the grid has been estimated by CFD. Also, CFD has been developed as a design tool to predict the coolant mixing ability of vane structures, that is to compare the relative peak spot temperatures around fuel rods at the same heat flux condition. These evaluations have been reflected to the new grid spacer design. The prototype grid was manufactured and some flow tests were performed to examine the thermal hydraulic performance, which were predicted by CFD. The experimental data of pressure loss was in good agreement with CFD prediction. The CFD prediction of flow behaviors at downstream of the mixing vanes was verified by detail cross-flow measurements at rod gaps by the rod LDV system. It is concluded that the applicability of the CFD evaluation method for the thermal hydraulic design of the grid is confirmed. (authors)

Studies on the assessment and validation of reactor dynamics models used in Finland

International Nuclear Information System (INIS)

Vanttola, T.

1993-10-01

Two reactor dynamics related computer codes of the calculation system at the Technical Research Centre of Finland have been assessed. The codes TRAB and SMATRA, have been examined from two points of view. First, models of some critical phenomena determining the worst fuel rod conditions during reactor transients have been evaluated on the basis of experimental information. Second, the the overall behaviour of the codes describing the dynamics of the reactor core and its cooling system has been studied on the basis of simulation of real transients and of performed safety analyses of selected accidents. The emphasis is on the VVER-440 reactors, but the generality of the methods has been demonstrated by showing that the key phenomena of the Chernobyl accident can be reproduced and analysed using the same calculation system. In the study the separate phenomena examined are single- and two-phase friction, post DNB heat transfer and critical heat flux in the VVER rod bundle. (60 refs., 11 figs., 4 tabs.)

Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model

International Nuclear Information System (INIS)

Oliveira, A.C.J.G. de.

1988-12-01

Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs

Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model; Aplicacao da teoria de perturbacao para calculos de sensibilidade em nucleos de reatores PWR, usando um modelo de dois canais

Energy Technology Data Exchange (ETDEWEB)

Oliveira, A.C.J.G. de

1988-12-01

Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs.

Convenience in white-collar crime: A resource perspective

Directory of Open Access Journals (Sweden)

Petter Gottschalk

2017-03-01

Full Text Available White-collar offenders have access to resources that make financial crime convenient. In the rare case of crime suspicion, resources are available in terms of professional attorney work, control over internal investigations, and public relations support. Hiring private investigators at an early stage of potential crime disclosure enables the organization to control the investigation mandate and influence the investigation process and the investigation output. Getting an early start on reconstruction of the past in terms of a fraud examination makes it possible for the suspect and the organization to influence what facts are relevant and how facts might be assessed in terms of possible violations of the penal code. Convenience aspects of private investigations are discussed in this article in terms of five internal investigations, two in the United States (General Motors and Lehman Brothers and three in Norway (Telenor VimpelCom, DNB Bank, and Norwegian Football Association. The aim of this research is to contribute insights into convenience associated with internal private investigations.

Comparisons of numerical simulations with ASTRID code against experimental results in rod bundle geometry for boiling flows

International Nuclear Information System (INIS)

Larrauri, D.; Briere, E.

1997-12-01

After different validation simulations of flows through cylindrical and annular channels, a subcooled boiling flow through a rod bundle has been simulated with ASTRID Steam-Water of software. The experiment simulated is called Poseidon. It is a vertical rectangular channel with three heating rods inside. The thermohydraulic conditions of the simulated flow were close to the DNB conditions. The simulation results were analysed and compared against the available measurements of liquid and wall temperatures. ASTRID Steam-Water produced satisfactory results. The wall and the liquid temperatures were well predicted in the different parts of the flow. The void fraction reached 40 % in the vicinity of the heating rods. The distribution of the different calculated variables showed that a three-dimensional simulation gives essential information for the analysis of the physical phenomena involved in this kind of flow. The good results obtained in Poseidon geometry will encourage future rod bundle flow simulations and analyses with ASTRID Steam-Water code. (author)

Critical heat flux tests for a 12 finned-element assembly

Energy Technology Data Exchange (ETDEWEB)

Yang, J., E-mail: Jun.Yang@cnl.ca; Groeneveld, D.C.; Yuan, L.Q.

2017-03-15

Highlights: • CHF tests for a 12 finned-fuel-element assembly at highly subcooled conditions. • Test approach to maximize experimental information and minimize heater failures. • Three series of tests were completed in vertical upward light water flow. • Bundle simulators of two axial power profiles and three heated lengths were tested. • Results confirm that the prediction method predicts lower CHF values than measured. - Abstract: An experimental study was undertaken to provide relevant data to validate the current critical heat flux (CHF) prediction method of the NRU driver fuel for safety analysis, i.e., to confirm no CHF occurrence below the predicted values. The NRU driver fuel assembly consists of twelve finned fuel elements arranged in two rings – three in the inner ring and nine in the outer ring. To satisfy the experimental objective tests at very high heat fluxes, very high mass velocities, and high subcoolings were conducted where the CHF mechanism is the departure from nucleate boiling (DNB). Such a CHF experiment can be very difficult, costly and time consuming since failure of the heating surface due to rupture or melting (physical burnout) is expected when the DNB type of CHF is reached. A novel experimental approach has been developed to maximize the amount of relevant experimental information on safe operating conditions in the tests, and to minimize any possible heater failures that inherently accompany the CHF occurrence at these conditions. Three series of tests using electrically heated NRU driver fuel simulators with three heated lengths and two axial power profiles (or axial heat flux distribution (AFD)) were completed in vertical upward light water flow. Each series of tests covered two mass flow rates, several heat flux levels, and local subcoolings that bound the ranges of interest for the analysis of postulated slow loss-of-regulation accident (LORA) and loss-of-flow accident (LOFA) scenarios. Tests for each mass flow rate of

Preliminary Analysis for K-DEMO Water Cooled Breeding Blanket Using MARS-KS

International Nuclear Information System (INIS)

Lee, Jeong-Hun; Kim, Geon-Woo; Park, Goon-Cherl; Cho, Hyoung-Kyu; Im, Kihak

2014-01-01

In the present study, thermal-hydraulic analyses for the blanket concept are being conducted using the Multidimensional Analysis of Reactor Safety (MARSKS) code, which has been used for the safety analysis of a pressurized water reactor. The purposes of the analyses are to verify the applicability of the code for the proposed blanket system, to investigate the departure of nucleate boiling (DNB) occurrence during the normal and transient conditions, and to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. In this paper, the thermal analysis results of the proposed blanket design using the MARS-KS code are presented for the normal operation and an accident condition of a reduced coolant flow rate. Afterwards, the plan for the whole blanket system analysis using MARSKS is introduced and the result of the first trial for the multiple blanket module analysis is summarized. In the present study, thermal-hydraulic analyses for the blanket concept were conducted using the MARS-KS code for a single blanket module. By comparing the MARS calculation results with the CFD analysis results, it was found that MARS-KS can be applied for the blanket thermal analysis with less number of computational meshes. Moreover, due to its capability on the two-phase flow analysis, it can be used for the transient or accident simulation where a phase change may be resulted in. In the future, the MARS-KS code will be applied for the anticipated transient and design based accident analyses. The investigation of the DNB occurrence during the normal and transient conditions will be of special interest of the analysis using it. After that, a methodology to simulate the entire blanket system was proposed by using the DLL version of MARS-KS. A supervisor program, which controls the multiple DLL files, was developed for the common header modelling. The program explicitly determines the flow rates of each module which can equalize

Status of the Negative Ion Based Heating and Diagnostic Neutral Beams for ITER

Science.gov (United States)

Schunke, B.; Bora, D.; Hemsworth, R.; Tanga, A.

2009-03-01

The current baseline of ITER foresees 2 Heating Neutral Beam (HNB's) systems based on negative ion technology, each accelerating to 1 MeV 40 A of D- and capable of delivering 16.5 MW of D0 to the ITER plasma, with a 3rd HNB injector foreseen as an upgrade option [1]. In addition a dedicated Diagnostic Neutral Beam (DNB) accelerating 60 A of H- to 100 keV will inject ≈15 A equivalent of H0 for charge exchange recombination spectroscopy and other diagnostics. Recently the RF driven negative ion source developed by IPP Garching has replaced the filamented ion source as the reference ITER design. The RF source developed at IPP, which is approximately a quarter scale of the source needed for ITER, is expected to have reduced caesium consumption compared to the filamented arc driven ion source. The RF driven source has demonstrated adequate accelerated D- and H- current densities as well as long-pulse operation [2, 3]. It is foreseen that the HNB's and the DNB will use the same negative ion source. Experiments with a half ITER-size ion source are on-going at IPP and the operation of a full-scale ion source will be demonstrated, at full power and pulse length, in the dedicated Ion Source Test Bed (ISTF), which will be part of the Neutral Beam Test Facility (NBTF), in Padua, Italy. This facility will carry out the necessary R&D for the HNB's for ITER and demonstrate operation of the full-scale HNB beamline. An overview of the current status of the neutral beam (NB) systems and the chosen configuration will be given and the ongoing integration effort into the ITER plant will be highlighted. It will be demonstrated how installation and maintenance logistics have influenced the design, notably the top access scheme facilitating access for maintenance and installation. The impact of the ITER Design Review and recent design change requests (DCRs) will be briefly discussed, including start-up and commissioning issues. The low current hydrogen phase now envisaged for start

Status of the Negative Ion Based Heating and Diagnostic Neutral Beams for ITER

International Nuclear Information System (INIS)

Schunke, B.; Bora, D.; Hemsworth, R.; Tanga, A.

2009-01-01

The current baseline of ITER foresees 2 Heating Neutral Beam (HNB's) systems based on negative ion technology, each accelerating to 1 MeV 40 A of D - and capable of delivering 16.5 MW of D 0 to the ITER plasma, with a 3rd HNB injector foreseen as an upgrade option. In addition a dedicated Diagnostic Neutral Beam (DNB) accelerating 60 A of H - to 100 keV will inject ≅15 A equivalent of H 0 for charge exchange recombination spectroscopy and other diagnostics. Recently the RF driven negative ion source developed by IPP Garching has replaced the filamented ion source as the reference ITER design. The RF source developed at IPP, which is approximately a quarter scale of the source needed for ITER, is expected to have reduced caesium consumption compared to the filamented arc driven ion source. The RF driven source has demonstrated adequate accelerated D - and H - current densities as well as long-pulse operation. It is foreseen that the HNB's and the DNB will use the same negative ion source. Experiments with a half ITER-size ion source are on-going at IPP and the operation of a full-scale ion source will be demonstrated, at full power and pulse length, in the dedicated Ion Source Test Bed (ISTF), which will be part of the Neutral Beam Test Facility (NBTF), in Padua, Italy. This facility will carry out the necessary R and D for the HNB's for ITER and demonstrate operation of the full-scale HNB beamline. An overview of the current status of the neutral beam (NB) systems and the chosen configuration will be given and the ongoing integration effort into the ITER plant will be highlighted. It will be demonstrated how installation and maintenance logistics have influenced the design, notably the top access scheme facilitating access for maintenance and installation. The impact of the ITER Design Review and recent design change requests (DCRs) will be briefly discussed, including start-up and commissioning issues. The low current hydrogen phase now envisaged for start

Supported liquid membrane stability in chiral resolution by chemically and physically modified membranes

Energy Technology Data Exchange (ETDEWEB)

Molinari, R.; Argurio, P. [Arcavata di Rende Univ. of Calabria, Arcavata di Rende, CS (Italy). Dept. of Chemical and Materials Engineering

2001-04-01

In the present work some stability studies on Supported Liquid Membranes (SLMs) to be used for chiral separations were realized. In particular, primary aim was to determine how a modification of the support surface influences the SLM stability. First, the procedure for support modification was optimised, making a screening of various compounds (sulphuric acid, nitric acid, chromic acid, sodium dodecyl sulphate (SDS), glycerol, oleic alcohol, propylene glycol (PPG), bovine serum albumin (BSA)) and testing their performance by means of contact angle measurements. Next, a second screening was realized by permeation tests in a stirred cell. Finally, to compare the stability of modified with unmodified support in a process of interest for chemical and/or biochemical industries, some permeation tests for resolution of DNB-DL-Leucine were realized in a re-circulation system. Results showed a better surface hydrophilization of chemically modified support and better stability of the sulphonated support. However, in operating conditions a little high stability of the unmodified support was obtained. [Italian] Nel presente lavoro sono stati realizzati degli studi di stabilita' di Membrane Liquide Supportate (SLMs) da impiegare in separazioni chirali. In particolare, obiettivo principale e' stato quello di determinare l'influenza che una modifica della superficie del supporto ha sulla stabilita' della SLM. Cosi', in un primo momento, e' stata ottimizzata le procedura di modifica del supporto, facendo una selezione tra vari composti (acido solforico, acido nitrico, acido cromico, sodio dodecil solfato (SDS), glicerolo, alcool oleico, glicole propilenico (PPG), siero di albumina bovina (BSA)) basata su misure dell'angolo di contatto. Successivamente, e' stata realizzata una seconda selezione mediante prove di permeazione in una cella agitata. Infine, con lo scopo di confrontare la stabilita' della SLM con supporto modificato rispetto

C-E setpoint methodology. C-E local power density and DNB LSSS and LCO setpoint methodology for analog protection systems

International Nuclear Information System (INIS)

1976-04-01

A description is presented of the methodology presently in use by Combustion Engineering to calculate Limiting Safety System Setting (LSSS) for the Local Power Density and Thermal Margin Trip Systems and Limiting Conditions for Operation (LCO) to assure that the specified acceptable fuel design limits are not exceeded during the design basis anticipated operational occurrences. The C-E Nuclear Steam Supply Systems for which the report is applicable are those incorporating the analog reactor protection system and licensed under the requirements of 10CFR50, Appendix A. The design basis events to be accommodated by the subject LSSS and LCO are discussed, and the methods to assure the required protection system response and initial required margin are described. The calculational techniques used to represent the specified acceptable fuel design limits in terms of monitored reactor parameters are provided. Using the resultant limits as a base, the methodology to synthesize the subject LSSS and LCO in terms of the parameters processed by the protection and monitoring systems is described

Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

Energy Technology Data Exchange (ETDEWEB)

Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

1985-07-01

This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

CASL VMA FY16 Milestone Report (L3:VMA.VUQ.P13.07) Westinghouse Mixing with COBRA-TF

Energy Technology Data Exchange (ETDEWEB)

Gordon, Natalie [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-09-30

COBRA-TF (CTF) is a low-resolution code currently maintained as CASL's subchannel analysis tool. CTF operates as a two-phase, compressible code over a mesh comprised of subchannels and axial discretized nodes. In part because CTF is a low-resolution code, simulation run time is not computationally expensive, only on the order of minutes. Hi-resolution codes such as STAR-CCM+ can be used to train lower-fidelity codes such as CTF. Unlike STAR-CCM+, CTF has no turbulence model, only a two-phase turbulent mixing coefficient, β. β can be set to a constant value or calculated in terms of Reynolds number using an empirical correlation. Results from STAR-CCM+ can be used to inform the appropriate value of β. Once β is calibrated, CTF runs can be an inexpensive alternative to costly STAR-CCM+ runs for scoping analyses. Based on the results of CTF runs, STAR-CCM+ can be run for specific parameters of interest. CASL areas of application are CIPS for single phase analysis and DNB-CTF for two-phase analysis.

Developments of fuel performance analysis codes in KEPCO NF

International Nuclear Information System (INIS)

Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S.

2012-01-01

The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF

A parametric thermohydraulic study an advanced pressurized light water reactor with a tight fuel rod lattice

International Nuclear Information System (INIS)

Dalle Donne, M.; Hame, W.

1982-12-01

A parametric thermohydraulic study for an Advanced Pressurized Light Water Reactor (APWR) with a tight fuel rod lattice has been performed. The APWR improves the uranium utilisation. The APWR core should be placed in a modern German PWR plant. Within this study about 200 different reactors have been calculated. The tightening of the fuel rod lattice implies a decrease of the net electrical output of the plant, which is greater for the heterogeneous reactor than for the homogeneous reactor. APWR cores mean higher core pressure drops and higher water velocities in the core region. The cores tend to be shorter and the number of fuel rods to be higher than for the PWR. At the higher fuel rod pitch to diameter ratios (p/d) the DNB limitation is more stringent than the limitation on the fuel rod linear rating given by the necessity of reflooding after a reactor accident. The contrary is true for the lower p/d ratios. Subcooled boiling in the highest rated coolant channels occurs for the most of the calculated reactors. (orig.) [de

Preliminary Study of the Onset of Nucleate Boiling (ONB) for the Thermal-hydraulic Design of HANARO Irradiation non-instrumented Capsule during the Natural Convection

Energy Technology Data Exchange (ETDEWEB)

Nam, Kyungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The HANARO reactor is an open-tank-in-pool type for easy access, and the capsules are being utilized for the irradiation test of materials and nuclear fuel in HANARO. The concept of the capsule is the direct contact with the coolant to cool the temperature of specimen down. To successfully accomplish the irradiation test, it is essential that the capsule should be designed considering the thermal margin such as the margin to Onset of Nucleate Boiling (ONB), the margin to Departure from Nucleate Boiling (DNB). In this paper, the preliminary study was performed by focusing on the ONB and the capsule design will be performed using the heat flux and temperature at ONB condition calculated in this paper. In this paper, the temperature and heat flux under ONB condition are simply calculated for the thermal design of fuel capsule for irradiation test. These values will be considered to design the non-instrumented capsule for natural circulation. To confirm the calculated value, detailed calculation will be performed using the one dimensional and multi-dimensional codes.

Moment inertia pump analysis used in the Rsg-Gas primary coolant loop under lofa condition

International Nuclear Information System (INIS)

Sudarmono; Setiyanto; Dhandhang, P.; Dibyo, S.; Royadi

1998-01-01

The moment inertia of primary cooling system analysis under LOFA condition has been done. It is potentially one of limiting design constraints of the RSG-GAS safety because the coolant flow rate reduces very rapidly under LOFA condition due to the low inertia circulation pumps. If a loss of flow accident occurs, the mass flow will decrease rapidly and the heat transfer coefficient between cladding and coolant will also decreases. As a consequence the fuel and cladding temperature will increase. The whole core was represented by the 1/4 sector and divided into 19 subchannels and 40 axial nodes. In the present study, moment inertia of pump analysis for RSG-GAS reactor was performed with COBRA-IV-I subchannel code. As the DNB correlation, W-3 Correlation was selected for base case. The flow and power transients under pump trip accident were determined from experiments. The result above compared with the design data are 75 kg m 2 and 81 Kg m 2 respectively. The result shows that the RSG-GAS requires the inertia more than 75 kg m 2

Vibration characteristics of a vertical round tube according to heat transfer regimes

International Nuclear Information System (INIS)

Lee, Yong Ho; Kim, Dae Hun; Chang, Soon Heung; Baek, Won Pil

2001-01-01

This paper presents the results of an experimental work on the effects of boiling heat transfer regimes on the vibration. the experiment has been performed using an electrically heated veritcal round tube through which water flows at atmospheric pressure. Vibration characteristics of the heated tube are changed significantly by heat transfer regimes and flow patterns. For single-phase liquid convection, the rod vibrations are negligible. However, On the beginning of subcooled nucleate boiling at tube exit, vibration level becomes very large. As bubble departure is occurred at the nucleation site of heated surface, the vibration decrease to saturated boiling region where thermal equilibrium quality becomes 0.0 at tube exit. In saturated boiling region, vibration amplitude increase with exit quality up to certain maximum value then decreases. At liquid film dryout condition, vibration could be regarded as negligible, however, these results cannot be extended to DNB-type CHF mechanism. Frequency analysis results of vibration signals suggested that excitation sources be different with heat transfer regimes. This study would contribute to improve the understanding of the relationship between boiling heat transfer and FIV

CFD application to advanced design for high efficiency spacer grid

International Nuclear Information System (INIS)

Ikeda, Kazuo

2014-01-01

Highlights: • A new LDV was developed to investigate the local velocity in a rod bundle and inside a spacer grid. • The design information that utilizes for high efficiency spacer grid has been obtained. • CFD methodology that predicts flow field in a PWR fuel has been developed. • The high efficiency spacer grid was designed using the CFD methodology. - Abstract: Pressurized water reactor (PWR) fuels have been developed to meet the needs of the market. A spacer grid is a key component to improve thermal hydraulic performance of a PWR fuel assembly. Mixing structures (vanes) of a spacer grid promote coolant mixing and enhance heat removal from fuel rods. A larger mixing vane would improve mixing effect, which would increase the departure from nucleate boiling (DNB) benefit for fuel. However, the increased pressure loss at large mixing vanes would reduce the coolant flow at the mixed fuel core, which would reduce the DNB margin. The solution is to develop a spacer grid whose pressure loss is equal to or less than the current spacer grid and that has higher critical heat flux (CHF) performance. For this reason, a requirement of design tool for predicting the pressure loss and CHF performance of spacer grids has been increased. The author and co-workers have been worked for development of high efficiency spacer grid using Computational Fluid Dynamics (CFD) for nearly 20 years. A new laser Doppler velocimetry (LDV), which is miniaturized with fiber optics embedded in a fuel cladding, was developed to investigate the local velocity profile in a rod bundle and inside a spacer grid. The rod-embedded fiber LDV (rod LDV) can be inserted in an arbitrary grid cell instead of a fuel rod, and has the advantage of not disturbing the flow field since it is the same shape as a fuel rod. The probe volume of the rod LDV is small enough to measure spatial velocity profile in a rod gap and inside a spacer grid. According to benchmark experiments such as flow velocity

CFD application to advanced design for high efficiency spacer grid

Energy Technology Data Exchange (ETDEWEB)

Ikeda, Kazuo, E-mail: kazuo3_ikeda@ndc.mhi.co.jp

2014-11-15

Highlights: • A new LDV was developed to investigate the local velocity in a rod bundle and inside a spacer grid. • The design information that utilizes for high efficiency spacer grid has been obtained. • CFD methodology that predicts flow field in a PWR fuel has been developed. • The high efficiency spacer grid was designed using the CFD methodology. - Abstract: Pressurized water reactor (PWR) fuels have been developed to meet the needs of the market. A spacer grid is a key component to improve thermal hydraulic performance of a PWR fuel assembly. Mixing structures (vanes) of a spacer grid promote coolant mixing and enhance heat removal from fuel rods. A larger mixing vane would improve mixing effect, which would increase the departure from nucleate boiling (DNB) benefit for fuel. However, the increased pressure loss at large mixing vanes would reduce the coolant flow at the mixed fuel core, which would reduce the DNB margin. The solution is to develop a spacer grid whose pressure loss is equal to or less than the current spacer grid and that has higher critical heat flux (CHF) performance. For this reason, a requirement of design tool for predicting the pressure loss and CHF performance of spacer grids has been increased. The author and co-workers have been worked for development of high efficiency spacer grid using Computational Fluid Dynamics (CFD) for nearly 20 years. A new laser Doppler velocimetry (LDV), which is miniaturized with fiber optics embedded in a fuel cladding, was developed to investigate the local velocity profile in a rod bundle and inside a spacer grid. The rod-embedded fiber LDV (rod LDV) can be inserted in an arbitrary grid cell instead of a fuel rod, and has the advantage of not disturbing the flow field since it is the same shape as a fuel rod. The probe volume of the rod LDV is small enough to measure spatial velocity profile in a rod gap and inside a spacer grid. According to benchmark experiments such as flow velocity

Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

Science.gov (United States)

Shirvan, Koroush

temperature was kept the same for the BWR-HD and ABWR which resulted in 4 °K cooler core inlet temperature for the BWR-HD given that its feedwater makes up a larger fraction of total core flow. The stability analysis using the STAB and S3K codes showed satisfactory results for the hot channel, coupled regional out-of-phase and coupled core-wide in-phase modes. A RELAPS model of the ABWR system was constructed and applied to six transients for the BWR-HD and ABWR. The 6MCPRs during all the transients were found to be equal or less for the new design and the core remained covered for both. The lower void coefficient along with smaller core volume proved to be advantages for the simulated transients. Helical Cruciform Fuel (HCF) rods were proposed in prior MIT studies to enhance the fuel surface to volume ratio. In this work, higher fidelity models (e.g. CFD instead of subchannel methods for the hydraulic behaviour) are used to investigate the resolution needed for accurate assessment of the HCF design. For neutronics, conserving the fuel area of cylindrical rods results in a different reactivity level with a lower void coefficient for the HCF design. In single-phase flow, for which experimental results existed, the friction factor is found to be sensitive to HCF geometry and cannot be calculated using current empirical models. A new approach for analysis of flow crisis conditions for HCF rods in the context of Departure from Nucleate Boiling (DNB) and dryout using the two phase interface tracking method was proposed and initial results are presented. It is shown that the twist of the HCF rods promotes detachment of a vapour bubble along the elbows which indicates no possibility for an early DNB for the HCF rods and in fact a potential for a higher DNB heat flux. Under annular flow conditions, it was found that the twist suppressed the liquid film thickness on the HCF rods, at the locations of the highest heat flux, which increases the possibility of reaching early dryout. It

Cross Matching of VIIRS Boat Detection and Vessel Monitoring System Tracks

Science.gov (United States)

Hsu, F. C.; Elvidge, C.; Zhizhin, M. N.; Baugh, K.; Ghosh, T.

2016-12-01

One approach to commercial fishing is to use use bright lights at night to attract catch. This is a widely used practice in East and Southeast Asia, but can also be found in other fisheries. In some cases, the deployed lighting exceeds 100,000 watts. Such lighting is distinctive in dark ocean and can even be seen from space with sensor such as Visible Infrared Imaging Radiometer Suite Day/Night Band (VIIRS-DNB). We have developed a VIIRS Boat Detection (VBD) system, which outputs lists of boat locations in near real time. One of the standard methods fishery agencies use to collect geospatial data on fishing boats is to require boats to carry Vessel Monitoring System beacons. We developed an algorithm to cross-match VBD data with VMS tracks. With this we are able to identify fishing boats that do not carry VMS beacons. In certain situations, this is an indicator of illegal fishing. The other application for this cross-matching is to define the VIIRS detection limits and developing a calibration to estimate deployed wattage. Here we demonstrate results of cross matching VBD and VMS for Indonesia as example to showcase its potential.

A heuristic application of critical power ratio to pressurized water reactor core design

International Nuclear Information System (INIS)

Ahn, Seung Hoon; Jeun, Gyoo Dong

2002-01-01

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR correlation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large

Experiment data report for semiscale Mod-1 test S-02-3 (blowdown heat transfer test)

International Nuclear Information System (INIS)

Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

1975-09-01

Recorded test data are presented for Test S-02-3 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-3 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,263 psig. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with a moderately heated core (75 percent design power level). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was also set at the 75 percent design level to achieve full core temperature differential. The flow resistance of the intact loop was based on core area scaling. During system depressurization, core power was reduced from the initial level of 1.2 MW in such a manner as to simulate the surface heat flux response of the LOFT nuclear fuel rods until such time that departure from nucleate boiling (DNB) occurs. Blowdown to the pressure suppression system was accomplished without simulated emergency core coolant injection or pressure suppression system coolant spray

Socio-economic impact of Trans-Siberian railway after the collapse of Soviet Union by integrated spatial data analysis

Science.gov (United States)

Uchida, Seina; Takeuchi, Wataru; Hatoyama, Kiichiro; Mazurov, Yuri

2016-06-01

How Russian cities have stood up again after the collapse of Soviet Union will be discussed in this paper. In order to know how the cities has managed the difficult period after the change of social system, transition of urban area, population, and nighttime light is searched. Although Far East will not stop as one of the most important area with abundant resources, overpopulation in towns and depopulation in countryside is going on. By searching the present situation, this research also aims to predict the future of Far East and Russia. First of all, Landsat data from 1987 to 2015 is collected over Moscow, Vladivostok, Novosibirsk, Tynda, and Blagoveshchensk and urban area is calculated by land cover classification. Secondly, population and retail turnover data are collected from year books in Russia. Thirdly, gross regional product (GRP) is estimated by nighttime light images from DMSP-OLS and VIIRS DNB dataset. In addition, these data are compared and difference of development stage after the collapse of Soviet Union between the unstable era (1990s-2000) and development era (2000-) will be discussed. It is expected that these analysis will give us useful information about Russian strategy for the future.

Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

International Nuclear Information System (INIS)

Bokhari, I.H.; Israr, M.; Pervez, S.

1999-01-01

Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

DNBR Prediction Using a Support Vector Regression

International Nuclear Information System (INIS)

Yang, Heon Young; Na, Man Gyun

2008-01-01

PWRs (Pressurized Water Reactors) generally operate in the nucleate boiling state. However, the conversion of nucleate boiling into film boiling with conspicuously reduced heat transfer induces a boiling crisis that may cause the fuel clad melting in the long run. This type of boiling crisis is called Departure from Nucleate Boiling (DNB) phenomena. Because the prediction of minimum DNBR in a reactor core is very important to prevent the boiling crisis such as clad melting, a lot of research has been conducted to predict DNBR values. The object of this research is to predict minimum DNBR applying support vector regression (SVR) by using the measured signals of a reactor coolant system (RCS). The SVR has extensively and successfully been applied to nonlinear function approximation like the proposed problem for estimating DNBR values that will be a function of various input variables such as reactor power, reactor pressure, core mass flowrate, control rod positions and so on. The minimum DNBR in a reactor core is predicted using these various operating condition data as the inputs to the SVR. The minimum DBNR values predicted by the SVR confirm its correctness compared with COLSS values

Critical power prediction by CATHARE2 of the OECD/NRC BFBT benchmark

Energy Technology Data Exchange (ETDEWEB)

Lutsanych, Sergii, E-mail: s.lutsanych@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122, San Piero a Grado, Pisa (Italy); Sabotinov, Luben, E-mail: luben.sabotinov@irsn.fr [Institut for Radiological Protection and Nuclear Safety (IRSN), 31 avenue de la Division Leclerc, 92262 Fontenay-aux-Roses (France); D’Auria, Francesco, E-mail: francesco.dauria@dimnp.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122, San Piero a Grado, Pisa (Italy)

2015-03-15

Highlights: • We used CATHARE code to calculate the critical power exercises of the OECD/NRC BFBT benchmark. • We considered both steady-state and transient critical power tests of the benchmark. • We used both the 1D and 3D features of the CATHARE code to simulate the experiments. • Acceptable prediction of the critical power and its location in the bundle is obtained using appropriate modelling. - Abstract: This paper presents an application of the French best estimate thermal-hydraulic code CATHARE 2 to calculate the critical power and departure from nucleate boiling (DNB) exercises of the International OECD/NRC BWR Fuel Bundle Test (BFBT) benchmark. The assessment activity is performed comparing the code calculation results with available in the framework of the benchmark experimental data from Japanese Nuclear Power Engineering Corporation (NUPEC). Two-phase flow calculations on prediction of the critical power have been carried out both in steady state and transient cases, using one-dimensional and three-dimensional modelling. Results of the steady-state critical power tests calculation have shown the ability of CATHARE code to predict reasonably the critical power and its location, using appropriate modelling.

[Determination of nitroaromatics and cyclo ketones in sea water' by gas chromatography coupled with activated carbon fiber solid-phase micro-extraction].

Science.gov (United States)

Ma, Hanna; Zhu, Mengya; Wang, Yalin; Sun, Tonghua; Jia, Jinping

2009-05-01

A gas chromatography (GC) coupled with solid-phase micro-extraction using a special activated carbon fiber (ACF) was developed for the analysis of 6 nitroaromatics and cyclic ketones, nitrobenzene (NB), 1,3-dinitrobenzene (1,3-DNB), 2,4-dinitrotoluene (2,4-DNT), 2,6-dinitrotoluene (2,6-DNT), isophorone, 1,4-naphthaquinone (1,4-NPQ), in sea water samples. The sample was extracted for 30 min under saturation of NaCl at 1,500 r/min and 60 degrees C in head space. The desorption was performance at 280 degrees C for 2 min. The linear ranges were from 0.01 to 400 microg/L. The limits of detection (LODs) were 1.4 - 3.2 ng/L. This method has been successfully applied to the determination of nitroaromatics and cyclic ketones in the sea water samples obtained from East China Sea. The concentrations of nitrobenzene, 1,3-dinitrobenzene and 2,6-dinitrotoluene in the sea water sample were 0.756, 0.944, 0.890 microg/L, respectively. The recoveries were 86.3% - 101.8% with the relative standard deviations (RSDs) of 3.7% -7.8%. The method is suitable for analyzing nitroaromatics and cyclic ketones at low concentration levels in sea water samples.

Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-15

For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

Silicon controlled rectifier polyphase bridge inverter commutated with gate-turn-off thyristor

Science.gov (United States)

Edwards, Dean B. (Inventor); Rippel, Wally E. (Inventor)

1986-01-01

A polyphase SCR inverter (10) having N switching poles, each comprised of two SCR switches (1A, 1B; 2A, 2B . . . NA, NB) and two diodes (D1B; D1B; D2A, D2B . . . DNA, DNB) in series opposition with saturable reactors (L1A, L1B; L2A, L2B . . . LNA, LNB) connecting the junctions between the SCR switches and diodes to an output terminal (1, 2 . . . 3) is commutated with only one GTO thyristor (16) connected between the common negative terminal of a dc source and a tap of a series inductor (14) connected to the positive terminal of the dc source. A clamp winding (22) and diode (24) are provided, as is a snubber (18) which may have its capacitance (c) sized for maximum load current divided into a plurality of capacitors (C.sub.1, C.sub.2 . . . C.sub.N), each in series with an SCR switch S.sub.1, S.sub.2 . . . S.sub.N). The total capacitance may be selected by activating selected switches as a function of load current. A resistor 28 and SCR switch 26 shunt reverse current when the load acts as a generator, such as a motor while braking.

Utilizing Suomi NPP's Day-Night Band to Assess Energy Consumption in Rural and Urban Areas as an Input for Poverty Analysis

Science.gov (United States)

Baldwin, H. B.; Klug, M.; Tapracharoen, K.; Visudchindaporn, C.

2017-12-01

While poverty in Thailand has decreased from 67% in 1986 to 13% in 2012, 6.7 million people were still living within 20% of the poverty line in 2014. Economic uncertainty caused by recurring droughts and decreasing agricultural prices puts this vulnerable part of the population at risk of dropping below the national poverty line in the future. In order to address this issue, the team worked with the Office of Science and Technology (OSTC) at the Royal Thai Embassy, Asian Disaster Preparedness Center (ADPC), and the NASA SERVIR Coordination Office to formulate a new method of analyzing poverty within Thailand. This project utilizes the monthly composite product for 2012-2015 produced by the Earth Observations Group (EOG) at National Oceanic and Atmospheric Administration (NOAA) and National Geophysical Data Center (NGDC). EOG created this product from satellite imagery from Suomi National Polar-Orbiting Visible Infrared Imaging Radiometer Suite's Day Night Band (Suomi NPP VIIRS DNB). Additionally, this project incorporated socio-economic data from Thailand's Ministry of Information and Communication Technology's National Statistical Office and Ministry of Education's National Education Information System to create an enhanced poverty index. This new poverty index will provide the Thai government a cost-effective way to analyze changes of poverty within the nation and inform policy making.

Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

International Nuclear Information System (INIS)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-01

For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

A PWR hot-rod model: Relap5/mod3.2.2.{gamma} as a subchannel code

Energy Technology Data Exchange (ETDEWEB)

Kirsten, I.C.; Jones, J.R. [British Energy, Barnwood, Gloucester (United Kingdom); Kimber, G.R. [Atomic Energy Authority Technology, Winfrith, Dorset (United Kingdom); Page, R. [National Nuclear Corp. Ltd., Cheshire (United Kingdom)

2001-07-01

The use of the PWR transient analysis code RELAP5 for detailed assessment of Departure from Nucleate Boiling (DNB) has previously implied coupling it in some way to a subchannel code, either by direct code-to-code coupling or by transferring core boundary conditions to the subchannel code. This paper shows an alternative by using a group of subchannels modelled in RELAP5 to represent a hot rod. The model consists of three parallel channels, each more refined than its neighbour: The first channel represents a quadrant of the core; the second a quadrant on a fuel assembly and the final channel represents a passage adjacent to a single fuel pin. The model is intended for use as part of point kinetics assessments and each channel is assigned a radial form factor designed to conservatively represent the hottest fuel pins in the reactor core. The main outputs from the model are minimum Departure from Nucleate Boiling Ratio (DNBR) and clad oxidation for the hot rod (lead pin). The DNBR results from the hot-rod model are benchmarked against the subchannel code COBRA 3-CP and the results are presented in this paper. Some of the modelling problems that needed to be resolved are also highlighted. (author)

SPoRT Participation in the GOES-R and JPSS Proving Grounds

Science.gov (United States)

Jedlovec, Gary; Fuell, Kevin; Smith, Matthew

2013-01-01

For the last several years, the NASA Short-term Prediction Research and Transition (SPoRT) project at has been working with the various algorithm working groups and science teams to demonstrate the utility of future operational sensors for GOES-R and the suite of instruments for the JPSS observing platforms. For GOES-R, imagery and products have been developed from polar-orbiting sensors such as MODIS and geostationary observations from SEVIRI, simulated imagery, enhanced products derived from existing GOES satellites, and data from ground-based observing systems to generate pseudo or proxy products for the ABI and GLM instruments. The suite of products include GOES-POES basic and RGB hybrid imagery, total lightning flash products, quantitative precipitation estimates, and convective initiation products. SPoRT is using imagery and products from VIIRS, CrIS, ATMS, and OMPS to show the utility of data and products from their operational counterparts on JPSS. The products include VIIRS imagery in swath form, the GOES-POES hybrid, a suite of RGB products including the air mass RGB using water vapor and ozone channels from CrIS, and several DNB products. Over a dozen SPoRT collaborative WFOs and several National Centers are involved in an intensive evaluation of the operational utility of these products.

Review of Available Data for Validation of Nuresim Two-Phase CFD Software Applied to CHF Investigations

Directory of Open Access Journals (Sweden)

D. Bestion

2009-01-01

Full Text Available The NURESIM Project of the 6th European Framework Program initiated the development of a new-generation common European Standard Software Platform for nuclear reactor simulation. The thermal-hydraulic subproject aims at improving the understanding and the predictive capabilities of the simulation tools for key two-phase flow thermal-hydraulic processes such as the critical heat flux (CHF. As part of a multi-scale analysis of reactor thermal-hydraulics, a two-phase CFD tool is developed to allow zooming on local processes. Current industrial methods for CHF mainly use the sub-channel analysis and empirical CHF correlations based on large scale experiments having the real geometry of a reactor assembly. Two-phase CFD is used here for understanding some boiling flow processes, for helping new fuel assembly design, and for developing better CHF predictions in both PWR and BWR. This paper presents a review of experimental data which can be used for validation of the two-phase CFD application to CHF investigations. The phenomenology of DNB and Dry-Out are detailed identifying all basic flow processes which require a specific modeling in CFD tool. The resulting modeling program of work is given and the current state-of-the-art of the modeling within the NURESIM project is presented.

Establishment and assessment of CHF data base for square-lattice rod bundles

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Seo, K. W.; Kim, K. K.; Zee, S. Q.

2002-02-01

A CHF data base is constructed for square-lattice rod bundles, and assessed with various existing CHF prediction models. The CHF data base consists of 10725 data points obtained from 147 test bundles with uniform axial power distributions and 29 test bundles with non-uniform axial power distributions. The local thermal-hydraulic conditions in the subchannels are calculated by employing a subchannel analysis code MATRA. The influence of turbulent mixing parameter on CHF is evaluated quantitatively for selected test bundles with representative cross sectional configurations. The performance of various CHF prediction models including empirical correlations for round tubes or rod bundles, theoretical DNB models such as sublayer dryout model and bubble crowding model, and CHF lookup table for round tubes, are assessed for the localized rod bundle CHF data base. In view of the analysis result, it reveals that the 1995 AECL-IPPE CHF lookup table method is one of promising models in the aspect of the prediction accuracy and the applicable range. As the result of analysis employing the CHF lookup table for 9113 data points with uniform axial heat profile, the mean and the standard deviation of P/M are calculated as 1.003 and 0.115 by HBM, 1.022 and 0.319 by DSM respectively

Initial Verification and Validation Assessment for VERA

Energy Technology Data Exchange (ETDEWEB)

Dinh, Nam [North Carolina State Univ., Raleigh, NC (United States); Athe, Paridhi [North Carolina State Univ., Raleigh, NC (United States); Jones, Christopher [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hetzler, Adam [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sieger, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-04-01

The Virtual Environment for Reactor Applications (VERA) code suite is assessed in terms of capability and credibility against the Consortium for Advanced Simulation of Light Water Reactors (CASL) Verification and Validation Plan (presented herein) in the context of three selected challenge problems: CRUD-Induced Power Shift (CIPS), Departure from Nucleate Boiling (DNB), and Pellet-Clad Interaction (PCI). Capability refers to evidence of required functionality for capturing phenomena of interest while capability refers to the evidence that provides confidence in the calculated results. For this assessment, each challenge problem defines a set of phenomenological requirements against which the VERA software is assessed. This approach, in turn, enables the focused assessment of only those capabilities relevant to the challenge problem. The evaluation of VERA against the challenge problem requirements represents a capability assessment. The mechanism for assessment is the Sandia-developed Predictive Capability Maturity Model (PCMM) that, for this assessment, evaluates VERA on 8 major criteria: (1) Representation and Geometric Fidelity, (2) Physics and Material Model Fidelity, (3) Software Quality Assurance and Engineering, (4) Code Verification, (5) Solution Verification, (6) Separate Effects Model Validation, (7) Integral Effects Model Validation, and (8) Uncertainty Quantification. For each attribute, a maturity score from zero to three is assigned in the context of each challenge problem. The evaluation of these eight elements constitutes the credibility assessment for VERA.

Three-dimensional transport coefficient model and prediction-correction numerical method for thermal margin analysis of PWR cores

International Nuclear Information System (INIS)

Chiu, C.

1981-01-01

Combustion Engineering Inc. designs its modern PWR reactor cores using open-core thermal-hydraulic methods where the mass, momentum and energy equations are solved in three dimensions (one axial and two lateral directions). The resultant fluid properties are used to compute the minimum Departure from Nuclear Boiling Ratio (DNBR) which ultimately sets the power capability of the core. The on-line digital monitoring and protection systems require a small fast-running algorithm of the design code. This paper presents two techniques used in the development of the on-line DNB algorithm. First, a three-dimensional transport coefficient model is introduced to radially group the flow subchannel into channels for the thermal-hydraulic fluid properties calculation. Conservation equations of mass, momentum and energy for this channels are derived using transport coefficients to modify the calculation of the radial transport of enthalpy and momentum. Second, a simplified, non-iterative numerical method, called the prediction-correction method, is applied together with the transport coefficient model to reduce the computer execution time in the determination of fluid properties. Comparison of the algorithm and the design thermal-hydraulic code shows agreement to within 0.65% equivalent power at a 95/95 confidence/probability level for all normal operating conditions of the PWR core. This algorithm accuracy is achieved with 1/800th of the computer processing time of its parent design code. (orig.)

Application of a statistical thermal design procedure to evaluate the PWR DNBR safety analysis limits

International Nuclear Information System (INIS)

Robeyns, J.; Parmentier, F.; Peeters, G.

2001-01-01

In the framework of safety analysis for the Belgian nuclear power plants and for the reload compatibility studies, Tractebel Energy Engineering (TEE) has developed, to define a 95/95 DNBR criterion, a statistical thermal design method based on the analytical full statistical approach: the Statistical Thermal Design Procedure (STDP). In that methodology, each DNBR value in the core assemblies is calculated with an adapted CHF (Critical Heat Flux) correlation implemented in the sub-channel code Cobra for core thermal hydraulic analysis. The uncertainties of the correlation are represented by the statistical parameters calculated from an experimental database. The main objective of a sub-channel analysis is to prove that in all class 1 and class 2 situations, the minimum DNBR (Departure from Nucleate Boiling Ratio) remains higher than the Safety Analysis Limit (SAL). The SAL value is calculated from the Statistical Design Limit (SDL) value adjusted with some penalties and deterministic factors. The search of a realistic value for the SDL is the objective of the statistical thermal design methods. In this report, we apply a full statistical approach to define the DNBR criterion or SDL (Statistical Design Limit) with the strict observance of the design criteria defined in the Standard Review Plan. The same statistical approach is used to define the expected number of rods experiencing DNB. (author)

Preliminary Study of ONB in Narrow-Vertical Rectangular Channel

International Nuclear Information System (INIS)

Omar, S. AL-Yahia; Jo, Daeseong

2015-01-01

The location where the vapor bubble can first exist at the heated surface is called 'onset of nucleate boiling (ONB). The subcooled boiling is highly efficient to remove the heat owing to the high heat transfer coefficient. The heat transfer is affected by the motion of the bulk liquid as well as the latent heat transport of the liquid microlayer between the bubble and the heated wall. However, with increasing in the wall temperature, the bubble growth will increase and may they aggregate at the heated surface forming a vapor film, which will prevent the heat transport from the wall and that leads to highly rise in wall temperature. This phenomenon called departure from nucleate boiling (DNB). Many experimental and numerical CFD methods were carried out to investigate the subcooled boiling because of its importance in the industrial applications. In the present study, vertical narrow rectangular channel heated from both side was simulated by using CFX-14 to investigate the subcooled wall boiling, and identical simulation is done by using TMAP to compare the ONB location between numerical simulation and empirical correlations that implemented in TMAP. The numerical results using CFX-14 are discussed and compared with the results obtained from TMAP. The coolant temperature increases gradually (linearly) in the downward direction owing to the uniform applied heat flux.

Preliminary Study of ONB in Narrow-Vertical Rectangular Channel

Energy Technology Data Exchange (ETDEWEB)

Omar, S. AL-Yahia; Jo, Daeseong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The location where the vapor bubble can first exist at the heated surface is called 'onset of nucleate boiling (ONB). The subcooled boiling is highly efficient to remove the heat owing to the high heat transfer coefficient. The heat transfer is affected by the motion of the bulk liquid as well as the latent heat transport of the liquid microlayer between the bubble and the heated wall. However, with increasing in the wall temperature, the bubble growth will increase and may they aggregate at the heated surface forming a vapor film, which will prevent the heat transport from the wall and that leads to highly rise in wall temperature. This phenomenon called departure from nucleate boiling (DNB). Many experimental and numerical CFD methods were carried out to investigate the subcooled boiling because of its importance in the industrial applications. In the present study, vertical narrow rectangular channel heated from both side was simulated by using CFX-14 to investigate the subcooled wall boiling, and identical simulation is done by using TMAP to compare the ONB location between numerical simulation and empirical correlations that implemented in TMAP. The numerical results using CFX-14 are discussed and compared with the results obtained from TMAP. The coolant temperature increases gradually (linearly) in the downward direction owing to the uniform applied heat flux.

Magnetic analysis of the magnetic field reduction system of the ITER neutral beam injector

Energy Technology Data Exchange (ETDEWEB)

Barrera, Germán, E-mail: german.barrera@ciemat.es [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Ahedo, Begoña; Alonso, Javier; Ríos, Luis [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Chareyre, Julien; El-Ouazzani, Anass [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 07/08, 08019 Barcelona (Spain)

2015-10-15

The neutral beam system for ITER consists of two heating and current drive neutral beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. For the correct operation of the beam, the ion source and the ion path until it is neutralized must operate under a very low magnetic field environment. To prevent the stray ITER field from penetrating inside those mentioned critical areas, a magnetic field reduction system (MFRS) will envelop the beam vessels and the high voltage transmission lines to ion source. This system comprises the passive magnetic shield (PMS), a box like assembly of thick low carbon steel plates, and the Active Correction and Compensation Coils (ACCC), a set of coils carrying a current which depends on the tokamak stray field. This paper describes the magnetic model and analysis results presented at the PMS and ACCC preliminary design review held in ITER organization in April 2013. The paper focuses on the magnetic model description and on the description of the analysis results. The iterative process for obtaining optimized currents in the coils is presented. The set of coils currents chosen among the many possible solutions, the magnetic field results in the interest regions and the fulfillment of the magnetic field requirements are described.

Investigation of thick-target neutron emission from Be-9(d,n)B-10 at E/sub d/ = 7 MeV for angles other than zero degrees

International Nuclear Information System (INIS)

Smith, D.L.; Meadows, J.W.; Guenther, P.T.

1985-01-01

Double-differential measurements of neutron emission from a thick beryllium target bombarded with 7-MeV deuterons are made for neutrons above 800 keV, over the angular range of 0 to 155 0 . The angular dependence of the neutron yield is found to be quite anisotropic. The importance of this anisotropy in integral neutron-induced reaction cross-section investigations is illustrated. 7 refs.,

Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

International Nuclear Information System (INIS)

Boone, M. L.; King, S. J.; Pulver, E. F.; Jeon, K.-L.; Esteves, R.; Kurincic, B.

2004-01-01

Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLO TM fuel rods), neutronic efficient components (i.e. ZIRLO TM Mid grids), ZIRLO TM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly

Control rod drop accident analysis for the mixed core project in Ling Ao NPS

International Nuclear Information System (INIS)

Zhang Shishun; Zhou Zhou; Xiao Min

2004-01-01

AFA-2G assemblies in Ling Ao NPS (LNPS) have been replaced gradually by AFA-3G assemblies from cycle 2 and subsequent cycles. the enrichment of the fuels will be increased from 3.2% to 3.7% from cycle 3 in Ling Ao. Therefore, the study of ling Ao mixed core and increased enrichment have been performed since 2001. Lots of accidents need to be re-analyzed in Ling Ao NPS in order to verify its safety requirements for the new fuel management. Control rod drop accident for LNPS was re-analyzed in 2001 in frame of FRAMATOME ANP analytical methodology. The analytical codes used in the accident analysis include SCIENCE, ESPADON, CINEMA, CANTAL and FLICA III. The control rod drop accident analysis is performed with respect to the 10 reference cycles of the generic fuel management design for Ling Ao mixed core and increased enrichment study. The pre-drop FδH for the first transition cycles and other cycles are 1.52 and 1.55, respectively. For detected dropped rod configurations, the negative flux rate protection system actuates a reactor trip. For the non-detected dropped rod configurations, the minimum DNBR values have been evaluated with conservative analysis methodology and assumptions and the DNBR fuel design limit is respected the analytical results shows that, for all the non-detected dropped rod configurations, the minimum DNB margin is about 2% which occurs in AFA-2G fuel assembly in the first transition cycle. (author)

One-dimensional two-phase thermal hydraulics (ENSTA course)

International Nuclear Information System (INIS)

Olive, J.

1995-11-01

This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends

One-dimensional two-phase thermal hydraulics (ENSTA course); Thermo-hydraulique diphasique monodimensionnelle. Cours ENSTA

Energy Technology Data Exchange (ETDEWEB)

Olive, J

1995-11-01

This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends.

Treatment of asthma: Identification of the practice behavior and the deviation from the guideline recommendations

Directory of Open Access Journals (Sweden)

Bhattacharyya Parthasarathi

2010-01-01

Full Text Available Background: Despite an exponential development of the understanding of the disease with availability of good therapy and feasibility of good control along with availability of globally accepted guidelines, there remains a significant gap between the guidelines and prevailing practice behavior for treating asthma all over the globe. This perhaps stands as the single most deterrent factor for good asthma care worldwide. The objective of the study is to analyze the asthma prescriptions to find out the available status of the practice behaviour and the deviations from the guideline in asthma practice. Materials and Methods: The asthma prescriptions of the referred patients presenting to the OPD services of the IPCR, Kolkata were photocopied and collected. They were further analyzed based on the available information upon a format being prepared on four major areas as qualifications, clinical recording habit, practice of evaluating patients, and treatment habit that stands apparent from the prescribed medications. The doctors were divided into three categories as a MBBS, b MD/DNB (medicine and respiratory medicine, and c DM (non respiratory sub-specialities and statistical analysis has been performed comparing the three groups as per the performance in the four pre-decided areas. Results: All the groups fall short of any guideline or text of asthma care in all the areas involved. Conclusion: The practice behaviour of our doctors for asthma care appears deficient in several areas and seems far from guideline recommendations. This needs further evaluation and adoption of appropriate interventions.

Computational multi-fluid dynamics predictions of critical heat flux in boiling flow

Energy Technology Data Exchange (ETDEWEB)

Mimouni, S., E-mail: stephane.mimouni@edf.fr; Baudry, C.; Guingo, M.; Lavieville, J.; Merigoux, N.; Mechitoua, N.

2016-04-01

Highlights: • A new mechanistic model dedicated to DNB has been implemented in the Neptune-CFD code. • The model has been validated against 150 tests. • Neptune-CFD code is a CFD tool dedicated to boiling flows. - Abstract: Extensive efforts have been made in the last five decades to evaluate the boiling heat transfer coefficient and the critical heat flux in particular. Boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. As a consequence, models dedicated to boiling flows have being improved. For example, Reynolds Stress Transport Model, polydispersion and two-phase flow wall law have been recently implemented. In a previous work, we have evaluated computational fluid dynamics results against single-phase liquid water tests equipped with a mixing vane and against two-phase boiling cases. The objective of this paper is to propose a new mechanistic model in a computational multi-fluid dynamics tool leading to wall temperature excursion and onset of boiling crisis. Critical heat flux is calculated against 150 tests and the mean relative error between calculations and experimental values is equal to 8.3%. The model tested covers a large physics scope in terms of mass flux, pressure, quality and channel diameter. Water and R12 refrigerant fluid are considered. Furthermore, it was found that the sensitivity to the grid refinement was acceptable.

Computational multi-fluid dynamics predictions of critical heat flux in boiling flow

International Nuclear Information System (INIS)

Mimouni, S.; Baudry, C.; Guingo, M.; Lavieville, J.; Merigoux, N.; Mechitoua, N.

2016-01-01

Highlights: • A new mechanistic model dedicated to DNB has been implemented in the Neptune_CFD code. • The model has been validated against 150 tests. • Neptune_CFD code is a CFD tool dedicated to boiling flows. - Abstract: Extensive efforts have been made in the last five decades to evaluate the boiling heat transfer coefficient and the critical heat flux in particular. Boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. As a consequence, models dedicated to boiling flows have being improved. For example, Reynolds Stress Transport Model, polydispersion and two-phase flow wall law have been recently implemented. In a previous work, we have evaluated computational fluid dynamics results against single-phase liquid water tests equipped with a mixing vane and against two-phase boiling cases. The objective of this paper is to propose a new mechanistic model in a computational multi-fluid dynamics tool leading to wall temperature excursion and onset of boiling crisis. Critical heat flux is calculated against 150 tests and the mean relative error between calculations and experimental values is equal to 8.3%. The model tested covers a large physics scope in terms of mass flux, pressure, quality and channel diameter. Water and R12 refrigerant fluid are considered. Furthermore, it was found that the sensitivity to the grid refinement was acceptable.

Study of transient connected with WWER-1000 cluster drop with subsequent working of automatic power controller

International Nuclear Information System (INIS)

Kuchin, A.; Ovdiienko, I.; Khalimonchuk, V.

2010-01-01

Results of calculation study of transient connected with drop of WWER-1000 cluster of working group are presented. Transient was considered in the mode of automatic power control without forming of warning protection signal due to reaching of dropped cluster of core bottom. Calculations are shown that given transient can cause valuable distortion of power distribution in axial direction. At that main increase of pin power is occurred in upper part of the core, whereas power in lower part is almost not changed. The additional increase of power in the upper part of core makes conditions for initiation of DNB. This effect can be observed if in initial state axial power distribution is displaced in upper part of core nearby to rest of supported power clusters of working group. It is necessary to define conservatively with taking into account assumed working group efficiency-in which row from extracted clusters of working group the displacement of axial power in the upper part is possible. Probability of such displacement and its localization in plane of core must be properly analyzed. The work was performed in framework of orders BMU SR 2511 and BMU R0801504 (SR2611). The report describes the opinion and view of the contractor-State Scientific and Technical Centre on Nuclear and Radiation Safety-and does not necessarily represent the opinion of the ordering party - BMU-BfS/GRS and TUEV SUED. (Authors)

The PARET code and the analysis of the SPERT I transients

Energy Technology Data Exchange (ETDEWEB)

Woodruff, William L [Argonne National Laboratory, Argonne (United States)

1983-09-01

The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients.

Experiment data report for Semiscale Mod-1 Test S-29-1 (integral test with asymmetrical break)

International Nuclear Information System (INIS)

Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

1976-07-01

Recorded test data are presented for Test S-29-1 of the Semiscale Mod-1 special heat transfer test series. This test is among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident (LOCA) in a pressurized-water reactor system. Test S-29-1 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,260 psia. An asymmetrical offset shear cold leg break was used to investigate the system response to a depressurization transient with a flow distribution different from that associated with a symmetrical cold leg break. System flow was set to achieve a core fluid temperature differential of 66 0 F at full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling (DNB) might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

Feasibility Study on Dual-Cooled Annular Fuel for OPR-1000 Power Uprate

International Nuclear Information System (INIS)

Chun, Tae Hyun; In, Wang Kee; Oh, Dong Suck

2010-04-01

A dual-cooled annular fuel (DCAF) is a highly promising concept as a high power density fuel for PWR power-uprate. The purpose of this study is to assess a feasibility of 120% core power for OPR-1000 with the DCAF. So the feasibility study were done through the code establishments for annular fuel analysis, evaluations of core physics, thermal-hydraulics and safety analyses at a 120% power with OPR-1000 and the preliminary economic benefits of 20% power-uprate. As results of the analyses, DCAF at 120% power showed sufficient margins available on DNB, PCT and fuel pellet temperature relative to the solid fuel at 100% power. However, judging from an anticipated wide range of the gap conductance variation in inner and outer clearances as fuel burn-up in the reactor core, irradiation behavior of DCAF has to be observed through research reactor test. On the other hand, the nuclear physics parameters like moderator temperature coefficient, power coefficient and so on comply with the technical specifications. An impact of 20% power-uprate on NSSS and BOP was also investigated, and accordingly some components and parts need to be changed were identified. Moreover, the economical benefits from the power-uprate was roughly estimated. It turned out that the power-uprating with DCAF could give an enormous profit even considering the expenses of components and parts to be replaced, additional fuel cycle cost and extended overhaul period

The PARET code and the analysis of the SPERT I transients

International Nuclear Information System (INIS)

Woodruff, William L.

1983-01-01

The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients

Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

International Nuclear Information System (INIS)

Bandini, G.; Caira, M.; Naviglio, A.; Sorabella, L.

1995-01-01

The MARS nuclear plant is equipped with a 600 MWth PWR type nuclear steam supply system, with completely innovative engineered core safeguards. The most relevant innovative safety system of this plant is its Emergency Core Cooling System, which is completely passive (with only one non static component). The Emergency Core Cooling System (ECCS) of the MARS reactor is natural-circulation, passive-type, and its intervention follows a core flow decrease, whatever was the cause. The operation of the system is based on a cascade of three fluid systems, functionally interfacing through heat exchangers; the first fluid system is connected to the reactor vessel and the last one includes an atmospheric-pressure condenser, cooled by external air. The infinite thermal capacity of the final heat sink provides the system an unlimited autonomy. The capability and operability of the system are based on its integrity and on the integrity of the primary coolant boundary (both of them are permanently enclosed in a pressurized containment; 100% redundancy is also foreseen) and on the operation of only one non static component (a check valve), with 400% redundancy. In the paper, all main thermal hydraulic transients occurring as a consequence of postulated accidents are analysed, to verify the capability of the passive-type ECCS to intervene always in time, without causing undue conditions of reduced coolability of the core (DNB, etc.), and to verify its capability to guarantee a long-term (indefinite) coolability of the core without the need of any external intervention. (author)

Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

International Nuclear Information System (INIS)

Joshi, Jaydeep; Yadav, Ashish; Gangadharan, Roopesh; Prasad, Rambilas; Ulahannan, Shino; Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun

2015-01-01

Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

Energy Technology Data Exchange (ETDEWEB)

Joshi, Jaydeep, E-mail: Jaydeep.joshi@iter-india.org [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Yadav, Ashish; Gangadharan, Roopesh [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Prasad, Rambilas [Madan Mohan Malaviya University of Technology, Gorakhpur, Uttar Pradesh 273001 (India); Ulahannan, Shino [Airframe Aerodesigns Pvt. Ltd., HAL Airport Exit Road, Old Airport Road, Bengaluru 17 (India); Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India)

2015-10-15

Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

Data and Geocomputation: Time Critical Mission Support for the 2017 Hurricane Season

Science.gov (United States)

Bhaduri, B. L.; Tuttle, M.; Rose, A.; Sanyal, J.; Thakur, G.; White, D.; Yang, H. H.; Laverdiere, M.; Whitehead, M.; Taylor, H.; Jacob, M.

2017-12-01

A strong spatial data infrastructure and geospatial analysis capabilities are nucleus to the decision-making process during emergency preparedness, response, and recovery operations. For over a decade, the U.S. Department of Energy's Oak Ridge National Laboratory has been developing critical data and analytical capabilities that provide the Federal Emergency Management Agency (FEMA) and the rest of the federal response community assess and evaluate impacts of natural hazards on population and critical infrastructures including the status of the national electricity and oil and natural gas networks. These capabilities range from identifying structures or buildings from very high-resolution satellite imagery, utilizing machine learning and high-performance computing, to daily assessment of electricity restoration highlighting changes in nighttime lights for the impacted region based on the analysis of NOAA JPSS VIIRS Day/Night Band (DNB) imagery. This presentation will highlight our time critical mission support efforts for the 2017 hurricane season that witnessed unprecedented devastation from hurricanes Harvey, Irma, and Maria. ORNL provided 90m resolution LandScan USA population distribution data for identifying vulnerable population as well as structure (buildings) data extracted from 1m imagery for damage assessment. Spatially accurate data for solid waste facilities were developed and delivered to the response community. Human activity signatures were assessed from large scale collection of open source social media data around points of interests (POI) to ascertain level of destruction. The electricity transmission system was monitored in real time from data integration from hundreds of utilities and electricity outage information were provided back to the response community via standardized web-services.

THYC qualification on Vatican-1 low pressure tests

International Nuclear Information System (INIS)

Duval, C.; Guichard, J.

1991-06-01

PWR cores or fuel assemblies are components of a nuclear power plant involving single and two-phase flows in rod bundles. The knowledge of the detailed two-phase and three-dimensional flow patterns is necessary to evaluate the singularity (grids) and bypass effects on the Departure from Nucleate Boiling (DNB) in reactor cores during incidental transients. For that purpose, since 1989, the VATICAN experiment has been performed at EDF as a part of the qualification program of the three-dimensional computer code THYC, developed by EDF. The qualification strategy of the THYC software for PWR cores is the following: assuming the theoretical or experimental knowledge of regular and singular pressure drops and grid turbulence sources in single-phase, pressure drop multipliers and relative velocity in two-phase flow, the VATICAN experiment allows to evaluate the diffusion phenomena in two-phase flow. It provides thermalhydraulic measurements on a mock-up of a part of 900 MWe PWR fuel assembly in single and two-phase flows, with power and quality gradients. The first configuration of the mock-up, with simple spacer grids, is studied (VATICAN-1). The specific effects of mixing spacer grids will be compared to these data through a second configuration. The last void fraction measurements, using a γ-ray technique, performed on VATICAN-1 low pressure tests allowed to qualify a set of closure relations, particularly a model of little two-phase diffusion, adapted to two-phase flows at low pressure (5.0MPa). The qualification of subcooled boiling and diffusion models will continue on next VATICAN and other experimental campaigns [fr

Land, Cryosphere, and Nighttime Environmental Products from Suomi NPP VIIRS: Overview and Status

Science.gov (United States)

Roman, Miguel O.; Justice, Chris; Csiszar, Ivan

2014-01-01

The Visible Infrared Imaging Radiometer Suite (VIIRS) instrument was launched in October 2011 as part of the Suomi National Polar-orbiting Partnership (S-NPP: http://npp.gsfc.nasa.gov/). VIIRS was designed to improve upon the capabilities of the operational Advanced Very High Resolution Radiometer (AVHRR) and provide observation continuity with NASA's Earth Observing System's (EOS) Moderate Resolution Imaging Spectroradiometer (MODIS). Since the VIIRS first-light images were received in November 2011, NASA and NOAA funded scientists have been working to evaluate the instrument performance and derived products to meet the needs of the NOAA operational users and the NASA science community. NOAA's focus has been on refining a suite of operational products known as Environmental Data Records (EDRs), which were developed according to project specifications under the former National Polar-orbiting Environmental Satellite System (NPOESS). The NASA S-NPP Science Team has focused on evaluating the EDRs for science use, developing and testing additional products to meet science data needs and providing MODIS data product continuity. This paper will present to-date findings of the NASA Science Team's evaluation of the VIIRS Land and Cryosphere EDRs, specifically Surface Reflectance, Land Surface Temperature, Surface Albedo, Vegetation Indices, Surface Type, Active Fires, Snow Cover, Ice Surface Temperature, and Sea Ice Characterization (http://viirsland.gsfc.nasa.gov/index.html). The paper will also discuss new capabilities being developed at NASA's Land Product Evaluation and Test Element (http://landweb.nascom.nasa.gov/NPP_QA/); including downstream data and products derived from the VIIRS Day/Night Band (DNB).

Void fraction prediction of NUPEC PSBT tests by CATHARE code

International Nuclear Information System (INIS)

Del Nevo, A.; Michelotti, L.; Moretti, F.; Rozzia, D.; D'Auria, F.

2011-01-01

The current generation of thermal-hydraulic system codes benefits of about sixty years of experiments and forty years of development and are considered mature tools to provide best estimate description of phenomena and detailed reactor system representations. However, there are continuous needs for checking the code capabilities in representing nuclear system, for drawing attention to their weak points, for identifying models which need to be refined for best-estimate calculations. Prediction of void fraction and Departure from Nucleate Boiling (DNB) in system thermal-hydraulics is currently based on empirical approaches. The database carried out by Nuclear Power Engineering Corporation (NUPEC), Japan addresses these issues. It is suitable for supporting the development of new computational tools based on more mechanistic approaches (i.e. three-field codes, two-phase CFD, etc.) as well as for validating current generation of thermal-hydraulic system codes. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The paper reviews the activity carried out by CATHARE2 code on the basis of the subchannel (four test sections) and presents rod bundle (different axial power profile and test sections) experiments available in the database in steady state and transient conditions. The results demonstrate the accuracy of the code in predicting the void fraction in different thermal-hydraulic conditions. The tests are performed varying the pressure, coolant temperature, mass flow and power. Sensitivity analyses are carried out addressing nodalization effect and the influence of the initial and boundary conditions of the tests. (author)

Analysis of independent failure assumptions on postulated secondary high energy line ruptures

International Nuclear Information System (INIS)

Hollingsworth, S.D.

1977-01-01

Postulated ruptures of the main steam piping in pressurized water reactors result in large amounts of steam being removed from the secondary system. Since the energy removal rate could be many times that of nominal design power, there may be a rapid cooldown of the primary coolant system and a positive addition of reactivity to the reactor core. The Westinghouse protection system design concept incorporates features that trip the reactor, isolate the main steamlines and provide for automatic alternate shutdown capability in the form of boric acid solution injection into the primary coolant system. At the most limiting time in life (end of life) the reactivity calculated to be inserted by the cooldown is sufficient to overcome the shutdown margin predicted to be available from control rods with the most reactive rod in the fully withdrawn position. Because the boron injected into the core may be delayed due to system responses, there is potential that the reactor core could return critical and return to power. The extremely adverse radial power distributions caused by the fully withdrawn control rod causes localized high power densities that could lead to reduced heat transfer capability (DNB). Because of the large amount of stored energy in the reactor coolant system at full power, the cooldown and subsequent return to power is more severe when calculated from a shutdown, hot zero power condition. It is shown that the protection system design has large margins to protect against adverse core effects following a steamline rupture

FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results

Energy Technology Data Exchange (ETDEWEB)

Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.

1997-07-01

The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime.

Feasibility study on thermal-hydraulic design of reduced-moderation PWR-type core

International Nuclear Information System (INIS)

Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

2000-03-01

At JAERI, a conceptual study on reduced-moderation water reactor (RMWR) has been performed as one of the advanced reactor system which is designed so as to realize the conversion ratio more than unity. In this reactor concept, the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated. Therefore, an evaluation of the core thermal margin becomes very important in the design of the RMWR. In this study, we have performed a feasibility evaluation on thermal-hydraulic design of RM-PWR type core (core thermal output: 2900 MWt, Rod gaps: 1 mm). In RM-PWR core, seed and blanket regions are exist. In the blanket region, power density is lower than that of the seed region. Then, evaluation was performed under setting a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because it is possible that the coolant boils in the seed region. In the feasibility evaluations, subchannel code COBRA-IV-I was used in combination with KfK DNB (departure nucleate boiling) correlation. When coolant mass flow rate to the blanket fuel assembly is reduced by 40%, and that to the seed fuel assembly is increased, coolant boiling is not occurred in the assembly region calculation. Provided that the channel boxes to the blanket fuel assembly are set up and coolant mass flow rate to the blanket fuel assembly is reduced by 40%, it is confirmed by the whole core calculation that the boiling of the coolant is not occurred and the RM-PWR core is feasible. (author)

FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results

International Nuclear Information System (INIS)

Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.

1997-07-01

The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime

A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

Energy Technology Data Exchange (ETDEWEB)

Marguet, S.D. [Electricite de France (EDF), 92 - Clamart (France)

1997-12-31

Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF`s neutronic code COCCINELLE uses the Rowland`s formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission`s products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on `low` configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.) 7 refs.

A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

International Nuclear Information System (INIS)

Marguet, S.D.

1997-01-01

Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF's neutronic code COCCINELLE uses the Rowland's formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission's products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on 'low' configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.)

Upper atmospheric gravity wave details revealed in nightglow satellite imagery

Science.gov (United States)

Miller, Steven D.; Straka, William C.; Yue, Jia; Smith, Steven M.; Alexander, M. Joan; Hoffmann, Lars; Setvák, Martin; Partain, Philip T.

2015-01-01

Gravity waves (disturbances to the density structure of the atmosphere whose restoring forces are gravity and buoyancy) comprise the principal form of energy exchange between the lower and upper atmosphere. Wave breaking drives the mean upper atmospheric circulation, determining boundary conditions to stratospheric processes, which in turn influence tropospheric weather and climate patterns on various spatial and temporal scales. Despite their recognized importance, very little is known about upper-level gravity wave characteristics. The knowledge gap is mainly due to lack of global, high-resolution observations from currently available satellite observing systems. Consequently, representations of wave-related processes in global models are crude, highly parameterized, and poorly constrained, limiting the description of various processes influenced by them. Here we highlight, through a series of examples, the unanticipated ability of the Day/Night Band (DNB) on the NOAA/NASA Suomi National Polar-orbiting Partnership environmental satellite to resolve gravity structures near the mesopause via nightglow emissions at unprecedented subkilometric detail. On moonless nights, the Day/Night Band observations provide all-weather viewing of waves as they modulate the nightglow layer located near the mesopause (∼90 km above mean sea level). These waves are launched by a variety of physical mechanisms, ranging from orography to convection, intensifying fronts, and even seismic and volcanic events. Cross-referencing the Day/Night Band imagery with conventional thermal infrared imagery also available helps to discern nightglow structures and in some cases to attribute their sources. The capability stands to advance our basic understanding of a critical yet poorly constrained driver of the atmospheric circulation. PMID:26630004

Multi-scale analysis of nuclear reactor thermal-hydraulics-first applications using the NEPTUNE platform

International Nuclear Information System (INIS)

Guelfi, A.; Boucker, M.; Mimouni, S.; Bestion, D.; Boudier, P.

2005-01-01

The NEPTUNE project aims at building a new two-phase flow thermal-hydraulics platform for nuclear reactor simulation. EDF (Electricite de France) and CEA (Commissariat a l'Energie Atomique) with the co-sponsorship of IRSN (Institut de Radioprotection et Surete Nucleaire) and FRAMATOME-ANP, are jointly developing the NEPTUNE multi-scale platform that includes new physical models and numerical methods for each of the computing scales. One usually distinguishes three different scales for industrial simulations: the 'system' scale, the 'component' scale (subchannel analysis) and CFD (Computational Fluid Dynamics). In addition DNS (Direct Numerical Simulation) can provide information at a smaller scale that can be useful for the development of the averaged scales. The NEPTUNE project also includes work on software architecture and research on new numerical methods for coupling codes since both are required to improve industrial calculations. All these R and D challenges have been defined in order to meet industrial needs and the underlying stakes (mainly the competitiveness and the safety of Nuclear Power Plants). This paper focuses on three high priority needs: DNB (Departure from Nucleate Boiling) prediction, directly linked to fuel performance; PTS (Pressurized Thermal Shock), a key issue when studying the lifespan of critical components and LBLOCA (Large Break Loss of Coolant Accident), a reference accident for safety studies. For each of these industrial applications, we provide a review of the last developments within the NEPTUNE platform and we present the first results. A particular attention is also given to physical validation and the needs for further experimental data. (authors)

Polar2Grid 2.0: Reprojecting Satellite Data Made Easy

Science.gov (United States)

Hoese, D.; Strabala, K.

2015-12-01

Polar-orbiting multi-band meteorological sensors such as those on the Suomi National Polar-orbiting Partnership (SNPP) satellite pose substantial challenges for taking imagery the last mile to forecast offices, scientific analysis environments, and the general public. To do this quickly and easily, the Cooperative Institute for Meteorological Satellite Studies (CIMSS) at the University of Wisconsin has created an open-source, modular application system, Polar2Grid. This bundled solution automates tools for converting various satellite products like those from VIIRS and MODIS into a variety of output formats, including GeoTIFFs, AWIPS compatible NetCDF files, and NinJo forecasting workstation compatible TIFF images. In addition to traditional visible and infrared imagery, Polar2Grid includes three perceptual enhancements for the VIIRS Day-Night Band (DNB), as well as providing the capability to create sharpened true color, sharpened false color, and user-defined RGB images. Polar2Grid performs conversions and projections in seconds on large swaths of data. Polar2Grid is currently providing VIIRS imagery over the Continental United States, as well as Alaska and Hawaii, from various Direct-Broadcast antennas to operational forecasters at the NOAA National Weather Service (NWS) offices in their AWIPS terminals, within minutes of an overpass of the Suomi NPP satellite. Three years after Polar2Grid development started, the Polar2Grid team is now releasing version 2.0 of the software; supporting more sensors, generating more products, and providing all of its features in an easy to use command line interface.

Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

International Nuclear Information System (INIS)

Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

2008-01-01

The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

A feasibility study on the extended cycle from the point of view of Non-LOCA safety analysis

International Nuclear Information System (INIS)

Lee, Chul Sin; Kim, Hee Chul; Kim, Jeong Jin; Baek, Seung Su; Lee, Byung Il; Jeon, Tae Hyun; Lee, Jeong Chan

1996-06-01

Extended cycle operation has many advantages as compared with standard cycle operation (12 months) from the viewpoint of operators' exposure to radiation dose, man-power for maintenance and the production of nuclear waste. And it is more economic than the standard cycle operation. If the extended cycle operation is adopted in the CE type nuclear plant, the effective management and enhancement of operation capability could be expected. A feasibility study on the extended cycle operation should be performed. The purpose of this technical report is to perform safety analysis using the design data for Yonggwang Nuclear Plant (YGN) 3 and 4 extended cycle operation and to evaluate qualitatively and quantitatively the safety related design basis events to verify the satisfaction of the safety criteria. Boron dilution and steam line break accidents were found to be most influenced by the change of physics data due to the adaptation of the extended cycle operation. For boron dilution accident, source range monitor ratio of 3.07 for YGN 3 cycle 2 was decreased to 2.92. The 3D reactivity feedback effect due to the local heatup in the vicinity of stuck CEA was credited in the analysis of steam line break. No return-to-power occurred for the steam line break with offsite power available and return-to-power occurred for the steam line break with loss of offsite power. For the steam line break with loss of offsite power, the safety margin was preserved with respect to fuel performance (DNB and LHGR) despite the return-to-power. 6 tabs., 10 figs., 5 refs. (Author) .new

Innovative Control concepts for German pressurized water reactors

International Nuclear Information System (INIS)

Brzozowski, Raphael; Kuhn, Andreas

2010-01-01

Controlling reactor power without any manual support is becoming more and more important. The READIG project (READIG = Reactor Instrumentation and Digital Control) power control system installed in unit 2 of the Philippsburg nuclear power station (KKP 2) requires no manual intervention except for specific strategy criteria settings. It was even possible to eliminate the power distribution set points. With minor adaptations, this concept can be applied in other PWR plants as well. KKP 2 is a PWR plant with particularly sophisticated core charges; as a consequence, the I and C systems were adapted accordingly. The increase in integral reactor power and the low-leakage core charges are the main reasons for lower limiting margins, especially in peak limiting. The standard control concept was supplemented in such a way that a more precise fine control concept for power distribution in the full-load regime is achieved. The READIG project fully utilizes the possibilities offered by digital TXS Technology, which is why use is also made of physical parameterization. The new power distribution control concept has these advantages: - Operation at small peak-/DNB-reactor output limitation margins. - Stable control without manual intervention also in load cycles and in the frequency control mode. - Simplified operation due to omission of the power distribution set point. - Reduction to zero of the frequency of L-bank steps at constant power with superimposed frequency control mode. - Reduction to zero of the frequency of D-bank steps at constant power with superimposed frequency control mode. - Lower quantities of demineralized water to be fed at constant power with superimposed frequency control mode (±1%). (orig.)

Two-phase flow regimes and mechanisms of critical heat flux under subcooled flow boiling conditions

International Nuclear Information System (INIS)

Le Corre, Jean-Marie; Yao, Shi-Chune; Amon, Cristina H.

2010-01-01

A literature review of critical heat flux (CHF) experimental visualizations under subcooled flow boiling conditions was performed and systematically analyzed. Three major types of CHF flow regimes were identified (bubbly, vapor clot and slug flow regime) and a CHF flow regime map was developed, based on a dimensional analysis of the phenomena and available experimental information. It was found that for similar geometric characteristics and pressure, a Weber number (We)/thermodynamic quality (x) map can be used to predict the CHF flow regime. Based on the experimental observations and the review of the available CHF mechanistic models under subcooled flow boiling conditions, hypothetical CHF mechanisms were selected for each CHF flow regime, all based on a concept of wall dry spot overheating, rewetting prevention and subsequent dry spot spreading. Even though the selected concept has not received much attention (in term or theoretical developments and applications) as compared to other more popular DNB models, its basis have often been cited by experimental investigators and is considered by the authors as the 'most-likely' mechanism based on the literature review and analysis performed in this work. The selected modeling concept has the potential to span the CHF conditions from highly subcooled bubbly flow to early stage of annular flow and has been numerically implemented and validated in bubbly flow and coupled with one- and three-dimensional (CFD) two-phase flow codes, in a companion paper. [Le Corre, J.M., Yao, S.C., Amon, C.H., in this issue. A mechanistic model of critical heat flux under subcooled flow boiling conditions for application to one and three-dimensional computer codes. Nucl. Eng. Des.].

Fuel design and operational experience in Loviisa NPP, future trends in fuel issues

International Nuclear Information System (INIS)

Terasvirta, R.

2001-01-01

This paper summarizes the past operational experience of nuclear fuel with reference to most significant design changes during the years. In general, the fuel behaviour in Loviisa NPP in terms of leaking fuel assemblies has been good. The major improvements by fuel design changes in Lovissa NPP, including rod elongation margin, change in the pellet design and manufacturing process, upper grid modifications, change of material in the spacer grids and reduction of the shroud tube thickness are discussed and related to the number of failed fuel assemblies. The detailed investigation of fuel failure rates as function of different fuel and operation characteristics allows to classify the leaking fuel assemblies according to the cause of failure. In a brief discussion concerning new changes in the safety guide for nuclear design limits, re-issued by the Finnish Safety Authority (STUK), the frequencies for class 1 and class 2 accidents are determined. Another change in this guide is the introduction of design limits for the number of fuel rods experiencing DNB in class 1 accidents and number of failed rods in class 2 accidents. It is concluded that as far as normal operation is concerned, there seems to be sufficiently large margin between present operational limits in Loviisa and the design limits. The real limits do not come from fuel behaviour in the normal operation or operational occurrences but from the accident behaviour. At the moment, fuel assembly burnup extension beyond 45 MWd/kgU is clearly out of the question before further information and positive results are obtained on high burnup fuel behaviour in accident conditions

Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

Energy Technology Data Exchange (ETDEWEB)

Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

2015-08-15

Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

Development of a coupling scheme between MCNP5 and subchanflow for the PIN- and fuel Assembly-Wise simulation of LWR and innovative reactors

International Nuclear Information System (INIS)

Ivanov, A.; Sanchez, V.; Imke, U.

2011-01-01

In order to increase the accuracy and the degree of spatial resolution of core design studies, coupled 3D neutronic (deterministic and Monte Carlo) and 3D thermal hydraulics (CFD and subchannel) codes are being developed worldwide. At KIT both deterministic and Monte Carlo codes were coupled with subchannel codes and applied to predict the safety-related design parameters such as pin power, maximal cladding and fuel temperature, DNB. These coupling approaches were revised and improved based on the experience gained. One particular example is replacing COBRA-TF with SUBCHANFLOW, in-house development subchannel code, in the COBRA-TF/MCNP coupling, accompanied with new way of radial mapping between the neutronic and thermal hydraulic domains. The new coupled system MCNP5/SUBCHANFLOW makes it possible to investigate variety of fuel assembly types (BWR, PWR or SCFR). Key issues in such a coupled system are the way in which thermal-hydraulic/neutronic feedbacks, accuracy of the Monte Carlo solutions and observation of convergence during the iterative solution are handled. Another key issue that might be considered is the optimal application of parallel computing. Using multi-processor computer architectures, it is possible to reduce the Monte- Carlo running time and obtain converged results within reasonable time limit. In particular it is shown that by exploiting the capabilities of multi-processor calculation, it is possible to investigate large fuel assemblies in a pin-by-pin manner with a resolution at pin and subchannel level. One of the most important issues addressed in the current work is the temperature effects on nuclear data. For the particular studies pseudo material approach was used, which produces interpolated results for Doppler broadened cross sections from NJOY pre-generated nuclear data. (author)

Computational model for turbulent flow around a grid spacer with mixing vane

International Nuclear Information System (INIS)

Tsutomu Ikeno; Takeo Kajishima

2005-01-01

Turbulent mixing coefficient and pressure drop are important factors in subchannel analysis to predict onset of DNB. However, universal correlations are difficult since these factors are significantly affected by the geometry of subchannel and a grid spacer with mixing vane. Therefore, we propose a computational model to estimate these factors. Computational model: To represent the effect of geometry of grid spacer in computational model, we applied a large eddy simulation (LES) technique in couple with an improved immersed-boundary method. In our previous work (Ikeno, et al., NURETH-10), detailed properties of turbulence in subchannel were successfully investigated by developing the immersed boundary method in LES. In this study, additional improvements are given: new one-equation dynamic sub-grid scale (SGS) model is introduced to account for the complex geometry without any artificial modification; the higher order accuracy is maintained by consistent treatment for boundary conditions for velocity and pressure. NUMERICAL TEST AND DISCUSSION: Turbulent mixing coefficient and pressure drop are affected strongly by the arrangement and inclination of mixing vane. Therefore, computations are carried out for each of convolute and periodic arrangements, and for each of 30 degree and 20 degree inclinations. The difference in turbulent mixing coefficient due to these factors is reasonably predicted by our method. (An example of this numerical test is shown in Fig. 1.) Turbulent flow of the problem includes unsteady separation behind the mixing vane and vortex shedding in downstream. Anisotropic distribution of turbulent stress is also appeared in rod gap. Therefore, our computational model has advantage for assessing the influence of arrangement and inclination of mixing vane. By coarser computational mesh, one can screen several candidates for spacer design. Then, by finer mesh, more quantitative analysis is possible. By such a scheme, we believe this method is useful

Experimental study of CHF enhancement using Fe{sub 3}O{sub 4} nanofluids in the subcooled boiling region

Energy Technology Data Exchange (ETDEWEB)

Choi, Young Jae; Kam, Dong Hoon; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2015-10-15

This study may give overall trends of CHF enhancement in the subcooled boiling region. In our experiment, subcooled flow boiling CHF enhancement phenomena in water and nano-coated surface was investigated in mass flux from 1000 to 5000 kg/m{sup 2}s. CHF enhancement of nanoparticles coated tube in DI water increased as exit quality get bigger at same massflux. Various methods to improve CHF characteristics are introduced, especially nanofluids are used for enhancing the CHF. Nanofluids is a colloidal suspension that nanoparticles are mixed with basic fluid. Normally the use of nanofluids as working fluid improves the flow boiling CHF characteristics. Lee et al. already researched the CHF characteristics using nanofluids. As exit quality increased from 0.07 to 0.74, CHF enhancement gradually decreased and approached zero. CHF enhancement was observed when exit quality was low and a DNB-like thermal crisis occurred. But CHF enhancement didn't occur for high exit quality, but LFD-type thermal crisis occurred. Because LFD phenomena are nearly unaffected by the surface conditions, CHF enhancement is not expected for annular flow with high exit quality. Kim et al. performed flow boiling CHF enhancement at subcooled region using alumina-water, zinc-oxide-water and diamond-water nanofluids. The CHF was enhanced by increasing wettability from nanoparticle deposition. CHF enhancement occurred in high mass flux (2000-2500 kg/m{sup 2}s), but CHF enhancement didn't occur in low mass flux (1500 kg/m{sup 2}s). The amount of nanoparticle deposition on each tube can be different during experiments by the several conditions such as deposition time, mass flux and heat flux. So, before the nanofluid experiment conducted, all tube are deposited in same condition of heat flux, concentration and time.

Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

International Nuclear Information System (INIS)

Tyagi, Himanshu; Soni, Jignesh; Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli; Gahlaut, Agrajit; Joshi, Jaydeep; Parmar, Deepak; Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun

2016-01-01

Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as

A user's guide to the PLTEMP/ANL code.

Energy Technology Data Exchange (ETDEWEB)

Kalimullah, M. (Nuclear Engineering Division)

2011-07-05

PLTEMP/ANL V4.1 is a FORTRAN program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of ''PLTEMP'' codes in use at ANL for the past 20 years. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates/tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as Onset-of-Nucleate boiling (ONB), departure from nucleate boiling (DNB), and onset of flow instability (FI). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst's time.

Prediction of Critical Heat Flux under Rolling Motion

Energy Technology Data Exchange (ETDEWEB)

Hwang, Jinseok; Lee, Yeongun; Park, Gooncherl [Seoul National Univ., Seoul (Korea, Republic of)

2013-05-15

The aim to this paper may be summarized as follows: identify the flow regime compare with existing void-quality relationship and void fraction at OAF derived from the vapor superficial velocity obtained by the churn-to annular flow criterion, develop and evaluate the correlation for accurate prediction of CHF ratio under rolling motion. Experimentally measured CHF results from the previous study were not well-predicted by existing CHF correlations developed for wide range of pressure under rolling motion in vertical tube. Specifically, existing correlations do not account for the dynamic motion parameter, such as tangential and centrifugal force. This study reviewed some existing correlation and experimental studies related to reduction and enhancement of CHF and heat transfer and flow behavior under heaving and rolling motion, and developed a CHF ratio correlation for upward flow vertical tube under rolling motion. Based upon dimensionless groups, equations and interpolation factor, an empirical CHF correlation has been developed which is consistent with experimental data for uniformly heated tubes internally cooled by R-134 under rolling motion. Flow regime was determined through the prediction method for annular flow. Non-dimensional number and function were decided by CHF mechanism of each region. Interaction of LFD and DNB regions is taken into account by means of power interpolation which is reflected void fraction at OAF. The suggested correlation predicted the CHF Ratio with reasonable accuracy, showing an average error of -0.59 and 2.51% for RMS. Rolling motion can affect bubble motion and liquid film behavior complexly by combination of tangential and centrifugal forces and mass flow than heaving motion. Through a search of literature and a comparison of previous CHF ratio results, this work can contribute to the study of boiling heat transfer and CHF for the purpose of enhancement or reduction the CHF of dynamic motion system, such as marine reactor.

Indigenous Manufacturing realization of TWIN Source

Science.gov (United States)

Pandey, R.; Bandyopadhyay, M.; Parmar, D.; Yadav, R.; Tyagi, H.; Soni, J.; Shishangiya, H.; Sudhir Kumar, D.; Shah, S.; Bansal, G.; Pandya, K.; Parmar, K.; Vuppugalla, M.; Gahlaut, A.; Chakraborty, A.

2017-04-01

TWIN source is two RF driver based negative ion source that has been planned to bridge the gap between single driver based ROBIN source (currently operational) and eight river based DNB source (to be operated under IN-TF test facility). TWIN source experiments have been planned at IPR keeping the objective of long term domestic fusion programme to gain operational experiences on vacuum immersed multi driver RF based negative ion source. High vacuum compatible components of twin source are designed at IPR keeping an aim on indigenous built in attempt. These components of TWIN source are mainly stainless steel and OFC-Cu. Being high heat flux receiving components, one of the major functional requirements is continuous heat removal via water as cooling medium. Hence for the purpose stainless steel parts are provided with externally milled cooling lines and that shall be covered with a layer of OFC-cu which would be on the receiving side of high heat flux. Manufacturability of twin source components requires joining of these dissimilar materials via process like electrode position, electron beam welding and vacuum brazing. Any of these manufacturing processes shall give a vacuum tight joint having proper joint strength at operating temperature and pressure. Taking the indigenous development effort vacuum brazing (in non-nuclear environment) has been opted for joining of dissimilar materials of twin source being one of the most reliable joining techniques and commercially feasible across the suppliers of country. Manufacturing design improvisation for the components has been done to suit the vacuum brazing process requirement and to ease some of the machining without comprising over the functional and operational requirements. This paper illustrates the details on the indigenous development effort, design improvisation to suits manufacturability, vacuum brazing basics and its procedures for twin source components.

Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

Energy Technology Data Exchange (ETDEWEB)

Tyagi, Himanshu, E-mail: htyagi@iter-india.org [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Soni, Jignesh [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Gahlaut, Agrajit [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Joshi, Jaydeep; Parmar, Deepak [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India)

2016-11-15

Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as

JPSS-1 VIIRS Pre-Launch Response Versus Scan Angle Testing and Performance

Science.gov (United States)

Moyer, David; McIntire, Jeff; Oudrari, Hassan; McCarthy, James; Xiong, Xiaoxiong; De Luccia, Frank

2016-01-01

The Visible Infrared Imaging Radiometer Suite (VIIRS) instruments on-board both the Suomi National Polar-orbiting Partnership (S-NPP) and the first Joint Polar Satellite System (JPSS-1) spacecraft, with launch dates of October 2011 and December 2016 respectively, are cross-track scanners with an angular swath of +/-56.06 deg. A four-mirror Rotating Telescope Assembly (RTA) is used for scanning combined with a Half Angle Mirror (HAM) that directs light exiting from the RTA into the aft-optics. It has 14 Reflective Solar Bands (RSBs), seven Thermal Emissive Bands (TEBs) and a panchromatic Day Night Band (DNB). There are three internal calibration targets, the Solar Diffuser, the BlackBody and the Space View, that have fixed scan angles within the internal cavity of VIIRS. VIIRS has calibration requirements of 2% on RSB reflectance and as tight as 0.4% on TEB radiance that requires the sensor's gain change across the scan or Response Versus Scan angle (RVS) to be well quantified. A flow down of the top level calibration requirements put constraints on the characterization of the RVS to 0.2%-0.3% but there are no specified limitations on the magnitude of response change across scan. The RVS change across scan angle can vary significantly between bands with the RSBs having smaller changes of approximately 2% and some TEBs having approximately 10% variation. Within aband, the RVS has both detector and HAM side dependencies that vary across scan. Errors in the RVS characterization will contribute to image banding and striping artifacts if their magnitudes are above the noise level of the detectors. The RVS was characterized pre-launch for both S-NPP and JPSS-1 VIIRS and a comparison of the RVS curves between these two sensors will be discussed.

A User's Guide to the PLTEMP/ANL Code

Energy Technology Data Exchange (ETDEWEB)

Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-07-07

PLTEMP/ANL V4.2 is a FORTRAN program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of ''PLTEMP'' codes in use at ANL for the past 20 years. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates/tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as Onset-of- Nucleate boiling (ONB), departure from nucleate boiling (DNB), and onset of flow instability (FI). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.

Características dendrométricas de um povoamento de nim indiano (Azadirachta indica A. Juss no semiárido paraibano

Directory of Open Access Journals (Sweden)

Francisco Tibério de Alencar Moreira

2012-11-01

Full Text Available O Nim (Azadirachia indica A. Juss é uma espécie de múltiplo uso que pertence a família Meliaceae. Por possuir múltiplos usos, o nim tem despertado a atenção e seus produtos têm sido cada vez mais utilizados na área de agricultura. No entanto, aspectos dendrométricos relacionados ao crescimento da espécie ainda são escassos. O presente trabalho teve como objetivo avaliar o crescimento da espécie nim plantada em região semiárida do Nordeste brasileiro em dois diferentes espaçamentos. A coleta de dados foi realizada em duas áreas localizadas na Fazenda Laranjeiras, localizada no município de São José de Espinharas, Estado da Paraíba. No plantio-1 o espaçamento usado foi de 5 m x 5 m semeado no ano de 1996, em uma área de 2100 m2 e no plantio-2 o espaçamento usado foi de 4 m x 4 m em 1998, em uma área de 1600 m2. As variáveis medidas foram: DNB-Diâmetro a 0,30 cm do nível do solo (cm, HF-Altura do fuste comercial (m, HT-Altura total (m e DC-Diâmetro de copa (m. A partir destes dados foram calculados o volume cilíndrico (m3/ha, volume real (m3/ha, volume empilhado (st/ha e a área basal (m2/ha. O crescimento do nim indiano na região semiárida paraibana apresentou comportamento semelhante ao de outras áreas experimentais estudadas. Quando se pretende produzir madeira com finalidade energética o espaçamento mais indicado, entre os dois avaliados, é o de menor espaçamento.

Velvet bean severe mosaic virus: a distinct begomovirus species causing severe mosaic in Mucuna pruriens (L.) DC.

Science.gov (United States)

Zaim, Mohammad; Kumar, Yogesh; Hallan, Vipin; Zaidi, A A

2011-08-01

Velvet bean [Mucuna pruriens (L.) DC] is one of the most important medicinal plants. It is used to treat many ailments, but is widely used for the treatment especially for Parkinson's disease because of the presence of 3,4-dihydroxyphenylalanine (L-dopa) in it. It was noticed in last 5 years that the plants in the field showed severe mosaic, downward curling of the leaves, stunting, etc. This is consistently observed over the years in India. The disease was transmitted by whiteflies and by grafting and the causal agent was found to be a bipartite begomovirus. The whole genome was amplified by rolling circle amplification (RCA) using ϕ-29 DNA polymerase and characterized. DNA-A and DNA-B shared a 124-nucleotide (nt) long highly conserved (98%) common region (CR). Comparisons with other begomovirus showed that DNA-A sequence has highest identity (76%) with an isolate of Mungbean yellow mosaic India virus (MYMIV; AY937195) reported from India. This data suggested that the present isolate is a new species of genus Begomovirus for which the name "Velvet bean severe mosaic virus" (VbSMV) is proposed. DNA-B has a maximum sequence identity of 49% with an isolate of Horsegram yellow mosaic virus (HgYMV; AM932426) reported from India. Infectious clones consisting of a 1.7 mer partial tandem repeat of DNA-A and a dimer of DNB-B were constructed and agro-inoculated to Macuna pruriens (L.) DC plants, which showed field observed symptoms 24 days post-infiltration (dpi). In phylogenetic analysis, DNA-A and DNA-B of the present isolate grouped with DNA-A of different begomoviruses reported from fabaceous crops. The study presents first ever molecular evidence of any disease in velvet bean and whole genome analysis of the causative virus which is a distinct bipartite species of Begomovirus.

Satellite-based Tropical Cyclone Monitoring Capabilities

Science.gov (United States)

Hawkins, J.; Richardson, K.; Surratt, M.; Yang, S.; Lee, T. F.; Sampson, C. R.; Solbrig, J.; Kuciauskas, A. P.; Miller, S. D.; Kent, J.

2012-12-01

Satellite remote sensing capabilities to monitor tropical cyclone (TC) location, structure, and intensity have evolved by utilizing a combination of operational and research and development (R&D) sensors. The microwave imagers from the operational Defense Meteorological Satellite Program [Special Sensor Microwave/Imager (SSM/I) and the Special Sensor Microwave Imager Sounder (SSMIS)] form the "base" for structure observations due to their ability to view through upper-level clouds, modest size swaths and ability to capture most storm structure features. The NASA TRMM microwave imager and precipitation radar continue their 15+ yearlong missions in serving the TC warning and research communities. The cessation of NASA's QuikSCAT satellite after more than a decade of service is sorely missed, but India's OceanSat-2 scatterometer is now providing crucial ocean surface wind vectors in addition to the Navy's WindSat ocean surface wind vector retrievals. Another Advanced Scatterometer (ASCAT) onboard EUMETSAT's MetOp-2 satellite is slated for launch soon. Passive microwave imagery has received a much needed boost with the launch of the French/Indian Megha Tropiques imager in September 2011, basically greatly supplementing the very successful NASA TRMM pathfinder with a larger swath and more frequent temporal sampling. While initial data issues have delayed data utilization, current news indicates this data will be available in 2013. Future NASA Global Precipitation Mission (GPM) sensors starting in 2014 will provide enhanced capabilities. Also, the inclusion of the new microwave sounder data from the NPP ATMS (Oct 2011) will assist in mapping TC convective structures. The National Polar orbiting Partnership (NPP) program's VIIRS sensor includes a day night band (DNB) with the capability to view TC cloud structure at night when sufficient lunar illumination exits. Examples highlighting this new capability will be discussed in concert with additional data fusion efforts.

A User's Guide to the PLTEMP/ANL Code

International Nuclear Information System (INIS)

Olson, Arne P.; Kalimullah, M.

2015-01-01

PLTEMP/ANL V4.2 is a FORTRAN program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of ''PLTEMP'' codes in use at ANL for the past 20 years. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates/tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as Onset-of- Nucleate boiling (ONB), departure from nucleate boiling (DNB), and onset of flow instability (FI). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst's time.

A user's guide to the PLTEMP/ANL code

International Nuclear Information System (INIS)

Kalimullah, M.

2011-01-01

PLTEMP/ANL V4.1 is a FORTRAN program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of ''PLTEMP'' codes in use at ANL for the past 20 years. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates/tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as Onset-of-Nucleate boiling (ONB), departure from nucleate boiling (DNB), and onset of flow instability (FI). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst's time.

Pressure loss coefficient and flow rate of side hole in a lower end plug for dual-cooled annular nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Shin, Chang-Hwan, E-mail: shinch@kaeri.re.kr; Park, Ju-Yong, E-mail: juyong@kaeri.re.kr; In, Wang-Kee, E-mail: wkin@kaeri.re.kr

2013-12-15

Highlights: • A lower end plug with side flow holes is suggested to provide alternative flow paths of the inner channel. • The inlet loss coefficient of the lower end plug is estimated from the experiment. • The flow rate through the side holes is estimated in a complete entrance blockage of inner channel. • The consequence in the reactor core condition is evaluated with a subchannel analysis code. - Abstract: Dual-cooled annular nuclear fuel for a pressurized water reactor (PWR) has been introduced for a significant increase in reactor power. KAERI has been developing a dual-cooled annular fuel for a power uprate of 20% in an optimized PWR in Korea, the OPR1000. This annular fuel can help decrease the fuel temperature substantially relative to conventional cylindrical fuel at a power uprate. Annular fuel has dual flow channels around itself; however, the inner flow channel has a weakness in that it is isolated unlike the outer flow channel, which is open to other neighbouring outer channels for a coolant exchange in the reactor core. If the entrance of the inner channel is, as a hypothetical event, completely blocked by debris, the inner channel will then experience a rapid increase in coolant temperature such that a departure from nucleate boiling (DNB) may occur. Therefore, a remedy to avoid such a postulated accident is indispensable for the safety of annular fuel. A lower end plug with side flow holes was suggested to provide alternative flow paths in addition to the central entrance of the inner channel. In this paper, the inlet loss coefficient of the lower end plug and the flow rate through the side holes were estimated from the experimental results even in a complete entrance blockage of the inner channel. An optimization for the side hole was also performed, and the results are applied to a subchannel analysis to evaluate the consequence in the reactor core condition.

Evaluation on the heat removal capacity of the first wall for water cooled breeder blanket of CFETR

Energy Technology Data Exchange (ETDEWEB)

Jiang, Kecheng, E-mail: jiangkecheng@ipp.ac.cn; Cheng, Xiaoman; Chen, Lei; Huang, Kai; Ma, Xuebin; Liu, Songlin

2016-02-15

Highlights: • Heat removal capacity of the FW is evaluated under BWR, PWR and He coolant inlet conditions. • Heat transfer property of the gas–liquid two phase and the two boiling crises are analyzed. • Heat removal capacity of water is larger than helium coolant. - Abstract: The water cooled ceramic breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). As an important component of the blanket, the FW should satisfy with the thermal requirements in any case. In this paper, three parameters including the heat removal capacity, coolant pressure drop as well as the temperature rise of the FW were investigated under different coolant velocity and heat flux from the plasma. Using the same first wall structure, two main water cooled schemes including Boiling Water Reactor (BWR, 7 MPa pressure and 265 °C temperature inlet) and Pressurized Water Reactor (PWR, 15 MPa pressure and 285 °C temperature inlet) conditions are discussed in the thermal hydraulic calculation. For further research, the thermal hydraulic characteristics of using helium as coolant (8 MPa pressure, 300 °C temperature inlet) are also explored to provide CFETR blanket design with more useful data supports. Without regard to the outlet coolant condition requirements of the blanket, the results indicate that the ultimate heat flux that the FW can resist is 2.2 MW/m{sup 2} at velocity of 5 m/s for BWR, 2.0 MW/m{sup 2} at velocity of 5 m/s for PWR and 0.87 MW/m{sup 2} for helium at velocity 100 m/s under the chosen operation condition. The detrimental departure from nucleate boiling (DNB) crisis would occur at the velocity of 1 m/s under the heat flux of 3 MW/m{sup 2} and dry out crisis appears at the velocity of less than 0.2 m/s with the heat flux of more than 1 MW/m{sup 2} for BWR. The further blanket/FW optimization design is provided with more useful data references according to the abundant calculation results.

JPSS Products, Applications and Training

Science.gov (United States)

Torres, J. R.; Connell, B. H.; Miller, S. D.

2017-12-01

The Joint Polar Satellite System (JPSS) is a new generation polar-orbiting operational environmental satellite system that will monitor the weather and environment around the globe. JPSS will provide technological and scientific improvements in environmental monitoring via high resolution satellite imagery and derived products that stand to improve weather forecasting capabilities for National Weather Service (NWS) forecasters and complement operational Geostationary satellites. JPSS will consist of four satellites, JPSS-1 through JPSS-4, where JPSS-1 is due to launch in Fall 2017. A predecessor, prototype and operational risk-reduction for JPSS is the Suomi-National Polar-orbiting Partnership (S-NPP) satellite, launched on 28 October 2011. The following instruments on-board S-NPP will also be hosted on JPSS-1: Visible Infrared Imaging Radiometer Suite (VIIRS), Cross-track Infrared Sounder (CrIS), Advanced Technology Microwave Sounder (ATMS), Ozone Mapping and Profiler Suite (OMPS) and the Clouds and Earth's Radiant Energy System (CERES). JPSS-1 instruments will provide satellite imagery, products and applications to users. The applications include detecting water and ice clouds, snow, sea surface temperatures, fog, fire, severe weather, vegetation health, aerosols, and sensing reflected lunar and emitted visible-wavelength light during the nighttime via the Day/Night Band (DNB) sensor included on VIIRS. Currently, there are only a few polar products that are operational for forecasters, however, more products will become available in the near future via Advanced Weather Interactive Processing System-II (AWIPS-II)-a forecasting analysis software package that forecasters can use to analyze meteorological data. To complement the polar products an wealth of training materials are currently in development. Denoted as the Satellite Foundational Course for JPSS (SatFC-J), this training will benefit NWS forecasters to utilize satellite data in their forecasts and daily

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

Energy Technology Data Exchange (ETDEWEB)

Brown, C.S., E-mail: csbrown3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Raleigh, NC 27695-7909 (United States); Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3870 (United States); Kucukboyaci, V., E-mail: kucukbvn@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Sung, Y., E-mail: sungy@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2016-12-01

Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

Investigation of forced convection heat transfer of supercritical pressure water in a vertically upward internally ribbed tube

International Nuclear Information System (INIS)

Wang Jianguo; Li Huixiong; Guo Bin; Yu Shuiqing; Zhang Yuqian; Chen Tingkuan

2009-01-01

similar to the DNB at subcritical pressures.

Extreme trace elements fractionation in Cenozoic nephelinites and phonolites from the Moroccan Anti-Atlas (Eastern Saghro)

Science.gov (United States)

Berger, Julien; Ennih, Nasser; Liégeois, Jean-Paul

2014-12-01

Nephelinites and phonolites from the Moroccan Anti-Atlas form a cogenetic series of volcanic rocks linked by a fractional crystallization process and showing continuous evolutionary trends for trace-elements. According to partial melting calculations, minor element data in olivine and review of published experimental studies, the most primitive nephelinites are low degree (~ 2%) partial melts from a carbonated LREE-rich spinel lherzolite. Sr-Nd-Pb isotopic compositions indicate the participation of both DM and HIMU end-members in the mantle source of nephelinites; the HIMU component is here interpreted as a relic of the shallow metasomatized Pan-African mantle. The phonolites show similar isotopic composition except for slightly more radiogenic Sr isotopic values. Fractional crystallization calculations were performed using trace-element mineral/bulk rock coefficients determined with new LA-ICP-MS data on minerals together with published equilibrium partition coefficients. The decrease of LREE, Sr and Ba with increasing differentiation is explained by fractionation of large amounts of apatite. Th, Nb and Zr display a behavior of very incompatible elements, reaching extreme concentration in most differentiated phonolites. Ta, Hf and MREE by contrast are characterized by a moderately incompatible to compatible behavior during differentiation. Fractionation of small amount of titanite, in which Ta, Hf and MREE are highly compatible compared to Nb, Zr and LREE (DNb/DTa: 2, DZr/DHf: 1.5 for titanite/phonolite ratios), explains the observed increase in Nb/Ta and Zr/Hf ratios with increasing silica content, from 18 and 40 in nephelinites to 70 and 80 in phonolites, respectively. Clinopyroxene also contributed to the fractionation of Hf from Zr in the very first steps of crystallization. The low values of Nb/Ta and Zr/Hf ratios observed in the two most differentiated Si-rich phonolites are probably a consequence of late stage segregation of volatile-rich agpaitic

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

International Nuclear Information System (INIS)

Brown, C.S.; Zhang, H.; Kucukboyaci, V.; Sung, Y.

2016-01-01

Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

Association between nighttime artificial light pollution and sea turtle nest density along Florida coast: A geospatial study using VIIRS remote sensing data.

Science.gov (United States)

Hu, Zhiyong; Hu, Hongda; Huang, Yuxia

2018-08-01

Artificial lighting at night has becoming a new type of pollution posing an important anthropogenic environmental pressure on organisms. The objective of this research was to examine the potential association between nighttime artificial light pollution and nest densities of the three main sea turtle species along Florida beaches, including green turtles, loggerheads, and leatherbacks. Sea turtle survey data was obtained from the "Florida Statewide Nesting Beach Survey program". We used the new generation of satellite sensor "Visible Infrared Imaging Radiometer Suite (VIIRS)" (version 1 D/N Band) nighttime annual average radiance composite image data. We defined light pollution as artificial light brightness greater than 10% of the natural sky brightness above 45° of elevation (>1.14 × 10 -11 Wm -2 sr -1 ). We fitted a generalized linear model (GLM), a GLM with eigenvectors spatial filtering (GLM-ESF), and a generalized estimating equations (GEE) approach for each species to examine the potential correlation of nest density with light pollution. Our models are robust and reliable in terms of the ability to deal with data distribution and spatial autocorrelation (SA) issues violating model assumptions. All three models found that nest density is significantly negatively correlated with light pollution for each sea turtle species: the higher light pollution, the lower nest density. The two spatially extended models (GLM-ESF and GEE) show that light pollution influences nest density in a descending order from green turtles, to loggerheads, and then to leatherbacks. The research findings have an implication for sea turtle conservation policy and ordinance making. Near-coastal lights-out ordinances and other approaches to shield lights can protect sea turtles and their nests. The VIIRS DNB light data, having significant improvements over comparable data by its predecessor, the DMSP-OLS, shows promise for continued and improved research about ecological effects of

Best estimate probabilistic safety assessment results for the Westinghouse Advanced Loop Tester (WALT)

International Nuclear Information System (INIS)

Wang, Guoqiang; Xu, Yiban; Oelrich, Robert L. Jr.; Byers, William A.; Young, Michael Y.; Karoutas, Zeses E.

2011-01-01

utilizing the correct conditions from the rectifier is critical for benchmarking the Dittus-Boelter forced convection as well as potential departure from nucleate boiling (DNB) margin gains. Summary and conclusions of this paper is important and useful for safe operation and/or more accurate data reduction of the WALT loop and other similar or relevant test facilities. (author)

Light pollution: measuring and modelling skyglow. An application in two Portuguese reserves

Science.gov (United States)

Lima, Raul Cerveira Pinto Sousa

Outdoors human-made lighting at night causes sky glow, one of the effects of light pollution. Sky glow is rising with the growth of world population. Urban inhabitants are increasingly deprived from a starry sky. However, since light propagates to regions far from where it is produced, light pollution spreads to places where few or none artificial light at night existed, disturbing the quality of the night sky. In this work we assess for the first time the sky brightness of two regions in Portugal, the Peneda-Geres National Park, and the recently created Starlight Reserve Dark Sky® Alqueva. We used a portable unit, a Unihedron Sky Quality Meter-L (SQM-L), to measure the luminance of the night sky. We also tested the SQM-L in a laboratory to a more thorough analysis of the device, and to check the effect of polarization on the unit, suggested by our observations and other users. Our results suggest that the SQM-L is not affected by any measurable effect of polarization, but some guidelines to use the SQM-L in the field are provided based on our work. The data from the field measurement was used to compare to one light pollution propagation model (Kocifaj, 2007), using VIIRS DNB satellite upwards radiance as input to the model. The results obtained from the model are favourably compared to the field measurements. We proceeded to a set of tests with the model to find the best fit. Our best results were achieved by analysing the data by night rather than the global set of data. Our first results were used to apply to the classification of the region of Alqueva to a Starlight Tourism Destination. That classification was attained during the course of this work (December 2011). A guideline on the Peneda-Geres National Park was also implemented after our first results were provided. We believe we have achieved a set of results in a set of parallel issues all related to light pollution that we hope may contribute to the current knowledge on this area of research.

Development of an artificial neural network model for on-line thermal margin estimation of a nuclear reactor core

International Nuclear Information System (INIS)

Kim, Hyun Koon

1992-02-01

One of the key safety parameters related to thermal margin in a Pressurized Water Reactor (PWR) core, is Departure from Nucleate Boiling Ratio (DNBR), which is to be assessed and continuously monitored during operation via either an analog or a digital monitoring system. The digital monitoring system, in general, allows more thermal margin than the analog system through the on-line computation of DNBR using the measured parameters as inputs to a simplified, fast running computer code. The purpose of this thesis is to develop an advanced method for on-line DNBR estimation by introducing an artifactual neural network model for best-estimation of DNBR at the given reactor operating conditions. the neural network model, consisting of three layers with five operating parameters in the input layer, provides real-time prediction accuracy of DNBR by training the network against the detailed simulation results for various operating conditions. The overall training procedure is developed to learn the characteristics of DNBR behaviour in the reactor core. First, a set of random combination of input variables is generated by Latin Hypercube Sampling technique performed on a wide range of input parameters. Second, the target values of DNBR to be referenced for training are calculated using a detailed simulation code, COBRA-IV. Third, the optimized training input data are selected. Then, training is performed using an Error Back Propagation algorithm. After completion of training, the network is tested on the examining data set in order to investigate the generalization capability of the network responses for the steady state operating condition as well as for the transient situations where DNB is of a primary concern. The test results show that the values of DNBR predicted by the neural network are maintained at a high level of accuracy for the steady state condition, and are in good agreements with the transient situation, although slightly conservative as compared to those

Enhancement of nuclear heat transfer in a typical pressurized water reactor by new spacer grids

International Nuclear Information System (INIS)

Nazifi, M.; Nematollahi, M.

2007-01-01

enhancement in nuclear heat transfer among the other mixing devices. Also the results show very clearly that the mixing vanes have a significant effect on both the DNB azimuthal location and the CHF value

Assessment of a non-uniform heat flux correction model to predicting CHF in PWR rod bundles

International Nuclear Information System (INIS)

Dae-Hyun, Hwang; Sung-Quun, Zee

2001-01-01

The full text follows. The prediction of CHF (critical heat flux) has been, in most cases, based on the empirical correlation. For PWR fuel assemblies the local parameter correlation requires the local thermal-hydraulic conditions usually calculated by a subchannel analysis code. The cross-sectional averaged fluid conditions of the subchannel, however, are not sufficient for determining CHF, especially for the cases of non-uniform axial heat flux distributions. Many investigators have studied the effect of the upstream heat flux on the CHF. In terms of the upstream memory effect, two different approaches have been considered as the limiting cases. The 'local conditions' hypothesis assumes that there is a unique relationship between the CHF and the local thermal-hydraulic conditions, and consequently there is no memory effect. In the 'overall power' hypothesis, on the other hand, it is assumed that the total power which can be fed into the tube with nonuniform heating will be the same as that for a uniformly heated tube of the same heated length with the same inlet conditions. Thus the CHF is totally influenced by the upstream heat flux distribution. In view of some experimental investigations such as the DeBortoli's test, it revealed that the two approaches are inadequate in general. It means that the local critical heat flux may be affected to some extent by the heat flux distribution upstream of the CHF location. Some correction-factor models have been suggested to take into account the upstream memory effect. Typically, Tong devised a correction factor on the basis of the heat balance of the superheated liquid layer that is spread underneath a highly viscous bubbly layer along the heated surface. His physical model suggested that the fluid enthalpy obtained from an energy balance of the superheated liquid layer is a representative quantity for the onset of DNB (departure nucleate boiling). A theoretically based correction factor model has been proposed by the

Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report; WTZ mit Russland. Transientenanalysen fuer wassergekuehlte Kernreaktoren. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Rohde, Ulrich [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Kozmenkov, Yaroslav [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Institute of Physics and Power Engineering, Obninsk (Russian Federation); Pivovarov, Valeri; Matveev, Yurij [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

2010-12-15

The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

Methodology for verification of heat transfer crisis in the nuclear fuel assemblies

International Nuclear Information System (INIS)

Sharaevsky, I. G.; Sharaevskaya, E. I.; Domashev, E. D.; Arkhypov, A. P.; Kolochko, V. N.

2003-01-01

Reliable operation of water-water type nuclear energy units and design of new generation reactors are not to be provided with wide application of best estimate ThermalHydraulic (TH) codes. It is accepted to consider that up-to-date versions of the codes are featured not only by wide range of NPPs equipment modeling and high ergonomic characteristics of realized in the codes interfaces but comprehensive substantiation of its governing component viz correlations and closure relations systems The pointed correlations and closure relations provide mathematical restraint of the main differential equations system which are necessary for adequate description of the main classes of two-phase flow TH regimes. The principal fact is that without physically justificated correlations and adequate closure relations first of all concerning heat transfer crisis at boiling (DNB) the acceptable reliability of numerical solutions cannot be guaranteed by the codes. But the significant part of realized in the codes correlations mainly on heat transfer crisis are based on the experimental data obtained more than 30 years ago for cylindrical channels. It is known that for TH reliability calculations of the WWERs core with rod fuel elements, such correlations can be applied with caution as it give significantly conservative values of critical heat flux especially at under pressure accident regimes. Moreover because of irregularity of the flow TH parameters on fuel rod elements cross-section distribution the heat transfer crisis regimes are originated only in separate 'hot' cells. Additionally it should be underlined that realized in the codes correlations and closure relations do not consider possibility occurring in the steam generating channels high frequency oscillation instability which poses a threat to the reactor safety. The high frequency oscillations can bring to the fuel elements destruction at heat fluxes much less than the critical ones. Now this type of oscillation

Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report

International Nuclear Information System (INIS)

Rohde, Ulrich; Pivovarov, Valeri; Matveev, Yurij

2010-12-01

The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

Gas Phase Vibrational Spectroscopy of Weakly Volatil Safe Taggants Using a Synchrotron Source

Science.gov (United States)

Cuisset, Arnaud; Hindle, Francis; Mouret, Gael; Gruet, Sebastien; Pirali, Olivier; Roy, Pascale

2013-06-01

The high performances of the AILES beamline of SOLEIL allow to study at medium resolution (0.5 cm^{-1}) the gas phase THz vibrational spectra of weakly volatil compounds. Between 2008 and 2010 we recorded and analyzed the THz/Far-IR spectra of phosphorous based nerve agents thanks to sufficient vapour pressures from liquid samples at room temperature. Recently, we extended these experiments towards the vibrational spectroscopy of vapour pressures from solid samples. This project is quite challenging since we target lower volatile compounds, and so requires very high sensitive spectrometers. Moreover a specially designed heated multipass-cell have been developped for the gas phase study of very weak vapor pressures. Thanks to skills acquired during initial studies and recent experiments performed on AILES with solid PAHs, we have recorded and assigned the gas phase vibrational fingerprints from the THz to the NIR spectral domain (10-4000 cm-1) of a set of targeted nitro-derivatives. The study was focused onto the para, ortho-mononitrotoluene (p-NT, o-NT), the 1,4 Dinitrobenzene (1,4 DNB), the 2,3-dimethyl-2,3-dinitrobutane (DMNB), and 2,4 and 2,6-dinitrotoluene (2,4-2,6 DNT), which are safe taggants widely used for the detection of commercial explosives. These taggants are usually added to plastic explosives in order to facilitate their vapour detection. Therefore, there is a continuous interest for their detection and identification in realistic conditions via optical methods. A first step consists in the recording of their gas phase vibrational spectra. These expected spectra focused onto molecules involved into defence and security domains are not yet available to date and will be very useful for the scientific community. This work is supported by the contract ANR-11-ASTR-035-01. A. Cuisset, G. Mouret, O. Pirali, P. Roy, F. Cazier, H. Nouali, J. Demaison, J. Phys. Chem. B, 2008, 112:, 12516-12525 I. Smirnova, A. Cuisset, R. Bocquet, F. Hindle, G. Mouret, O

Health status of and health-care provision to asylum seekers in Germany: protocol for a systematic review and evidence mapping of empirical studies.

Science.gov (United States)

Schneider, Christine; Mohsenpour, Amir; Joos, Stefanie; Bozorgmehr, Kayvan

2014-11-29

There are more than 100,000 asylum seekers registered in Germany, who are granted limited access to health services. This study aims to provide a systematic overview of the empirical literature on the health status of and health-care provision to asylum seekers in Germany in order to consolidate knowledge, avoid scientific redundance, and identify research gaps. A systematic review and evidence mapping of empirical literature on the health status of and health-care provision to asylum seekers in Germany will be performed. We will apply a three-tiered search strategy: 1. search in databases (PubMed/MEDLINE, Web of Science, IBSS, Sociological Abstracts, Worldwide Political Science Abstracts, CINAHL, Sowiport, Social Sciences Citation Index, ASSIA, MedPilot, DNB), dissertation and theses databases, and the internet (Google); 2. screening references of included studies; 3. contacting authors and civil society organizations for grey literature. Included will be studies which report quantitative and/or qualitative data or review articles on asylum seekers in Germany, published in German or English language. Outcome measures will include physical, mental, or social well-being, and all aspects of health-care provision (access, availability, affordability, and quality). Search results will be screened for eligibility by screening titles, abstracts and full texts. Data extraction comprises information on study characteristics, research aims, and domains of health or health-care services analyzed. The quality of studies will be appraised and documented by appropriate assessment tools. A descriptive evidence map will be drawn by categorizing all included articles by research design and the health conditions and/or domains of health-care provision analyzed. The body of evidence will be evaluated, and a narrative evidence synthesis will be performed by means of a multi-level approach, whereby quantitative and qualitative evidence are analyzed as separate streams and the product

A Scientometric Visualization Analysis for Night-Time Light Remote Sensing Research from 1991 to 2016

Directory of Open Access Journals (Sweden)

Kai Hu

2017-08-01

Full Text Available In this paper, we conducted a scientometric analysis based on the Night-Time Light (NTL remote sensing related literature datasets retrieved from Science Citation Index Expanded and Social Science Citation Index in Web of Science core collection database. Using the methods of bibliometric and Social Network Analysis (SNA, we drew several conclusions: (1 NTL related studies have become a research hotspot, especially after 2011 when the second generation of NTL satellites, the Suomi National Polar-orbiting Partnership (S-NPP Satellite with the Visible Infrared Imaging Radiometer Suite (VIIRS sensor was on board. In the same period, the open-access policy of the long historical dataset of the first generation satellite Defense Meteorological Satellite Program’s Operational Linescan System (DMSP/OLS started. (2 Most related studies are conducted by authors from USA and China, and the USA takes the lead in the field. We identified the biggest research communities constructed by co-authorships and the related important authors and topics by SNA. (3 By the visualization and analysis of the topic evolution using the co-word and co-cited reference networks, we can clearly see that: the research topics change from hardware oriented studies to more real-world applications; and from the first generation of the satellite DMSP/OLS to the second generation of satellite S-NPP. Although the Day Night Band (DNB of the S-NPP exhibits higher spatial and radiometric resolution and better calibration conditions than the first generation DMSP/OLS, the longer historical datasets in DMSP/OLS are still important in long-term and large-scale human activity analysis. (4 In line with the intuitive knowledge, the NTL remote sensing related studies display stronger connections (such as interpretive frame, context, and academic purpose to the social sciences than the general remote sensing discipline. The citation trajectories are visualized based on the dual-maps, thus the

Physical validation issue of the NEPTUNE two-phase modelling: validation plan to be adopted, experimental programs to be set up and associated instrumentation techniques developed

International Nuclear Information System (INIS)

Pierre Peturaud; Eric Hervieu

2005-01-01

, successively in close connection with the in-PWR core DNB and PTS applications, and (ii) high priority experimental programs, with respect to porous medium multi-field and interfacial area transport, in connection with LB-LOCA application. These experimental programs will require the use of specific instrumentation to provide local characteristics of both liquid and vapour phases, such as local void fraction, local interfacial area concentration and local liquid velocity. To fulfill these needs, the following measurement techniques have respectively been developed/enhanced and assessed: X-ray tomography, 4-sensor optical probes and hot-film anemometry. (authors)

Experimental determination of trace-element partitioning between pargasite and a synthetic hydrous andesitic melt

Science.gov (United States)

Brenan, J. M.; Shaw, H. F.; Ryerson, F. J.; Phinney, D. L.

1995-10-01

In order to more fully establish a basis for quantifying the role of amphibole in trace-element fractionation processes, we have measured pargasite/silicate melt partitioning of a variety of trace elements (Rb, Ba, Nb, Ta, Hf, Zr, Ce, Nd, Sm, Yb), including the first published values for U, Th and Pb. Experiments conducted at 1000°C and 1.5 GPa yielded large crystals free of compositional zoning. Partition coefficients were found to be constant at total concentrations ranging from ˜ 1 to > 100 ppm, indicating Henry's Law is oparative over this interval. Comparison of partition coefficients measured in this study with previous determinations yields good agreement for similar compositions at comparable pressure and temperature. The compatibility of U, Th and Pb in amphibole decreases in the order Pb > Th > U. Partial melting or fractional crystallization of amphibole-bearing assemblages will therefore result in the generation of excesses in 238U activity relative to 230Th, similar in magnitude to that produced by clinopyroxene. The compatibility of Pb in amphibole relative to U or Th indicates that melt generation in the presence of residual amphibole will result in the long-term enrichment in Pb relative to U or Th in the residue. This process is therefore incapable of producing the depletion in Pb relative to U or Th inferred from the Pb isotopic composition of MORB and OIB. Comparison of partition coefficients measured in this study with previous values for clinopyroxene allows some distinction to be made between expected trace-element fractionations produced during dry (cpx present) and wet (cpx + amphibole present) melting. Rb, Ba, Nb and Ta are dramatically less compatible in clinopyroxene than in amphibole, whereas Th, U, Hf and Zr have similar compatibilities in both phases. Interelement fractionations, such as DNb/DBa are also different for clinopyroxene and amphibole. Changes in certain ratios, such as Ba/Nb, Ba/Th, and Nb/Th within comagmatic suites may

A Validation of Subchannel Based CHF Prediction Model for Rod Bundles

International Nuclear Information System (INIS)

Hwang, Dae-Hyun; Kim, Seong-Jin

2015-01-01

A large number of CHF data base were procured from various sources which included square and non-square lattice test bundles. CHF prediction accuracy was evaluated for various models including CHF lookup table method, empirical correlations, and phenomenological DNB models. The parametric effect of the mass velocity and unheated wall has been investigated from the experimental result, and incorporated into the development of local parameter CHF correlation applicable to APWR conditions. According to the CHF design criterion, the CHF should not occur at the hottest rod in the reactor core during normal operation and anticipated operational occurrences with at least a 95% probability at a 95% confidence level. This is accomplished by assuring that the minimum DNBR (Departure from Nucleate Boiling Ratio) in the reactor core is greater than the limit DNBR which accounts for the accuracy of CHF prediction model. The limit DNBR can be determined from the inverse of the lower tolerance limit of M/P that is evaluated from the measured-to-predicted CHF ratios for the relevant CHF data base. It is important to evaluate an adequacy of the CHF prediction model for application to the actual reactor core conditions. Validation of CHF prediction model provides the degree of accuracy inferred from the comparison of solution and data. To achieve a required accuracy for the CHF prediction model, it may be necessary to calibrate the model parameters by employing the validation results. If the accuracy of the model is acceptable, then it is applied to the real complex system with the inferred accuracy of the model. In a conventional approach, the accuracy of CHF prediction model was evaluated from the M/P statistics for relevant CHF data base, which was evaluated by comparing the nominal values of the predicted and measured CHFs. The experimental uncertainty for the CHF data was not considered in this approach to determine the limit DNBR. When a subchannel based CHF prediction model

Dinitrobenzamide mustard prodrugs - hypoxic cytotoxins and dual substrates for E.coli nitroreductase

International Nuclear Information System (INIS)

Patterson, A.V.; Hogg, A.; Pullen, S.; Degenkolbe, A.; Li, D.; Chappell, A.; Ying, S.; Atwell, G.J.; Denny, W.A.; Anderson, R.F.; Wilson, W.R.

2003-01-01

Conditional replicating adenoviral vectors (CRAds) have received considerable attention as therapeutic tools in combination with radiotherapy. Viral distribution and micro-regional geometry are likely to be important issues in the treatment of human solid tumours with gene therapy, particularly following intravenous virus administration. The use of CRAds that are 'armed' with enzyme/prodrug systems may overcome some of the perceived limitations; CRAds can redistribute and self-amplify in a cytolytic fashion whilst prodrug metabolites may elicit a local bystander effect. Either or both of these cytotoxic properties could have favourable interactions with radiotherapy (IR). Nevertheless, they may be insufficient to avoid pockets of vector-naive tumour cells beyond the diffusion limits of cytotoxic prodrug metabolites, such as when perivascular seeding occurs. Under such circumstances hypoxic tumour cells may represent the least accessible compartment for vector transfection; the same tumour subpopulation that is likely to be radioresistant. E.coli nitroreductase (NTR) can bioactivate dinitrobenzamide mustards (DNBMs) and is a promising enzyme/prodrug system for 'arming' CRAds. Notably DNMBs can also be activated by endogenous human reductases under low oxygen conditions providing an opportunity to identify dual hypoxic cytotoxins/NTR substrates that may circumvent some of the geometry issues and provide complementarity with IR. To identify a prodrug for NTR that is also active as a hypoxic cytotoxin in vivo. From a set of 164 DNB prodrugs, 19 with favourable activity in vitro against a panel of four NTR-expressing cancer cells were selected and screened for activity as hypoxic cytotoxins in vitro. Measured E17 values ranged from -444 to -366 mV. Seven DNBMs possessed acceptable hypoxic selectivity against the human NSCLC cell line A549WT or clones engineered to overexpress either a human single-electron reductase, cytochrome P450 reductase (A549P450R), or oxic

CFD simulation and validation of turbulent mixing in a rod bundle with vaned spacer grids based on LDV test

International Nuclear Information System (INIS)

Chen Xi; Li Songwei; Li Zhongchun; Du Sijia; Zhang Yu; Peng Huanhuan

2017-01-01

Spacer grids with mixing vanes are generally used in fuel assemblies of Pressurized Water Reactor (PWR), because that mixing vanes could enhance the lateral turbulent mixing in subchannels. Thus, heat exchangements are more efficient, and the value of departure from nucleate boiling (DNB) is greatly increased. Actually turbulent mixing is composed of two kinds of flows: swirling flow inside the subchannel and cross flow between subchannels. Swirling flow could induce mixing between hot water near the rod and cold water in the center of the subchannel, and may accelerate deviation of the bubbles from the rod surface. Besides, crossing flow help to mixing water between hot subchannels and cold subchannels, which impact relatively large flow area. As a result, how to accurately capture and how to predict the complicated mixing phenomenon are of great concernments. Recently many experimental studies has been conducted to provide detailed turbulent mixing in rod bundle, among which Laser Doppler Velocimetry method is widely used. With great development of Computational Fluid Dynamics, CFD has been validated as an analysis method for nuclear engineering, especially for single phase calculation. This paper presents the CFD simulation and validation of the turbulent mixing induced by spacer grid with mixing vanes in rod bundles. Experiment data used for validation came from 5 x 5 rod bundle test with LDV technology, which is organized by Science and Technology on Reactor System Design Technology Laboratory. A 5 x 5 rod bundle with two spacer grids were used. Each rod has dimension of 9.5 mm in outer diameter and distance between rods is 12.6 mm. Two axial bulk velocities were conducted at 3.0 m/s for high Reynolds number and 1.0 m/s for low Reynolds number. Working pressure was 1.0 bar, and temperature was about 25degC. Two different distances from the downstream of the mixing spacer grid and one from upstream were acquired. Mean axial velocities and turbulent intensities

Flow regimes and mechanistic modeling of critical heat flux under subcooled flow boiling conditions

Science.gov (United States)

Le Corre, Jean-Marie

the post-DNB heater temperature up to the point of heater melting. Validation of the proposed model was performed using detailed measured wall boiling parameters near CHF, thereby bypassing most needed constitutive relations. It was found that under limiting nucleation conditions; a peak wall temperature at the time of bubble departure can be reached at CHF preventing wall cooling by quenching. The simulations show that the resulting dry patch can survive the surrounding quenching event, preventing further nucleation and leading to a fast heater temperature increase. For more practical applications, the model was applied at known CHF conditions in simple geometry coupled with one-dimensional and three-dimensional (CFD) codes. It was found that, in the case where CHF occurs under bubbly flow conditions, the local wall superheat underneath nucleating bubbles is predicted to reach the Leidenfrost temperature. However, a better knowledge of statistical variations in wall boiling parameters would be necessary to correctly capture the CHF trends with mass flux (or Weber number). In addition, consideration of relevant parameter influences on the Leidenfrost temperature and consideration of interfacial microphysics at the wall would allow improved simulation of the wall rewetting prevention and subsequent dry patch spreading.

A User Guide to PARET/ANL

Energy Technology Data Exchange (ETDEWEB)

Olson, A. P. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Marin-Lafleche, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-01-01

first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.

Development of real-time core monitoring system models with accuracy-enhanced neural network and its application

International Nuclear Information System (INIS)

Koo, Bon Hyun

1994-02-01

conservative prediction for the wide range of core DNBR while the fine weight set can provide a more accurate yet conservative prediction of core DNBR over a specified DNBR range. With this scheme, the DNBR can be predicted more accurately when the core DNBR approaches the safety setpoint. The weighted system error backpropagation method has also been found to be very effective for the accuracy enhancement and can have a variety of applications if proper weighted function is chosen. For actual applications the uncertainty factor as a function of output was introduced to provide the conservative predictions. A sensitivity analysis was then peformed by taking the partial derivative of the DNBR with respect to the core power to secure a sensitivity coefficient. This coefficient was used to retrieve the available power margin which means how far the core power is away from the DNB occurrence. Based on this analysis, it can be concluded that the backpropagation network training algorithm can be used as a tool for the prediction of the core safety parameters and that the developed system models using the neural network can be used for the accurate yet conservative prediction of the core HCF and DNBR on a real-time basis

Critical heat flux near the critical pressure in heater rod bundle cooled by R-134A fluid: Effects of unheated rods and spacer grid

International Nuclear Information System (INIS)

Chun, Se-Y.; Shin, C.W.; Hong, S. D.; Moon, S. K.

2007-01-01

the case of the DNB (departure from nucleate boiling) at normal pressure conditions. The existence of a threshold pressure at which the CHF phenomenon disappears has been observed near the critical pressure. In the region where the pressure passes across the threshold pressure, CHF does not occur and the wall temperature variations increase monotonously according to the power level applied to the heater rods. The effects of unheated rods and spacer grid with mixing vane on the critical power have been investigated. The effect of unheated rods in the rod bundle on the critical power becomes smaller as the pressure approaches the critical pressure, and when the pressure exceeds 3.9 MPa, the unheated rods have little effect on the critical power. In the case of the rod bundle with the mixing vane spacer grids, the critical power shows larger value compared to that for the spacer grids without mixing vane. This trend is kept up to the pressure of 4.0 MPa (P/Pc =9.85) very close to the critical pressure

High-resolution flow structure measurements in a rod bundle

Energy Technology Data Exchange (ETDEWEB)

Ylönen, A. T.

2013-07-01

Flow behaviour inside a rod bundle has been an active research topic since the early days of the nuclear power industry. Of particular interest in previous studies have been topics such as flow mixing, two-phase flow structure and mapping of two-phase flow transitions. The optimisation of fuel element design can only be achieved by truly understanding the nature of flow. The ultimate goal in this research is to enhance the heat transfer and increase the critical heat flux, which would improve the fuel economy. A better understanding of the flow would also improve nuclear safety as departure from nucleate boiling (DNB) can be predicted more accurately. The motivation for the current project (SUBFLOW) was to increase knowledge of the complex flow phenomena inside a rod bundle. A dedicated sub-channel flow test facility was designed and constructed at the Paul Scherrer Institut (PSI), Villigen, Switzerland. An adiabatic test loop has an up-scaled (1:2.6) vertical fuel rod bundle model with a 4 × 4 geometry. For the very first time, the wire-mesh sensor measurement technique was implemented in a rod bundle as two 64×64 conductivity wire-mesh sensors were installed in the upper part of the test section. The measurement technique enables one to study single- and two-phase flow behaviour with high spatial and temporal resolution. The research topics addressed in this thesis cover a wide range of flow conditions with and without a spacer grid in a rod bundle. The experimental campaign was started by studying natural mixing of a passive scalar to characterise the development of turbulent diffusion in an injection sub-channel and, later on, cross-mixing between adjacent sub-channels. The results were also used in comparison with the in-house CFD code PSI-Boil that is being developed at PSI. The code could estimate the mixing inside the sub-channel and the transition to cross-mixing with a good accuracy. As a natural transition, the SUBFLOW experiments were continued by

High-resolution flow structure measurements in a rod bundle

International Nuclear Information System (INIS)

Ylönen, A. T.

2013-01-01

Flow behaviour inside a rod bundle has been an active research topic since the early days of the nuclear power industry. Of particular interest in previous studies have been topics such as flow mixing, two-phase flow structure and mapping of two-phase flow transitions. The optimisation of fuel element design can only be achieved by truly understanding the nature of flow. The ultimate goal in this research is to enhance the heat transfer and increase the critical heat flux, which would improve the fuel economy. A better understanding of the flow would also improve nuclear safety as departure from nucleate boiling (DNB) can be predicted more accurately. The motivation for the current project (SUBFLOW) was to increase knowledge of the complex flow phenomena inside a rod bundle. A dedicated sub-channel flow test facility was designed and constructed at the Paul Scherrer Institut (PSI), Villigen, Switzerland. An adiabatic test loop has an up-scaled (1:2.6) vertical fuel rod bundle model with a 4 × 4 geometry. For the very first time, the wire-mesh sensor measurement technique was implemented in a rod bundle as two 64×64 conductivity wire-mesh sensors were installed in the upper part of the test section. The measurement technique enables one to study single- and two-phase flow behaviour with high spatial and temporal resolution. The research topics addressed in this thesis cover a wide range of flow conditions with and without a spacer grid in a rod bundle. The experimental campaign was started by studying natural mixing of a passive scalar to characterise the development of turbulent diffusion in an injection sub-channel and, later on, cross-mixing between adjacent sub-channels. The results were also used in comparison with the in-house CFD code PSI-Boil that is being developed at PSI. The code could estimate the mixing inside the sub-channel and the transition to cross-mixing with a good accuracy. As a natural transition, the SUBFLOW experiments were continued by

AREVA modeling and predictive capacities to support PWR fuel assembly upgrading

International Nuclear Information System (INIS)

Canat, J. N.; Mollard, P.; Gentet, G.; Uyeda, G.

2008-01-01

optimization. It explains how they have supported AREVA in performing a wide ranging review of the various components resulting from the technologies acquired after the merging of FRAMATOME with Siemens Nuclear activities (now resulting in AREVA NP): AFA 3G TM , HTP TM and Mark B TM . Today, AREVA offers to each reactor worldwide the best suited fuel assembly, fully assessed and featuring upgraded components issued from well proven technologies, such as AGORA fuel assemblies or other very advanced versions of fuel assemblies. In all areas of fuel design, computational codes, methods and modeling have improved. The paper details the main progress achieved in a variety of fields. Special attention will be paid to improvements performed in the thermal hydraulics area where the Computational Fluid Dynamics STAR CD tool has been used in several Fuel Assembly design and development applications. This progress was made possible by recently available computer hardware, allowing cost effective and easy access of parallel computing in cluster arrays. Specific applications such as the optimization of the vane shape of PWR grids for higher DNB performance, the evaluation of pressure loss coefficients for complex components such as the anti debris bottom nozzle, computation of the flow velocity distribution and consequent fuel rod vibration regime will be detailed. As for fuel rod design, 3 D modeling of the PCI phenomenon provides a basis for improvements of industrial 1 D predictive tools with a better level of accuracy in discriminating hardware solutions. The developments of the next product generations are reaping the benefits of all these improvements, paying off in optimized designs and upgraded performance and robustness levels, for the final benefit of the utilities, especially with the prospect of fuel supply to the forthcoming Generation III reactors

Neutronics and thermal-hydraulics analysis of KUHFR

Energy Technology Data Exchange (ETDEWEB)

Woodruff, W L [Argonne National Laboratory, Argonne, IL (United States); Mishima, K [KURRI, Osaka (Japan)

1983-08-01

control rod worth with reduced enrichment has not yet determined, but only a small decrease in worth is expected. These BOL boron poisoned fuels are also used as the fresh fuel feed for the equilibrium fuel cycle studies contained in this report. The first three cases shown have matching cycle lengths in the equilibrium cycle, while the last case has a considerably longer cycle length. These results are similarly reflected in the 'Maximum Cycle Lengths' shown for unpoisoned BOL cores. Thus, the first three case can be considered comparable. The last case might be considered as an option for an extended cycle length design. The cycle length for this case is increased by about 21%. Obviously, by decreasing the uranium density in the fuel meat (to 2.7 g/cm{sup 3}), the cycle length for this design could be reduced to match that of the other cases. Thermal-hydraulic calculations have been carried out in order to study the safety aspects of the use of reduced enrichment uranium fuel for the KUHFR. The calculations were based on what is outlined in the Safety Analysis Report for the KUHFR and also the IAEA Guidebook for the RERTR program. Only a few combinations of hydraulic parameters have been tested because the reactor safety cannot be discussed without any nuclear physics considerations. For example, any variations in fuel coolant channels may change not only flow velocities but also power peaking factors which may affect the assessment of reactor safety. For this reason, the thermal-hydraulic calculations were carried out only for those specific cases on which neutronics analysis has been already performed. Low enriched uranium (LEU) cases are also included in this study as initial feasibility studies for potential conversion. The computer code PLTEMP has been developed to calculate the flow distribution in the core, fuel plate temperatures and DNB heat fluxes.

Introduzione

Directory of Open Access Journals (Sweden)

Mauro Guerrini

2018-01-01

, bringing our rich cultural heritage and our Italian cataloging background and perspective to an international dimension? I have no doubt that working in an international dimension is an imperative and an advantageous choice for all parties concerned. [1] “Towards an international cataloguing code: 10 questions to Barbara Tillett”, by Mauro Guerrini, in: International cataloguing and bibliographic control, IFLA, vol. 34, no. 1 (Jan.-Mar. 2005, p. 18-20. [2] An example of the influence they have been able to exert is the change with regard to the omission of initial articles in Preferred Title for the Work, following AACR tradition and NACO practice.  In response to a DNB proposal (6JSC/Chair/3 http://rda-jsc.org/archivedsite/working2.html#chair-63 RSC relegated the omission to an alternative instruction. [1] The letter is reproduced in: Mauro Guerrini, “Il dibattito in Italia sulle norme di catalogazione per autori dalla Conferenza di Parigi alle RICA: una prima ricognizione”. In: “Il linguaggio della biblioteca. Scritti in onore di Diego Maltese”, edited by Mauro Guerrini. Milano: Editrice Bibliografica, 1996, p. 626-675.