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Sample records for core power density

  1. Three-dimensional Core Design of a Super Fast Reactor with a High Power Density

    International Nuclear Information System (INIS)

    Cao, Liangzhi; Oka, Yoshiaki; Ishiwatari, Yuki; Ikejiri, Satoshi; Ju, Haitao

    2010-01-01

    The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/cm 3 . The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied

  2. Core design and fuel rod analyses of a super fast reactor with high power density

    International Nuclear Information System (INIS)

    Ju, Haitao; Cao, Liangzhi; Lu, Haoliang; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) that is presently researched in a Japanese project. One of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. A preliminary core has an average power density of 158.8W/cc. In this paper, the principle of improving the average power density is studied and the core design is improved. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. This power density is competitive with that of typical Liquid Metal Fast Breeder Reactors (LMFBR). In order to ensure the fuel rod integrity of this core design, the fuel rod behaviors on the normal operating condition are analyzed using FEMAXI-6 code. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are taken from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak(MPP), Maximum Discharge Burnup(MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900degC. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be

  3. Formulation of detector response function to calculate the power density profiles using in-core neutron detectors

    International Nuclear Information System (INIS)

    Ahmed, S. A.; Peter, J. K.; Semmler, W.; Shultis, J. K.

    2007-01-01

    By measuring neutron fluxes at different locations throughout a core, it's possible to derive the power-density profile P k (W cm - 3), at an axial depth z of fuel rod k. Micro-pocket fission detectors (MPFD) have been fabricated to perform such in-core neutron flux measurements. The purpose of this study is to develop a mathematical model to obtain axial power density distributions in the fuel rods from the in-core responses of the MPFDs

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

    Science.gov (United States)

    Shirvan, Koroush

    Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater

  5. Analysis of the methodical component of core power density field calculation error on the basis of Mochovce-1 commissioning tests

    International Nuclear Information System (INIS)

    Brik, A.

    2009-01-01

    In the first decade of June 2008, during the power commissioning of the reactor at the Mochovce NPP unit 1, the experiment with reducing the thermal power of core almost to the balance-of-plant (BOP) needs was performed. After the reactor has operated for seven hours at low power (about 200 220 MW (thermal)), its power was increased (at a rate of about 0.25% of N nom /min) to the initial level, close to 107% (1471 MW). During the experiment, core parameters, which were subsequently used for comparing the measured data with the results of experiment simulation calculations, were recorded in the reactor in-core monitoring system database. Calculated and measured levels of critical concentrations of boric acid were compared, along with power density distributions by fuel elements and assemblies obtained both by the KRUIZ in-core monitoring system and on the basis of calculations simulating reactor operation in accordance with the given core power variation schedule. The final stage consisted of assessing the methodical component of power density micro- and macro-fields calculation error in the core of Mochovce-1 reactor operating with varying load. (author)

  6. Analysis of the methodical component of core power density field calculation error on the basis of Mochovce-1 commissioning tests

    International Nuclear Information System (INIS)

    Brik, A.

    2009-01-01

    In the first decade of June 2008, during the power commissioning of the reactor at Mochovce NPP unit 1, the experiment with reducing the thermal power of core almost to the balance-of-plant needs was performed. After the reactor has operated for seven hours at low power (about 200 220 MW (thermal)), its power was increased (at a rate of about 0.25% of N nom /min) to the initial level, close to 107% (1471 MW). During the experiment, core parameters, which were subsequently used for comparing the measured data with the results of experiment simulation calculations, were recorded in the reactor in-core monitoring system's database. Calculated and measured levels of critical concentrations of boric acid were compared, along with power density distributions by fuel elements and assemblies obtained both by the KRUIZ in-core monitoring system and on the basis of calculations simulating reactor operation in accordance with the given core power variation schedule. The final stage consisted of assessing the methodical component of power density micro- and macro-fields' calculation error in the core of Mochovce-1 reactor operating with varying load. (Authors)

  7. Research on instability design method without occurring boiling transition for hyper ABWR plants of extended core power density

    International Nuclear Information System (INIS)

    Okamoto, T.; Hotta, A.; Ama, T.

    2008-01-01

    The hyper ABWR (Advanced Boiling Water Reactor) project aims to develop an advanced BWR concept that is competitive in the global market with both highly economic and safety features. Expecting plant construction within the coming ten years, a research program for substantiating the basic design of a high core power density ABWR was conducted. By inheriting the conventional ABWR design, it is possible to reduce construction costs. In order to achieve the rated core power of over 1650MWe which is almost equivalent to that of the EPR (European Pressurized Water Reactor), the core power density of ABWR will be up-rated by at least 25%. Three key subjects linked to this target were recognized. They are, (1) fuel design applicable to the high power density core, (2) improvement of the evaluation method for the coupled neutronic and thermal-hydraulic instability under a wider power-flow operating range, and (3) improvement of the steam separator performance under high quality conditions. In this paper, the second subject has been focused on. In the second subject, the uncertainty approach was introduced in the instability analysis where the best-estimate plant simulator was combined with a direct prediction of boiling transition by the sub-channel code. By employing the CSAU like method, a safety evaluation system that enables to include influences of uncertainties has been developed. Based on the correlation between the time margin for reaching the boiling transition under power oscillations and the decay ratio in the power-flow operation map, an automatic power oscillation suppressing system was designed. The set-point for activating suppression mechanisms (i.e. scram or SRI) could be determined based on this correlation. It was proposed that the present conservative acceptance criterion of the deterministic decay ratio can be replaced with a more rational one of the time margin with including uncertainties. (author)

  8. TRIGA research reactors with higher power density

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1994-01-01

    The recent trend in new or upgraded research reactors is to higher power densities (hence higher neutron flux levels) but not necessarily to higher power levels. The TRIGA LEU fuel with burnable poison is available in small diameter fuel rods capable of high power per rod (≅48 kW/rod) with acceptable peak fuel temperatures. The performance of a 10-MW research reactor with a compact core of hexagonal TRIGA fuel clusters has been calculated in detail. With its light water coolant, beryllium and D 2 O reflector regions, this reactor can provide in-core experiments with thermal fluxes in excess of 3 x 10 14 n/cm 2 ·s and fast fluxes (>0.1 MeV) of 2 x 10 14 n/cm 2 ·s. The core centerline thermal neutron flux in the D 2 O reflector is about 2 x 10 14 n/cm 2 ·s and the average core power density is about 230 kW/liter. Using other TRIGA fuel developed for 25-MW test reactors but arranged in hexagonal arrays, power densities in excess of 300 kW/liter are readily available. A core with TRIGA fuel operating at 15-MW and generating such a power density is capable of producing thermal neutron fluxes in a D 2 O reflector of 3 x 10 14 n/cm 2 ·s. A beryllium-filled central region of the core can further enhance the core leakage and hence the neutron flux in the reflector. (author)

  9. Inference of core barrel motion from neutron noise spectral density. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, J.C.; Shahrokhi, F.; Kryter, R.C.

    1977-03-15

    A method was developed for inference of core barrel motion from the following statistical descriptors: cross-power spectral density, autopower spectral density, and amplitude probability density. To quantify the core barrel motion in a typical pressurized water reactor (PWR), a scale factor was calculated in both one- and two-dimensional geometries using forward, variational, and perturbation methods of discrete ordinates neutron transport. A procedure for selection of the proper frequency band limits for the statistical descriptors was developed. It was found that although perturbation theory is adequate for the calculation of the scale factor, two-dimensional geometric effects are important enough to rule out the use of a one-dimensional approximation for all but the crudest calculations. It was also found that contributions of gamma rays can be ignored and that the results are relatively insensitive to the cross-section set employed. The proper frequency band for the statistical descriptors is conveniently determined from the coherence and phase information from two ex-core power range neutron monitors positioned diametrically across the reactor vessel. Core barrel motion can then be quantified from the integral of the band-limited cross-power spectral density of two diametrically opposed ex-core monitors or, if the coherence between the pair is greater than or equal to 0.7, from a properly band-limited amplitude probability density function. Wide-band amplitude probability density functions were demonstrated to yield erroneous estimates for the magnitude of core barrel motion.

  10. Reactor Core Design and Analysis for a Micronuclear Power Source

    Directory of Open Access Journals (Sweden)

    Hao Sun

    2018-03-01

    Full Text Available Underwater vehicle is designed to ensure the security of country sea boundary, providing harsh requirements for its power system design. Conventional power sources, such as battery and Stirling engine, are featured with low power and short lifetime. Micronuclear reactor power source featured with higher power density and longer lifetime would strongly meet the demands of unmanned underwater vehicle power system. In this paper, a 2.4 MWt lithium heat pipe cooled reactor core is designed for micronuclear power source, which can be applied for underwater vehicles. The core features with small volume, high power density, long lifetime, and low noise level. Uranium nitride fuel with 70% enrichment and lithium heat pipes are adopted in the core. The reactivity is controlled by six control drums with B4C neutron absorber. Monte Carlo code MCNP is used for calculating the power distribution, characteristics of reactivity feedback, and core criticality safety. A code MCORE coupling MCNP and ORIGEN is used to analyze the burnup characteristics of the designed core. The results show that the core life is 14 years, and the core parameters satisfy the safety requirements. This work provides reference to the design and application of the micronuclear power source.

  11. Research on Power Output Characteristics of Magnetic Core in Energy Harvesting Devices

    Directory of Open Access Journals (Sweden)

    Rong-Ping GUO

    2014-07-01

    Full Text Available Magnetic core is the dominant factor in the performance of current transformer energy harvesting devices. The power output model of the magnetic core is established and verified through experiments. According to the actual application requirements, the concept of power density is proposed. The relationships of power density to air gap, material and dimension of the magnetic core are analyzed and verified through experiments.

  12. Core power capability verification for PWR NPP

    International Nuclear Information System (INIS)

    Xian Chunyu; Liu Changwen; Zhang Hong; Liang Wei

    2002-01-01

    The Principle and methodology of pressurized water reactor nuclear power plant core power capability verification for reload are introduced. The radial and axial power distributions of normal operation (category I or condition I) and abnormal operation (category II or condition II) are simulated by using neutronics calculation code. The linear power density margin and DNBR margin for both categories, which reflect core safety, are analyzed from the point view of reactor physics and T/H, and thus category I operating domain and category II protection set point are verified. Besides, the verification results of reference NPP are also given

  13. Neutronic studies of the long life core concept: Part 1, Design and performance of 1000 MWe uranium oxide fueled low power density LMR cores

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1987-04-01

    The parametric behavior of some key neutronic performance parameters for low power density LMR cores fueled with uranium oxide is investigated. The results are compared to reference homogeneous and heterogeneous cores with normal fuel management and Pu fueling. It can be concluded that with respect to minimizing the initial fissile mass and thereby economizing on the inventory costs and carrying charges, the superior neutron economy of the LMR fuel cycle is best exploited through normal fuel management with Pu recycling. In the once-through mode the LMR fuel cycle has disadvantages due to a higher fissile inventory and is not competitive with the LWR fuel cycle

  14. Studies on the inhomogeneous core density of a fluidized bed nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Van der Hagen, T.H.J.J.; Van Dam, H.; Hoogenboom, J.E.; Khotylev, V.A. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.; Harteveld, W.; Mudde, R.F.

    1997-12-31

    Results are reported on the expected time dependent core density profile of a fluidized-bed nuclear fission reactor. Core densities have been measured in a test facility by the gamma-transmission technique. Bubble and particle-cluster sizes, positions, velocities and frequencies could be determined. Neutronic studies have been performed on the influence of core voids on reactivity using Monte-Carlo and neutron-transport codes. Fuel-particle importance has been determined. Point-kinetic parameters have been calculated for linking reactivity perturbations to power fluctuations. (author)

  15. Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru [National Research Nuclear University MEPhI (Russian Federation); Pinegin, A. A. [National Research Center Kurchatov Institute (Russian Federation)

    2016-12-15

    The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

  16. Fragmentation of massive dense cores down to ≲ 1000 AU: Relation between fragmentation and density structure

    International Nuclear Information System (INIS)

    Palau, Aina; Girart, Josep M.; Estalella, Robert; Fuente, Asunción; Fontani, Francesco; Sánchez-Monge, Álvaro; Commerçon, Benoit; Hennebelle, Patrick; Busquet, Gemma; Bontemps, Sylvain; Zapata, Luis A.; Zhang, Qizhou; Di Francesco, James

    2014-01-01

    In order to shed light on the main physical processes controlling fragmentation of massive dense cores, we present a uniform study of the density structure of 19 massive dense cores, selected to be at similar evolutionary stages, for which their relative fragmentation level was assessed in a previous work. We inferred the density structure of the 19 cores through a simultaneous fit of the radial intensity profiles at 450 and 850 μm (or 1.2 mm in two cases) and the spectral energy distribution, assuming spherical symmetry and that the density and temperature of the cores decrease with radius following power-laws. Even though the estimated fragmentation level is strictly speaking a lower limit, its relative value is significant and several trends could be explored with our data. We find a weak (inverse) trend of fragmentation level and density power-law index, with steeper density profiles tending to show lower fragmentation, and vice versa. In addition, we find a trend of fragmentation increasing with density within a given radius, which arises from a combination of flat density profile and high central density and is consistent with Jeans fragmentation. We considered the effects of rotational-to-gravitational energy ratio, non-thermal velocity dispersion, and turbulence mode on the density structure of the cores, and found that compressive turbulence seems to yield higher central densities. Finally, a possible explanation for the origin of cores with concentrated density profiles, which are the cores showing no fragmentation, could be related with a strong magnetic field, consistent with the outcome of radiation magnetohydrodynamic simulations.

  17. Fragmentation of massive dense cores down to ≲ 1000 AU: Relation between fragmentation and density structure

    Energy Technology Data Exchange (ETDEWEB)

    Palau, Aina; Girart, Josep M. [Institut de Ciències de l' Espai (CSIC-IEEC), Campus UAB-Facultat de Ciències, Torre C5-parell 2, E-08193 Bellaterra, Catalunya (Spain); Estalella, Robert [Departament d' Astronomia i Meteorologia (IEEC-UB), Institut de Ciències del Cosmos, Universitat de Barcelona, Martí i Franquès, 1, E-08028 Barcelona (Spain); Fuente, Asunción [Observatorio Astronómico Nacional, P.O. Box 112, E-28803 Alcalá de Henares, Madrid (Spain); Fontani, Francesco; Sánchez-Monge, Álvaro [Osservatorio Astrofisico di Arcetri, INAF, Lago E. Fermi 5, I-50125 Firenze (Italy); Commerçon, Benoit; Hennebelle, Patrick [Laboratoire de Radioastronomie, UMR CNRS 8112, École Normale Supérieure et Observatoire de Paris, 24 rue Lhomond, F-75231 Paris Cedex 05 (France); Busquet, Gemma [INAF-Istituto di Astrofisica e Planetologia Spaziali, Area di Recerca di Tor Vergata, Via Fosso Cavaliere 100, I-00133 Roma (Italy); Bontemps, Sylvain [Université de Bordeaux, LAB, UMR 5804, F-33270 Floirac (France); Zapata, Luis A. [Centro de Radioastronomía y Astrofísica, Universidad Nacional Autónoma de México, P.O. Box 3-72, 58090 Morelia, Michoacán (Mexico); Zhang, Qizhou [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); Di Francesco, James, E-mail: palau@ieec.uab.es [Department of Physics and Astronomy, University of Victoria, P.O. Box 355, STN CSC, Victoria, BC, V8W 3P6 (Canada)

    2014-04-10

    In order to shed light on the main physical processes controlling fragmentation of massive dense cores, we present a uniform study of the density structure of 19 massive dense cores, selected to be at similar evolutionary stages, for which their relative fragmentation level was assessed in a previous work. We inferred the density structure of the 19 cores through a simultaneous fit of the radial intensity profiles at 450 and 850 μm (or 1.2 mm in two cases) and the spectral energy distribution, assuming spherical symmetry and that the density and temperature of the cores decrease with radius following power-laws. Even though the estimated fragmentation level is strictly speaking a lower limit, its relative value is significant and several trends could be explored with our data. We find a weak (inverse) trend of fragmentation level and density power-law index, with steeper density profiles tending to show lower fragmentation, and vice versa. In addition, we find a trend of fragmentation increasing with density within a given radius, which arises from a combination of flat density profile and high central density and is consistent with Jeans fragmentation. We considered the effects of rotational-to-gravitational energy ratio, non-thermal velocity dispersion, and turbulence mode on the density structure of the cores, and found that compressive turbulence seems to yield higher central densities. Finally, a possible explanation for the origin of cores with concentrated density profiles, which are the cores showing no fragmentation, could be related with a strong magnetic field, consistent with the outcome of radiation magnetohydrodynamic simulations.

  18. Development of An On-Line, Core Power Distribution Monitoring System

    International Nuclear Information System (INIS)

    Tunc ALdemir; Don Miller; Peng Wang

    2007-01-01

    The objective of the proposed work was to develop a software package that can construct in three-dimensional core power distributions using the signals from constant temperature power sensors distributed in the reactor core. The software developed uses a mode-based state/parameter estimation technique that is particularly attractive when there are model uncertainties and/or large signal noise. The software yields the expected value of local power at the detector locations and points in between, as well as the probability distribution of the local power density

  19. Development of An On-Line, Core Power Distribution Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Tunc ALdemir; Don Miller; Peng Wang

    2007-10-02

    The objective of the proposed work was to develop a software package that can construct in three-dimensional core power distributions using the signals from constant temperature power sensors distributed in the reactor core. The software developed uses a mode-based state/parameter estmation technique that is particularly attractive when there are model uncertainties and/or large signal noise. The software yields the expected value of local power at the detector locations and points in between, as well as the probability distribution of the local power density

  20. Synthesis of Axial Power Distribution Using 5-Level Ex-core Detector in a Core Protection System

    International Nuclear Information System (INIS)

    Koo, Bon-Seung; Lee, Chung-Chan; Zee, Sung-Quun

    2007-01-01

    In ABB-CE digital plants, Core Protection Calculator System (CPCS) is used for a core protection based on several online measured system parameters including 3- level safety grade ex-core detector signals. The CPCS provides four independent channels for the departure from a nucleate boiling ratio (DNBR) and local power density (LPD) trip signals to the reactor protection system. Each channel consists of a core protection calculator (CPC) and a control element assembly calculator (CEAC). The cubic spline synthesis technique has been used in online calculations of the core axial power distributions using 3-level ex-core detector signals in CPC. The pre-determined cubic spline function sets are used depending on the characteristics of the ex-core detector responses. But this method shows large power distribution errors for the extremely skewed axial shapes due to restrictive function sets and an incorrect SAM value. Especially thus situation is worse at a higher burnup. To solve these problems, the cubic spline function sets are improved and it is demonstrated that the axial power shapes can be synthesized more accurately with the new function sets than those of a conventional CPC. In this paper, synthesis of an axial power distribution using a 5-level ex-core detector is described and the axial power distributions are compared between 3-level and 5-level ex-core detector systems

  1. Core-shell N-doped active carbon fiber@graphene composites for aqueous symmetric supercapacitors with high-energy and high-power density

    Science.gov (United States)

    Xie, Qinxing; Bao, Rongrong; Xie, Chao; Zheng, Anran; Wu, Shihua; Zhang, Yufeng; Zhang, Renwei; Zhao, Peng

    2016-06-01

    Graphene wrapped nitrogen-doped active carbon fibers (ACF@GR) of a core-shell structure were successfully prepared by a simple dip-coating method using natural silk as template. Compared to pure silk active carbon, the as-prepared ACF@GR composites exhibit high specific surface area in a range of 1628-2035 m2 g-1, as well as superior energy storage capability, an extremely high single-electrode capacitance of 552.8 F g-1 was achieved at a current density of 0.1 A g-1 in 6 M KOH aqueous electrolyte. The assembled aqueous symmetric supercapacitors are capable of deliver both high energy density and high power density, for instance, 17.1 Wh kg-1 at a power density of 50.0 W kg-1, and 12.2 Wh kg-1 at 4.7 kW kg-1 with a retention rate of 71.3% for ACF@GR1-based supercapacitor.

  2. On-line generation of core monitoring power distribution in the SCOMS couppled with core design code

    International Nuclear Information System (INIS)

    Lee, K. B.; Kim, K. K.; In, W. K.; Ji, S. K.; Jang, M. H.

    2002-01-01

    The paper provides the description of the methodology and main program module of power distribution calculation of SCOMS(SMART COre Monitoring System). The simulation results of the SMART core using the developed SCOMS are included. The planar radial peaking factor(Fxy) is relatively high in SMART core because control banks are inserted to the core at normal operation. If the conventional core monitoring method is adapted to SMART, highly skewed planar radial peaking factor Fxy yields an excessive conservatism and reduces the operation margin. In addition to this, the error of the core monitoring would be enlarged and thus operating margin would be degraded, because it is impossible to precalculate the core monitoring constants for all the control banks configurations taking into account the operation history in the design stage. To get rid of these drawbacks in the conventional power distribution calculation methodology, new methodology to calculate the three dimensional power distribution is developed. Core monitoring constants are calculated with the core design code (MASTER) which is on-line coupled with SCOMS. Three dimensional (3D) power distribution and the several peaking factors are calculated using the in-core detector signals and core monitoring constant provided at real time. Developed methodology is applied to the SMART core and the various core states are simulated. Based on the simulation results, it is founded that the three dimensional peaking factor to calculate the Linear Power Density and the pseudo hot-pin axial power distribution to calculate the Departure Nucleate Boiling Ratio show the more conservative values than those of the best-estimated core design code, and SCOMS adapted developed methodology can secures the more operation margin than the conventional methodology

  3. Development of UCMS for Analysis of Designed and Measured Core Power Distribution

    International Nuclear Information System (INIS)

    Moon, Sang Rae; Hong, Sun Kwan; Yang, Sung Tae

    2009-01-01

    In this study, reactor core loading patterns were determined by calculating and verifying the factors affecting peak power and important core safety variables were reconciled with their design criteria using a newly designed unified core management system. Core loading patterns are designed for quadrant cores under the assumption that the power distribution of the reactor core is the same among symmetric fuel assemblies within the core. Actual core power distributions measured during core operation may differ slightly from their designed data. Reactor engineers monitor these differences between the designed and measured data by performing a surveillance procedure every month according to the technical specification requirements. It is difficult to monitor overall power distribution behavior throughout the assemblies using the current procedure because it requires the reactor engineer to compare the designed data with only the maximum value of the power peaking factor and the relative power density. It is necessary to enhance this procedure to check the primary variables such as core power distribution, because long cycle operation, high burnup, power up-rate, and improved fuel can change the environment in the core. To achieve this goal, a web-based Unified Core Management System (UCMS) was developed. To build the UCMS, a database system was established using reactor design data such as that in the Nuclear Design Report (NDR) and automated core analysis codes for all light water reactor power plants. The UCMS is designed to help reactor engineers to monitor important core variables and core safety margins by comparing the measured core power distribution with designed data for each fuel assembly during the cycle operation in nuclear power plants

  4. Optimal power and distribution control for weakly-coupled-core reactor

    International Nuclear Information System (INIS)

    Oohori, Takahumi; Kaji, Ikuo

    1977-01-01

    A numerical procedure has been devised for obtaining the optimal power and distribution control for a weakly-coupled-core reactor. Several difficulties were encountered in solving this optimization problem: (1) nonlinearity of the reactor kinetics equations; (2) neutron-leakage interaction between the cores; (3) localized power changes occurring in addition to the total power changes; (4) constraints imposed on the states - e.g. reactivity, reactor period. To obviate these difficulties, use is made of the generalized Newton method to convert the problem into an iterative sequence of linear programming problems, after approximating the differential equations and the integral performance criterion by a set of discrete algebraic equations. In this procedure, a heuristic but effective method is used for deriving an initial approximation, which is then made to converge toward the optimal solution. Delayed-neutron one-group point reactor models embodying transient temperature feed-back to the reactivity are used in obtaining the kinetics equations for the weakly-coupled-core reactor. The criterion adopted for determining the optimality is a norm relevant to the deviations of neutron density from the desired trajectories or else to the time derivatives of the neutron density; uniform control intervals are prescribed. Examples are given of two coupled-core reactors with typical parameters to illustrate the results obtained with this procedure. A comparison is also made between the coupled-core reactor and the one-point reactor. (auth.)

  5. A study of HANARO core conversion using high density U-Mo fuel

    International Nuclear Information System (INIS)

    Lee, K.H.; Lee, C.S.; Lee, B.C.; Park, S.J.; Kim, H.; Kim, C.K.

    2002-01-01

    Currently, HANARO is using 3.15gU/cc U3Si/Al as a driver fuel. HANARO has seven vertical irradiation holes in the core region. Three of them including a central trap are located in the inner region of the core and mainly being used for material irradiation tests. Four of them are located in the reflector tank but cooled by primary coolant. They are used for fuel irradiation tests or radioisotope development tests. For minimum core modification using high density U-Mo fuels, no dimension change is assumed in the current fuel rods and the cladding thickness remains the same in this study. The high density U-Mo fuel will have up to about twice the linear uranium loading of a current HANARO driver fuel. Using this high density fuel 8 fuel sites can be replaced with irradiation sites. Three kinds of conceptual cores are considered using 5 gU/cc U-7Mo/Al and 16 gU/cc U-7Mo. The increase of the linear heat generation rate due to the decrease of total fuel length can be overcome by more uniform radial and axial power distribution using different uranium densities and different fuel meat diameters are introduced into those cores. The new core has 4.54 times larger surface-to-volume ratio than the reference core. The core uranium loading, linear heat generation rate, excess reactivity, and control rod worth as well as the neutron spectra are analysed for each core. (author)

  6. Early-type galaxy core phase densities

    International Nuclear Information System (INIS)

    Carlberg, R. G.; Hartwick, F. D. A.

    2014-01-01

    Early-type galaxies have projected central density brightness profile logarithmic slopes, γ', ranging from about 0 to 1. We show that γ' is strongly correlated, r = 0.83, with the coarse grain phase density of the galaxy core, Q 0 ≡ ρ/σ 3 . The luminosity-γ' correlation is much weaker, r = –0.51. Q 0 also serves to separate the distribution of steep core profiles, γ' > 0.5, from shallow profiles, γ' < 0.3, although there are many galaxies of intermediate slope, at intermediate Q 0 , in a volume-limited sample. The transition phase density separating the two profile types is approximately 0.003 M ☉ pc –3 km –3 s 3 , which is also where the relation between Q 0 and core mass shows a change in slope, the rotation rate of the central part of the galaxy increases, and the ratio of the black hole to core mass increases. These relations are considered relative to the globular cluster inspiral core buildup and binary black hole core scouring mechanisms for core creation and evolution. Mass-enhanced globular cluster inspiral models have quantitative predictions that are supported by data, but no single model yet completely explains the correlations.

  7. Method of estimating thermal power distribution of core of BWR type reactor

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1982-01-01

    Purpose: To accurately and rapidly predict the thermal power of the core of a BWR they reactor at load follow-up operating time. Method: A parameter value corrected from a correction coefficient deciding unit and a xenon density distribution value predicted and calculated from a xenon density distributor are inputted to a thermal power distribution predicting devise, the status amount such as coolant flow rate or the like predetermined at this and next high power operating times is substituted for physical model to predict and calculate the thermal power distribution. The status amount of a nuclear reactor at the time of operating in previous high power corresponding to the next high power operation to be predicted is read from the status amount of the reactor stored in time series manner is a reactor core status memory, and the physical model used in the prediction and calculation of the thermal power distribution at the time of next high power operation is corrected. (Sikiya, K.)

  8. 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

  9. Determination of power density distribution of fuel assemblies for research reactor by directly measuring the strontium-91 activities

    International Nuclear Information System (INIS)

    Yuan, Liq-Ji

    1987-01-01

    This work described the investigations of reactor core power peaking and three dimensional power density distribution of present core configuration of Tsing Hua Open-pool reactor (THOR). An experimental program, based on non-destructive fuel gamma scanning of 91 Sr activities, provides the data of fission density distribution for individual fuel pin of four-rod TRIGA-LEU cluster or for MTR-type fuel assembly. The informations are essentially important for the safety of reactor operation and for fuel management especially for the mixed loading with three different types of fuel at present. The relative power peaking values and the power density distribution for present core are discussed. (author)

  10. Study on in-core fuel management for CNP1500 nuclear power plant

    International Nuclear Information System (INIS)

    Li Dongsheng

    2005-10-01

    CNP1500 is a four-loop PWR nuclear power plant with light water as moderator and coolant. The reactor core is composed of 205 AFA-3GXL fuel assemblies. The active core height at cold is 426.4 cm and equivalent diameter is 347.0 cm. The reactor thermal output is 4250 MW, and average linear power density is 179.5 W/cm. The cycle length of equilibrium cycle core is 470 equivalent full power days. For all cycles, the moderator temperature coefficients at all conditions are negative values, the nuclear enthalpy rise factors F ΔH at hot full power, all control rods out and equilibrium xenon are less than the limit value, the maximum discharge assembly burnup is less 55000 MW·d/tU, and the shutdown margin values at the end of life meet design criteria. The low-leakage core loading reduces radiation damage on pressure vessel and is beneficial to prolong use lifetime of it. The in-core fuel management design scheme and main calculation results for CNP1500 nuclear power plant are presented. (author)

  11. Fuel density effect on parameter of reactivity coefficient of the Innovative Research Reactor core

    International Nuclear Information System (INIS)

    Rokhmadi; Tukiran S

    2013-01-01

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research reactor in Indonesia right now is already 25 year old. Therefor, it is needed to design a new research reactor as a alternative called it innovative research reactor (IRR) and then as an exchanger for old research reactor. The aim of this research is to complete RRI core design data as a requirement for design license. Calculation done is to get the RRI core reactivity coefficients with 5 x 5 core configuration and 20 MW of power, has more than 40 days cycle of length. The RRI core reactivity coefficient calculation is done for new U-"9Mo-Al fuel with variation of densities. The calculation is done by using WIMSD-5B and BATAN-FUEL computer codes. The result of calculation for conceptual design showed that the equilibrium RRI core with 5 x 5 configuration, 450 g, 550 g and 700 g of fuel loadings have negative reactivity coefficients of fuel temperature, moderator temperature, void fraction and density of moderator but the values of the reactivities are very variation. This results has met the safety criteria for RRI core conceptual design. (author)

  12. Optimal Design and Analysis of the Stepped Core for Wireless Power Transfer Systems

    Directory of Open Access Journals (Sweden)

    Xiu Zhang

    2016-01-01

    Full Text Available The key of wireless power transfer technology rests on finding the most suitable means to improve the efficiency of the system. The wireless power transfer system applied in implantable medical devices can reduce the patients’ physical and economic burden because it will achieve charging in vitro. For a deep brain stimulator, in this paper, the transmitter coil is designed and optimized. According to the previous research results, the coils with ferrite core can improve the performance of the wireless power transfer system. Compared with the normal ferrite core, the stepped core can produce more uniform magnetic flux density. In this paper, the finite element method (FEM is used to analyze the system. The simulation results indicate that the core loss generated in the optimal stepped ferrite core can reduce about 10% compared with the normal ferrite core, and the efficiency of the wireless power transfer system can be increased significantly.

  13. BRIGHTEST CLUSTER GALAXIES AND CORE GAS DENSITY IN REXCESS CLUSTERS

    International Nuclear Information System (INIS)

    Haarsma, Deborah B.; Leisman, Luke; Donahue, Megan; Bruch, Seth; Voit, G. Mark; Boehringer, Hans; Pratt, Gabriel W.; Pierini, Daniele; Croston, Judith H.; Arnaud, Monique

    2010-01-01

    We investigate the relationship between brightest cluster galaxies (BCGs) and their host clusters using a sample of nearby galaxy clusters from the Representative XMM-Newton Cluster Structure Survey. The sample was imaged with the Southern Observatory for Astrophysical Research in R band to investigate the mass of the old stellar population. Using a metric radius of 12 h -1 kpc, we found that the BCG luminosity depends weakly on overall cluster mass as L BCG ∝ M 0.18±0.07 cl , consistent with previous work. We found that 90% of the BCGs are located within 0.035 r 500 of the peak of the X-ray emission, including all of the cool core (CC) clusters. We also found an unexpected correlation between the BCG metric luminosity and the core gas density for non-cool-core (non-CC) clusters, following a power law of n e ∝ L 2.7±0.4 BCG (where n e is measured at 0.008 r 500 ). The correlation is not easily explained by star formation (which is weak in non-CC clusters) or overall cluster mass (which is not correlated with core gas density). The trend persists even when the BCG is not located near the peak of the X-ray emission, so proximity is not necessary. We suggest that, for non-CC clusters, this correlation implies that the same process that sets the central entropy of the cluster gas also determines the central stellar density of the BCG, and that this underlying physical process is likely to be mergers.

  14. Determination of power peak factor using control rods, ex-core detectors and neural networks

    International Nuclear Information System (INIS)

    Souza, Rose Mary Gomes do Prado

    2005-01-01

    This work presents a methodology based on the artificial neural network technique to predict in real time the power peak factor in a form that can be implemented in reactor protection systems. The neural network inputs were those available in the reactor protection systems, namely, the axial and quadrant power differences obtained from measured ex-core detector signals, and the position of control rods. The response of ex core detector signals was measured in experiments especially performed in the IPEN/MB-01 zero-power reactor. Several reactor states with different power density distribution were obtained by positioning the control rods in different configurations. The power distribution and its peak factor were calculated for each of these reactor states using the Citation code. The obtained results show that the power peak factor correlates well with the control rod position and the quadrant power difference, and with a lesser degree with the axial power differences. The data presented an inherent organisation and could be classified into different classes of power peak factor behaviour as a function of position of control rods, axial power difference and quadrant power difference. The RBF networks were able to identify classes and interpolate the power peak factor values. The relative error for the power peak factor estimation ranged from 0.19 % to 0.67 %, less than the one that was obtained performing a power density distribution map with in-core detectors. It was observed that the positions of control rods bear the detailed and localised information about the power density distribution, and that the axial and the quadrant power difference describe its global variations in the axial and radial directions. The results showed that the RBF and MLP networks produced similar results, and that a neural network correlation can be implemented in power reactor protection systems. (author)

  15. Safety characteristics of mid-sized MOX fueled liquid metal reactor core of high converter type in the initiating phase of unprotected loss of flow accident. Effect of low specific fuel power density on ULOF behavior brought by employment of large diameter fuel pins

    International Nuclear Information System (INIS)

    Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

    2003-07-01

    Safety characteristics in core disruptive accidents (CDAs) of mid-sized MOX fueled liquid metal reactor core of high converter type have been examined by using the CDA initiating phase analysis code SAS4A. The design concept of high converter type reactor core has been studied as one of options in the category of sodium-cooled reactor in Phase II of Feasibility Study on Commercialized Fast Reactor Cycle System. An unprotected loss-of-flow accident (ULOF) has been selected as a representative CDA initiator for this study. A core concept of high converter type, which employed a large diameter fuel pin of 11.1 mm with 1.2 m core height to get a large fuel volume fraction in the core to achieve high internal conversion ratio was proposed in JFY2001. Each fuel subassembly of the core (abbreviated here as UPL120)was provided with an upper sodium plenum directly above the core to reduce the sodium void reactivity worth. Because of the large fuel pin diameter, average specific fuel power density (31 kW/kg-MOX) of UPL120 is about one half of those of conventional large MOX cores. The reactivity worth of sodium voiding is 6$ in the whole core, and -1$ in the all upper plenums. Initiating phase of ULOF accident in UPL120 under the conditions of nominal design and best estimate analysis resulted in a slightly super-prompt critical power burst. The causes of the super-prompt criticality have been identified twofold: (a) the low specific fuel power density of core reduced the effectiveness of prompt negative reactivity feedback of Doppler and axial fuel expansion effects upon increase in reactor power, and (b) the longer core height compared with conventional 1m cores brought, together with the lower specific power density, a remarkable delay in insertion of negative fuel dispersion reactivity after the onset of fuel disruption in sodium voided subassembly due to the lower linear heat rating in the top portion of the core. During the delay, burst-type fuel failures in sodium un

  16. High-resolution gamma ray attenuation density measurements on mining exploration drill cores, including cut cores

    Science.gov (United States)

    Ross, P.-S.; Bourke, A.

    2017-01-01

    Physical property measurements are increasingly important in mining exploration. For density determinations on rocks, one method applicable on exploration drill cores relies on gamma ray attenuation. This non-destructive method is ideal because each measurement takes only 10 s, making it suitable for high-resolution logging. However calibration has been problematic. In this paper we present new empirical, site-specific correction equations for whole NQ and BQ cores. The corrections force back the gamma densities to the "true" values established by the immersion method. For the NQ core caliber, the density range extends to high values (massive pyrite, 5 g/cm3) and the correction is thought to be very robust. We also present additional empirical correction factors for cut cores which take into account the missing material. These "cut core correction factors", which are not site-specific, were established by making gamma density measurements on truncated aluminum cylinders of various residual thicknesses. Finally we show two examples of application for the Abitibi Greenstone Belt in Canada. The gamma ray attenuation measurement system is part of a multi-sensor core logger which also determines magnetic susceptibility, geochemistry and mineralogy on rock cores, and performs line-scan imaging.

  17. Apparatus for measurement of tree core density

    International Nuclear Information System (INIS)

    Blincow, D.W.

    1975-01-01

    Apparatus is described for direct measurement of the density of a core sample from a tree. A radiation source and detector with a receptacle for the core therebetween, an integrator unit for the detector output, and an indicating meter driven by the integrator unit are described

  18. Modelling of core protection and monitoring system for PWR nuclear power plant simulator

    International Nuclear Information System (INIS)

    Jung Kun Lee; Byoung Sung Han

    1997-01-01

    A nuclear power plant simulator was developed for Younggwang units 3 and 4 nuclear power plant (YGN Nos 3 and 4) in Korea; it has been in operation on training center since November 1996. The core protection calculator (CPC) and the core operating limit supervisory system (COLSS) for the simulator were also developed. The CPC is a digital computer-based core protection system, which performs on-line calculation of departure from nucleate boiling ratio (DNBR) and local power density (LPD). It initiates reactor trip when the core conditions exceed designated DNBR or LPD limitations. The COLSS is designed to assist operators by implementing the limiting conditions for operations in the technical specifications. With these systems, it is possible to increase capacity factor and safety of nuclear power plants, because the COLSS data can show accurate operation margin to plant operators and the CPC can protect reactor core. In this study, the function of CPC/COLSS is analyzed in detail, and then simulation model for CPC/COLSS is presented based on the function. Compared with the YGN Nos 3 and 4 plant operation data and CEDIPS/COLSS FORTRAN code test results, the predictions with the model show reasonable results. (Author)

  19. Density Anomalies in the Mantle and the Gravitational Core-Mantle Interaction

    Science.gov (United States)

    Kuang, Weijia; Liu, Lanbo

    2003-01-01

    Seismic studies suggest that the bulk of the mantle is heterogeneous, with density variations in depth as well as in horizontal directions (latitude and longitude). This density variation produces a three- dimensional gravity field throughout the Earth. On the other hand, the core density also varies in both time and space, due to convective core flow. Consequently, the fluid outer core and the solid mantle interact gravitationally due to the mass anomalies in both regions. This gravitational core-mantle interaction could play a significant role in exchange of angular momentum between the core and the mantle, and thus the change in Earth's rotation on time scales of decades and longer. Aiming at estimating the significance of the gravitational core-mantle interaction on Earth's rotation variation, we introduce in our MoSST core dynamics model a heterogeneous mantle, with a density distribution derived from seismic results. In this model, the core convection is driven by the buoyancy forces. And the density variation is determined dynamically with the convection. Numerical simulation is carried out with different parameter values, intending to extrapolate numerical results for geophysical implications.

  20. Determination of power density in VVER-1000 Mock-Up in LR-0 reactor

    Directory of Open Access Journals (Sweden)

    Košál Michal

    2017-01-01

    Full Text Available The pin power density is an important quantity which has to be monitored during the reactor operation, for two main reasons. Firstly, it is part of the limits and conditions of safe operation and, secondly, it is source term in neutron transport calculations used for the adequate assessing of the state of core structures and pressure vessel material. It is often calculated using deterministic codes which may have problems with an adequate definition of boundary conditions in subcritical regions. This may lead to overestimation of real situation, and therefore the validation of the utility codes contributes not only to better fuel utilization, but also to more precise description of radiation situation in structural components of core. Current paper presents methods developed at LR-0 reactor, as well as selected results for pin power density measurement in peripheral regions of VVER-1000 mock-up. The presented data show that the results of a utility diffusion code at core boundary overestimate the measurement. This situation, however satisfactory safe, may lead to unduly conservative approach in the determination of radiation damage of core structures.

  1. Approach to improve the axial power distribution for the application of a core protection system

    International Nuclear Information System (INIS)

    Koo, Bon Seung; Cho, Jin Young; Song, Jae Seung; Lee, Chung Chan

    2008-01-01

    A Core Protection Calculator System (CPCS) is a digital computer based on a safety system for generating trip signals based on a calculation of the Departure from Nucleate Boiling Ratio (DNBR) and the Local Power Density (LPD) by using several on-line measured system parameters including 3-level ex-core detector signals. A few approaches to improve the axial power distribution for the application of a core protection system were performed. For the Yonggwang unit 3 (cycle 1), axial power distributions were synthesized by applying the cubic spline method and compared with the neutronics code results. Several new cubic spline function sets were generated for the drastically distorted axial shapes for a 3-level ex-core detector system. In addition, synthesized axial shapes with a 5-level ex-core detector signals were compared with the conventional 3-level detector results. It demonstrates that the newly generated function sets appear to be better than that of the conventional CPC from the aspect of an axial power synthesis, particularly for the heavily distorted shapes. Moreover, synthesis of an axial power distribution using 5-level ex-core detector signals appears to be better than that of the 3-level ex-core detector signals. From the above results, improvement of the thermal margin is expected because of an uncertainty decreasing a core protection system. (authors)

  2. Overview of the TITAN-II reversed-field pinch aqueous fusion power core design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Creedon, R.L.; Grotz, S.; Cheng, E.T.; Sharafat, S.; Cooke, P.I.H.

    1988-03-01

    TITAN-II is a compact, high power density Reversed-Field Pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO/sub 3/ and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the Reversed-Field Pinch reactor to operate at a neutron wall loading of 18 MWm/sup 2/. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a passive safety assurance design. 13 refs., 3 figs., 1 tab.

  3. Overview of the TITAN-II reversed-field pinch aqueous fusion power core design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Creedon, R.L.; Cheng, E.T. (General Atomic Co., San Diego, CA (USA)); Grotz, S.P.; Sharafat, S.; Cooke, P.I.H. (California Univ., Los Angeles (USA). Dept. of Mechanical, Aerospace and Nuclear Engineering; California Univ., Los Angeles, CA (USA). Inst. for Plasma and Fusion Research); TITAN Research Group

    1989-04-01

    TITAN-II is a compact, high-power-density Reversed-Field Pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO/sub 3/ and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the Reversed-Field Pinch reactor to operate at a neutron wall loading of 18 MW/m/sup 2/. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a level 2 of passive safety assurance design. (orig.).

  4. O Electromagnetic Power Waves and Power Density Components.

    Science.gov (United States)

    Petzold, Donald Wayne

    1980-12-01

    On January 10, 1884 Lord Rayleigh presented a paper entitled "On the Transfer of Energy in the Electromagnetic Field" to the Royal Society of London. This paper had been authored by the late Fellow of Trinity College, Cambridge, Professor J. H. Poynting and in it he claimed that there was a general law for the transfer of electromagnetic energy. He argued that associated with each point in space is a quantity, that has since been called the Poynting vector, that is a measure of the rate of energy flow per unit area. His analysis was concerned with the integration of this power density vector at all points over an enclosing surface of a specific volume. The interpretation of this Poynting vector as a true measure of the local power density was viewed with great skepticism unless the vector was integrated over a closed surface, as the development of the concept required. However, within the last decade or so Shadowitz indicates that a number of prominent authors have argued that the criticism of the interpretation of Poynting's vector as a local power density vector is unjustified. The present paper is not concerned with these arguments but instead is concerned with a decomposition of Poynting's power density vector into two and only two components: one vector which has the same direction as Poynting's vector and which is called the forward power density vector, and another vector, directed opposite to the Poynting vector and called the reverse power density vector. These new local forward and reverse power density vectors will be shown to be dependent upon forward and reverse power wave vectors and these vectors in turn will be related to newly defined forward and reverse components of the electric and magnetic fields. The sum of these forward and reverse power density vectors, which is simply the original Poynting vector, is associated with the total electromagnetic energy traveling past the local point. Another vector which is the difference between the forward

  5. Experimental core electron density of cubic boron nitride

    DEFF Research Database (Denmark)

    Wahlberg, Nanna; Bindzus, Niels; Bjerg, Lasse

    as well as experimental result. The redistribution of electron density will, if not accounted for, result in increased thermal parameters. It is estimated that 1.7-2 electrons is transferred from boron to nitrogen. [1]: N. Bindzus, T. Straasø, N. Wahlberg, J. Becker, L. Bjerg, N. Lock, A.-C. Dippel, and B......Experimental core electron density of cubic boron nitride Nanna Wahlberg*, Niels Bindzus*, Lasse Bjerg*, Jacob Becker*, and Bo B. Iversen* *Aarhus University, Department of Chemistry, CMC, Langelandsgade 140, 8000 Århus, Denmark The resent progress in powder diffraction provides data of quality...... obtained. The displacement parameters reported here are significantly lower than those previously reported, stressing the importance of an adequate description of the core density. The charge transfer from boron to nitrogen clearly affects the inner electron density, which is evident from theoretical...

  6. Results and interpretation of noise measurements using in-core self powered neutron detector strings at Unit 2 of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Gloeckler, O.; Por, G.; Valko, J.

    1986-11-01

    In-core neutron noise and fuel assembly outlet temperature noise measurements were performed at Unit 2 of Paks Nuclear Power Plant. Characteristics of the reactor and the noise measuring equipment are briefly described. The in-core Rhodium emitter selfpowered neutron detector strings positioned axially above the other show high coherence and linear phase at low frequencies indicating a marked transport effect, not regularly measured in PWRs. The coherence between horizontally placed neutron detectors is small and the phase is zero. A transport effect of different nature is obtained between neutron detectors (in-core and ex-core) and fuel assembly outlet thermocouples. The observed characteristics depend on reactor and fuel assembly power in a way supporting interpretation in terms of coolant density and void content changes and power feedback effects. During routine analysis vibration of 1.1 Hz appeared as a strong peak in the power spectra. The control assembly that was responsible for the observed behaviour could be localized with high certainty. (author)

  7. Investigating the impact of uneven magnetic flux density distribution on core loss estimation

    DEFF Research Database (Denmark)

    Niroumand, Farideh Javidi; Nymand, Morten; Wang, Yiren

    2017-01-01

    is calculated according to an effective flux density value and the macroscopic dimensions of the cores. However, the flux distribution in the core can alter by core shapes and/or operating conditions due to nonlinear material properties. This paper studies the element-wise estimation of the loss in magnetic......There are several approaches for loss estimation in magnetic cores, and all these approaches highly rely on accurate information about flux density distribution in the cores. It is often assumed that the magnetic flux density evenly distributes throughout the core and the overall core loss...

  8. HEDPIN: a computer program to estimate pinwise power density

    International Nuclear Information System (INIS)

    Cappiello, M.W.

    1976-05-01

    A description is given of the digital computer program, HEDPIN. This program, modeled after a previously developed program, POWPIN, provides a means of estimating the pinwise power density distribution in fast reactor triangular pitched pin bundles. The capability also exists for computing any reaction rate of interest at the respective pin positions within an assembly. HEDPIN was developed in support of FTR fuel and test management as well as fast reactor core design and core characterization planning and analysis. The results of a test devised to check out HEDPIN's computational method are given, and the realm of application is discussed. Nearly all programming is in FORTRAN IV. Variable dimensioning is employed to make efficient use of core memory and maintain short running time for small problems. Input instructions, sample problem, and a program listing are also given

  9. Safeguarding subcriticality during loading and shuffling operations in the higher density of the RSG-GAS's silicide core

    International Nuclear Information System (INIS)

    Sembiring, T.M.; Kuntoro, I.

    2003-01-01

    The core conversion program of the RSG-GAS reactor is to convert the all-oxide to all-silicide core. The silicide equilibrium core with fuel meat density of 3.55 gU cm -3 is an optimal core for RSG-GAS reactor and it can significantly increase the operation cycle length from 25 to 32 full power days. Nevertheless, the subcriticality of the shutdown core and the shutdown margin are lower than of the oxide core. Therefore, the deviation of subcriticality condition in the higher silicide core caused by the fuel loading and shuffling error should be reanalysed. The objective of this work is to analyse the sufficiency of the subcriticality condition of the shutdown core to face the worst condition caused by an error during loading and shuffling operations. The calculations were carried out using the 2-dimensional multigroup neutron diffusion code of Batan-FUEL. In the fuel handling error, the calculated results showed that the subcriticality condition of the shutdown higher density silicide equilibrium core of RSG-GAS can be maintained. Therefore, all fuel management steps are fixed in the present reactor operation manual can be applied in the higher silicide equilibrium core of RSG-GAS reactor. (author)

  10. Enhancing power density of biophotovoltaics by decoupling storage and power delivery

    Science.gov (United States)

    Saar, Kadi L.; Bombelli, Paolo; Lea-Smith, David J.; Call, Toby; Aro, Eva-Mari; Müller, Thomas; Howe, Christopher J.; Knowles, Tuomas P. J.

    2018-01-01

    Biophotovoltaic devices (BPVs), which use photosynthetic organisms as active materials to harvest light, have a range of attractive features relative to synthetic and non-biological photovoltaics, including their environmentally friendly nature and ability to self-repair. However, efficiencies of BPVs are currently lower than those of synthetic analogues. Here, we demonstrate BPVs delivering anodic power densities of over 0.5 W m-2, a value five times that for previously described BPVs. We achieved this through the use of cyanobacterial mutants with increased electron export characteristics together with a microscale flow-based design that allowed independent optimization of the charging and power delivery processes, as well as membrane-free operation by exploiting laminar flow to separate the catholyte and anolyte streams. These results suggest that miniaturization of active elements and flow control for decoupled operation and independent optimization of the core processes involved in BPV design are effective strategies for enhancing power output and thus the potential of BPVs as viable systems for sustainable energy generation.

  11. Density functional theory studies of screw dislocation core structures in bcc metals

    DEFF Research Database (Denmark)

    Frederiksen, Søren Lund; Jacobsen, Karsten Wedel

    2003-01-01

    The core structures of (I 11) screw dislocations in bee metals are studied using density functional theory in the local-density approximation. For Mo and Fe, direct calculations of the core structures show the cores to be symmetric with respect to 180degrees rotations around an axis perpendicular...... to symmetric core structures for all the studied metals....

  12. Effect of burnup history by moderator density on neutron-physical characteristics of WWER-1000 core

    International Nuclear Information System (INIS)

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

    2011-01-01

    Results of assessment of burnup history effect by moderator density on neutron physical characteristics of WWER-1000 core are presented on example of stationary fuel loading with Russian design fuel assembly TWSA and AER benchmark for Khmelnitsky NPP that was proposed by TUV and SSTC NRC at nineteenth symposium. Assessment was performed by DYN3D code and cross section library sets generated by HELIOS code. Burnup history was taken into account by preparing of numerous cross section sets with different isotopic composition each of which was obtained by burning under different moderator density. For analysis of history effect 20 cross section sets were prepared for each fuel assembly corresponded to each of 20 axial layers of reactor core model for DYN3D code. Four fuel cycles were modeled both for stationary fuel loading with TWSA and AER benchmark for Khmelnitsky NPP to obtain steady value of error due to neglect of burnup history effect. Main attention of study was paid to effect of burnup history by moderator density to axial power distribution. Results of study for AER benchmark were compared with experimental values of axial power distribution for fuel assemblies of first, second, third and fourth year operation. (Authors)

  13. Determining the axial power profile of partly flooded fuel in a compact core assembled in reactor LR-0

    International Nuclear Information System (INIS)

    Košťál, Michal; Švadlenková, Marie; Baroň, Petr; Rypar, Vojtěch; Milčák, Ján

    2016-01-01

    Highlights: • Fission density in partly flooded compact core. • Calculation of fission density axial profile. • Significant calculational under prediction of experimental axial profile. - Abstract: Measurement and calculation of the axial power profile near the boundary of a moderated and non-moderated core is used to analyze the suitability of the neutron-physical process description, mainly the angular cross-section of a water-moderated uranium system. This is also an important issue because it affects the radiation situation above the partly flooded core of a water-moderated reactor. Axial power profiles of various fuel pins irradiated on reactor LR-0 were measured and the results were compared with MCNP6 code calculations using the ENDF/B-VII.0 nuclear data library. The calculated power profile in positions above the moderator level significantly underestimates experimental results. This might be caused by an improper description of the angular distribution of scattered neutrons in a water-moderated uranium system.

  14. A NEW RECIPE FOR OBTAINING CENTRAL VOLUME DENSITIES OF PRESTELLAR CORES FROM SIZE MEASUREMENTS

    International Nuclear Information System (INIS)

    Tassis, Konstantinos; Yorke, Harold W.

    2011-01-01

    We propose a simple analytical method for estimating the central volume density of prestellar molecular cloud cores from their column density profiles. Prestellar cores feature a flat central part of the column density and volume density profiles of the same size indicating the existence of a uniform-density inner region. The size of this region is set by the thermal pressure force which depends only on the central volume density and temperature of the core, and can provide a direct measurement of the central volume density. Thus, a simple length measurement can immediately yield a central density estimate independent of any dynamical model for the core and without the need for fitting. Using the radius at which the column density is 90% of the central value as an estimate of the size of the flat inner part of the column density profile yields an estimate of the central volume density within a factor of two for well-resolved cores.

  15. Direct measurement of the transition from edge to core power coupling in a light-ion helicon source

    Science.gov (United States)

    Piotrowicz, P. A.; Caneses, J. F.; Showers, M. A.; Green, D. L.; Goulding, R. H.; Caughman, J. B. O.; Biewer, T. M.; Rapp, J.; Ruzic, D. N.

    2018-05-01

    We present time-resolved measurements of an edge-to-core power transition in a light-ion (deuterium) helicon discharge in the form of infra-red camera imaging of a thin stainless steel target plate on the Proto-Material Exposure eXperiment device. The time-resolved images measure the two-dimensional distribution of power deposition in the helicon discharge. The discharge displays a mode transition characterized by a significant increase in the on-axis electron density and core power coupling, suppression of edge power coupling, and the formation of a fast-wave radial eigenmode. Although the self-consistent mechanism that drives this transition is not yet understood, the edge-to-core power transition displays characteristics that are consistent with the discharge entering a slow-wave anti-resonant regime. RF magnetic field measurements made across the plasma column, together with the power deposition results, provide direct evidence to support the suppression of the slow-wave in favor of core plasma production by the fast-wave in a light-ion helicon source.

  16. On-line generation of three-dimensional core power distribution using incore detector signals to monitor safety limits

    International Nuclear Information System (INIS)

    Jang, Jin Wook; Lee, Ki Bog; Na, Man Gyun; Lee, Yoon Joon

    2004-01-01

    It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the Linear Power Density (LPD) and the Departure from Nucleate Boiling Ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER(Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. Through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation

  17. Nuclear detectors for in-core power-reactors

    International Nuclear Information System (INIS)

    Duchene, Jean; Verdant, Robert.

    1979-12-01

    Nuclear reactor control is commonly obtained through neutronic measurements, ex-core and in-core. In large size reactors flux instabilities may take place. For a good monitoring of them, local in-core power measurements become particularly useful. This paper intends to review the questions about neutronic sensors with could be used in-core. A historical account about methods is given first, from early power reactors with brief description of each system. Sensors presently used (ionization fission chambers, self-powered detectors) are then considered and also those which could be developped such as gamma thermometers. Their physical basis, main characteristics and operation modes are detailed. Preliminary tests and works needed for an extension of their life-time are indicated. As an example present irradiation tests at the CEA are then proposed. Two tables will help comparing the characteristics of each type in terms of its precise purpose: fuel monitoring, safety or power control. Finally a table summarizes the kind of sensors mounted on working power reactors and another one is a review of characteristics for some detectors from obtainable commercial sheets [fr

  18. Determination of the in-core power and the average core temperature of low power research reactors using gamma dose rate measurements

    International Nuclear Information System (INIS)

    Osei Poku, L.

    2012-01-01

    Most reactors incorporate out-of-core neutron detectors to monitor the reactor power. An accurate relationship between the powers indicated by these detectors and actual core thermal power is required. This relationship is established by calibrating the thermal power. The most common method used in calibrating the thermal power of low power reactors is neutron activation technique. To enhance the principle of multiplicity and diversity of measuring the thermal neutron flux and/or power and temperature difference and/or average core temperature of low power research reactors, an alternative and complimentary method has been developed, in addition to the current method. Thermal neutron flux/Power and temperature difference/average core temperature were correlated with measured gamma dose rate. The thermal neutron flux and power predicted using gamma dose rate measurement were in good agreement with the calibrated/indicated thermal neutron fluxes and powers. The predicted data was also good agreement with thermal neutron fluxes and powers obtained using the activation technique. At an indicated power of 30 kW, the gamma dose rate measured predicted thermal neutron flux of (1* 10 12 ± 0.00255 * 10 12 ) n/cm 2 s and (0.987* 10 12 ± 0.00243 * 10 12 ) which corresponded to powers of (30.06 ± 0.075) kW and (29.6 ± 0.073) for both normal level of the pool water and 40 cm below normal levels respectively. At an indicated power of 15 kW, the gamma dose rate measured predicted thermal neutron flux of (5.07* 10 11 ± 0.025* 10 11 ) n/cm 2 s and (5.12 * 10 11 ±0.024* 10 11 ) n/cm 2 s which corresponded to power of (15.21 ± 0.075) kW and (15.36 ± 0.073) kW for both normal levels of the pool water and 40 cm below normal levels respectively. The power predicted by this work also compared well with power obtained from a three-dimensional neutronic analysis for GHARR-1 core. The predicted power also compares well with calculated power using a correlation equation obtained from

  19. CRISS power spectral density

    International Nuclear Information System (INIS)

    Vaeth, W.

    1979-04-01

    The correlation of signal components at different frequencies like higher harmonics cannot be detected by a normal power spectral density measurement, since this technique correlates only components at the same frequency. This paper describes a special method for measuring the correlation of two signal components at different frequencies: the CRISS power spectral density. From this new function in frequency analysis, the correlation of two components can be determined quantitatively either they stem from one signal or from two diverse signals. The principle of the method, suitable for the higher harmonics of a signal as well as for any other frequency combinations is shown for the digital frequency analysis technique. Two examples of CRISS power spectral densities demonstrates the operation of the new method. (orig.) [de

  20. Evaluation for the models of neutron diffusion theory in terms of power density distributions of the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Shimakawa, Satoshi; Nojiri, Naoki; Fujimoto, Nozomu

    2003-10-01

    In the case of evaluations for the highest temperature of the fuels in the HTTR, it is very important to expect the power density distributions accurately; therefore, it is necessary to improve the analytical model with the neutron diffusion and the burn-up theory. The power density distributions are analyzed in terms of two models, the one mixing the fuels and the burnable poisons homogeneously and the other modeling them heterogeneously. Moreover these analytical power density distributions are compared with the ones derived from the gross gamma-ray measurements and the Monte Carlo calculational code with continuous energy. As a result the homogeneous mixed model isn't enough to expect the power density distributions of the core in the axial direction; on the other hand, the heterogeneous model improves the accuracy. (author)

  1. Response of a core coherent density oscillation on electron cyclotron resonance heating in Heliotron J plasma

    Science.gov (United States)

    Kobayashi, T.; Kobayashi, S.; Lu, X. X.; Kenmochi, N.; Ida, K.; Ohshima, S.; Yamamoto, S.; Kado, S.; Kokubu, D.; Nagasaki, K.; Okada, H.; Minami, T.; Otani, Y.; Mizuuchi, T.

    2018-01-01

    We report properties of a coherent density oscillation observed in the core region and its response to electron cyclotron resonance heating (ECH) in Heliotron J plasma. The measurement was performed using a multi-channel beam emission spectroscopy system. The density oscillation is observed in a radial region between the core and the half radius. The poloidal mode number is found to be 1 (or 2). By modulating the ECH power with 100 Hz, repetition of formation and deformation of a strong electron temperature gradient, which is likely ascribed to be an electron internal transport barrier, is realized. Amplitude and rotation frequency of the coherent density oscillation sitting at the strong electron temperature gradient location are modulated by the ECH, while the poloidal mode structure remains almost unchanged. The change in the rotation velocity in the laboratory frame is derived. Assuming that the change of the rotation velocity is given by the background E × B velocity, a possible time evolution of the radial electric field was deduced.

  2. Analysis and study on core power capability with margin method

    International Nuclear Information System (INIS)

    Liu Tongxian; Wu Lei; Yu Yingrui; Zhou Jinman

    2015-01-01

    Core power capability analysis focuses on the power distribution control of reactor within the given mode of operation, for the purpose of defining the allowed normal operating space so that Condition Ⅰ maneuvering flexibility is maintained and Condition Ⅱ occurrences are adequately protected by the reactor protection system. For the traditional core power capability analysis methods, such as synthesis method or advanced three dimension method, usually calculate the key safety parameters of the power distribution, and then verify that these parameters meet the design criteria. For PWR with on-line power distribution monitoring system, core power capability analysis calculates the most power level which just meets the design criteria. On the base of 3D FAC method of Westinghouse, the calculation model of core power capability analysis with margin method is introduced to provide reference for engineers. The core power capability analysis of specific burnup of Sanmen NPP is performed with the margin method. The results demonstrate the rationality of the margin method. The calculation model of the margin method not only helps engineers to master the core power capability analysis for AP1000, but also provides reference for engineers for core power capability analysis of other PWR with on-line power distribution monitoring system. (authors)

  3. Relative power density distribution calculations of the Kori unit 1 pressurized water reactor with full-scope explicit modeling of monte carlo simulation

    International Nuclear Information System (INIS)

    Kim, J. O.; Kim, J. K.

    1997-01-01

    Relative power density distributions of the Kori unit 1 pressurized water reactor calculated by Monte Carlo modeling with the MCNP code. The Kori unit 1 core is modeled on a three-dimensional representation of the one-eighth of the reactor in-vessel component with reflective boundaries at 0 and 45 degrees. The axial core model is based on half core symmetry and is divided into four axial segments. Fission reaction density in each rod is calculated by following 100 cycles with 5,000 test neutrons in each cycle after starting with a localized neutron source and ten noncontributing settle cycles. Relative assembly power distributions are calculated from fission reaction densities of rods in assembly. After 100 cycle calculations, the system coverages to a κ value of 1.00039 ≥ 0.00084. Relative assembly power distribution is nearly the same with that of the Kori unit 1 FSAR. Applicability of the full-scope Monte Carlo simulation in the power distribution calculation is examined by the relative root mean square error of 2.159%. (author)

  4. Fission-powered in-core thermoacoustic sensor

    Energy Technology Data Exchange (ETDEWEB)

    Garrett, Steven L. [Graduate Program in Acoustics, Penn State University, University Park, Pennsylvania 16802 (United States); Smith, James A. [Fundamental Fuel Properties, Idaho National Laboratory, Idaho Falls, Idaho 83415 (United States); Smith, Robert W. M. [Applied Research Laboratory, Penn State University, State College, Pennsylvania 16804 (United States); Heidrich, Brenden J. [Nuclear Science User Facilities, Idaho National Laboratory, Idaho Falls, Idaho 83415 (United States); Heibel, Michael D. [Global Technology Development, Westinghouse Electric Company, Cranberry Township, Pennsylvania 16066 (United States)

    2016-04-04

    A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

  5. Lamination effects on a 3D model of the magnetic core of power transformers

    Directory of Open Access Journals (Sweden)

    Poveda-Lerma Antonio

    2017-12-01

    Full Text Available In this paper the lamination effect on the model of a power transformer’s core with stacked E-I structure is analyzed. The distribution of the magnetic flux in the laminations depends on the stacking method. In this work it is shown, using a 3D FEM model and an experimental prototype, that the non-uniform distribution of the flux in a laminated E-I core with alternate-lap joint stack increases substantially the average value of the magnetic flux density in the core, compared with a butt joint stack. Both the simulated model and the experimental tests show that the presence of constructive air-gaps in the E-I junctions gives rise to a zig-zag flux in the depth direction. This inter-lamination flux reduces the magnetic flux density in the I-pieces and increases substantially the magnetic flux density in the E-pieces, with highly saturated points that traditional 2D analysis cannot reproduce. The relation between the number of laminations included in the model, and the computational resourses needed to build it, is also evaluated in this work.

  6. Aerosol core nuclear reactor for space-based high energy/power nuclear-pumped lasers

    International Nuclear Information System (INIS)

    Prelas, M.A.; Boody, F.P.; Zediker, M.S.

    1987-01-01

    An aerosol core reactor concept can overcome the efficiency and/or chemical activity problems of other fuel-reactant interface concepts. In the design of a laser using the nuclear energy for a photon-intermediate pumping scheme, several features of the aerosol core reactor concept are attractive. First, the photon-intermediate pumping concept coupled with photon concentration methods and the aerosol fuel can provide the high power densities required to drive high energy/power lasers efficiently (about 25 to 100 kW/cu cm). Secondly, the intermediate photons should have relatively large mean free paths in the aerosol fuel which will allow the concept to scale more favorably. Finally, the aerosol core reactor concept can use materials which should allow the system to operate at high temperatures. An excimer laser pumped by the photons created in the fluorescer driven by a self-critical aerosol core reactor would have reasonable dimensions (finite cylinder of height 245 cm and radius of 245 cm), reasonable laser energy (1 MJ in approximately a 1 millisecond pulse), and reasonable mass (21 kg uranium, 8280 kg moderator, 460 kg fluorescer, 450 kg laser medium, and 3233 kg reflector). 12 references

  7. Hollow-core fibers for high power pulse delivery

    DEFF Research Database (Denmark)

    Michieletto, Mattia; Lyngsø, Jens K.; Jakobsen, Christian

    2016-01-01

    We investigate hollow-core fibers for fiber delivery of high power ultrashort laser pulses. We use numerical techniques to design an anti-resonant hollow-core fiber having one layer of non-touching tubes to determine which structures offer the best optical properties for the delivery of high power...... picosecond pulses. A novel fiber with 7 tubes and a core of 30 mu m was fabricated and it is here described and characterized, showing remarkable low loss, low bend loss, and good mode quality. Its optical properties are compared to both a 10 mu m and a 18 mu m core diameter photonic band gap hollow......-core fiber. The three fibers are characterized experimentally for the delivery of 22 picosecond pulses at 1032nm. We demonstrate flexible, diffraction limited beam delivery with output average powers in excess of 70W. (C) 2016 Optical Society of America...

  8. The system of the measurement of reactor power and the monitoring of core power distribution

    International Nuclear Information System (INIS)

    Li Xianfeng

    1999-01-01

    The author mainly describes the measurement of the reactor power and the monitoring of the core power distribution in DAYA BAY nuclear power plant, introduces the calibration for the measurement system. Ex-core nuclear instrumentation system (RPN) and LOCA surveillance system (LSS) are the most important system for the object. they perform the measurement of the reactor power and the monitoring of the core power distribution on-line and timely. They also play the important roles in the reactor control and the reactor protection. For the same purpose there are test instrumentation system (KME) and in-core instrumentation system (RIC). All of them work together ensuring the exact measurement and effective monitoring, ensuring the safety of the reactor power plant

  9. High Power Density Power Electronic Converters for Large Wind Turbines

    DEFF Research Database (Denmark)

    Senturk, Osman Selcuk

    . For these VSCs, high power density is required due to limited turbine nacelle space. Also, high reliability is required since maintenance cost of these remotely located wind turbines is quite high and these turbines operate under harsh operating conditions. In order to select a high power density and reliability......In large wind turbines (in MW and multi-MW ranges), which are extensively utilized in wind power plants, full-scale medium voltage (MV) multi-level (ML) voltage source converters (VSCs) are being more preferably employed nowadays for interfacing these wind turbines with electricity grids...... VSC solution for wind turbines, first, the VSC topology and the switch technology to be employed should be specified such that the highest possible power density and reliability are to be attained. Then, this qualitative approach should be complemented with the power density and reliability...

  10. Results from core-edge experiments in high Power, high performance plasmas on DIII-D

    Directory of Open Access Journals (Sweden)

    T.W. Petrie

    2017-08-01

    Full Text Available Significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q⊥P ∝ [PSOL x IP] 0.92 for PSOL= 8−19MW and IP= 1.0–1.4MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-D plasmas may be problematical at high power and H98 (≥ 1.5 due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q⊥P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot” but also that heating near the slot opening is a significant source for impurity contamination of the core.

  11. Use of in-core self-powered detectors in the nuclear reactor control system

    International Nuclear Information System (INIS)

    Mitel'man, M.G.; Andreev, L.G.; Batenin, I.V.

    1975-01-01

    The paper describes the results of experiments conducted at the Obninsk atomic power station on the establishment of an integrated system for the control and automatic regulation of the distribution of energy liberation. The pick-ups used for this system are direct-charge detectors with rhodium emitters. The system, which consists of 12 integral DPZ-7 detectors distributed evenly on 2 mutually perpendicular diamaters and 2 assemblies, each containing 4 1n direct-charge detectors, monitors the distribution of neutron flow density according to core height. Increased signals from the detectors were fed into the summation block, in which the outgoing signal was also adjusted. The half-life of the rhodium was calculated and the non-inertia of the entire system ascertained. The total operating time of the system for the automatic control of reactor power is approximately 1 h. The influence of interference was not determined. During the experiment, the power of the reactor remained constant, i.e. its regulation by the direct-charge detectors inside the core was effective. (author)

  12. Core fueling to produce peaked density profiles in large tokamaks

    International Nuclear Information System (INIS)

    Mikkelsen, D.R.; McGuire, K.M.; Schmidt, G.L.; Zweben, S.J.

    1994-06-01

    Peaking the density profile increases the usable bootstrap current and the average fusion power density; this could reduce the current drive power and increase the net output of power producing tokamaks. The use of neutral beams and pellet injection to produce peaked density profiles is assessed. We show that with radially ''hollow'' diffusivity profiles (and no particle pinch) moderately peaked density profiles can be produced by particle source profiles which are peaked off-axis. The fueling penetration requirements can therefore be relaxed and this greatly improves the feasibility of generating peaked density profiles in large tokamaks. In particular, neutral beam fueling does not require MeV particle energy. Even with beam voltages of ∼200 keV, however, exceptionally good particle confinement, τ p much-gt τ E is required to achieve net electrical power generation. In system with no power production requirement (e.g., neutron sources) neutral beam fueling should be capable of producing peaked density profiles in devices as large as ITER. Fueling systems with low energy cost per particle (such as cryogenic pellet injection) must be used in power producing tokamaks when τ p ∼ τ E . Simulations with pellet injection speeds of 7 km/sec show the peaking factor, n eo /left-angle n e right-angle, approaching 2

  13. A performance analysis for MHD power cycles operating at maximum power density

    International Nuclear Information System (INIS)

    Sahin, Bahri; Kodal, Ali; Yavuz, Hasbi

    1996-01-01

    An analysis of the thermal efficiency of a magnetohydrodynamic (MHD) power cycle at maximum power density for a constant velocity type MHD generator has been carried out. The irreversibilities at the compressor and the MHD generator are taken into account. The results obtained from power density analysis were compared with those of maximum power analysis. It is shown that by using the power density criteria the MHD cycle efficiency can be increased effectively. (author)

  14. Estimation of reactor core calculation by HELIOS/MASTER at power generating condition through DeCART, whole-core transport code

    International Nuclear Information System (INIS)

    Kim, H. Y.; Joo, H. G.; Kim, K. S.; Kim, G. Y.; Jang, M. H.

    2003-01-01

    The reactivity and power distribution errors of the HELIOS/MASTER core calculation under power generating conditions are assessed using a whole core transport code DeCART. For this work, the cross section tablesets were generated for a medium sized PWR following the standard procedure and two group nodal core calculations were performed. The test cases include the HELIOS calculations for 2-D assemblies at constant thermal conditions, MASTER 3D assembly calculations at power generating conditions, and the core calculations at HZP, HFP, and an abnormal power conditions. In all these cases, the results of the DeCART code in which pinwise thermal feedback effects are incorporated are used as the reference. The core reactivity, assemblywise power distribution, axial power distribution, peaking factor, and thermal feedback effects are then compared. The comparison shows that the error of the HELIOS/MASTER system in the core reactivity, assembly wise power distribution, pin peaking factor are only 100∼300 pcm, 3%, and 2%, respectively. As far as the detailed pinwise power distribution is concerned, however, errors greater than 15% are observed

  15. Core radial power profile effect on system and core cooling behavior during reflood phase of PWR-LOCA with CCTF data

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Iguchi, Tadashi; Murao, Yoshio

    1985-01-01

    In the reactor safety assessment during reflood phase of a PWR-LOCA, it is assumed implicitly that the core thermal hydraulic behavior is evaluated by the one-dimensional model with an average power rod. In order to assess the applicability of the one-dimensional treatment, integral tests were performed with various core radial power profiles using the Cylindrical Core Test Facility (CCTF) whose core includes about 2,000 heater rods. The CCTF results confirm that the core radial power profile has weak effect on the thermal hydraulic behavior in the primary system except core. It is also confirmed that the core differential pressure in the axial direction is predicted by the one-dimensional core model with an average power rod even in the case with a steep radial power profile in the core. Even though the core heat transfer coefficient is dependent on the core radial power profile, it is found that the error of the peak clad surface temperature calculation is less than 15 K using the one-dimensional model in the CCTF tests. The CCTF results support the one-dimensional treatment assumed in the reactor safety assessment. (author)

  16. State-space model predictive control method for core power control in pressurized water reactor nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Guo Xu; Wu, Jie; Zeng, Bifan; Wu, Wangqiang; Ma, Xiao Qian [School of Electric Power, South China University of Technology, Guangzhou (China); Xu, Zhibin [Electric Power Research Institute of Guangdong Power Grid Corporation, Guangzhou (China)

    2017-02-15

    A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  17. Power-Energy Simulation for Multi-Core Processors in Bench-marking

    Directory of Open Access Journals (Sweden)

    Mona A. Abou-Of

    2017-01-01

    Full Text Available At Microarchitectural level, multi-core processor, as a complex System on Chip, has sophisticated on-chip components including cores, shared caches, interconnects and system controllers such as memory and ethernet controllers. At technological level, architects should consider the device types forecast in the International Technology Roadmap for Semiconductors (ITRS. Energy simulation enables architects to study two important metrics simultaneously. Timing is a key element of the CPU performance that imposes constraints on the CPU target clock frequency. Power and the resulting heat impose more severe design constraints, such as core clustering, while semiconductor industry is providing more transistors in the die area in pace with Moore’s law. Energy simulators provide a solution for such serious challenge. Energy is modelled either by combining performance benchmarking tool with a power simulator or by an integrated framework of both performance simulator and power profiling system. This article presents and asses trade-offs between different architectures using four cores battery-powered mobile systems by running a custom-made and a standard benchmark tools. The experimental results assure the Energy/ Frequency convexity rule over a range of frequency settings on different number of enabled cores. The reported results show that increasing the number of cores has a great effect on increasing the power consumption. However, a minimum energy dissipation will occur at a lower frequency which reduces the power consumption. Despite that, increasing the number of cores will also increase the effective cores value which will reflect a better processor performance.

  18. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-U.

    Science.gov (United States)

    Faust, I; Delgado-Aparicio, L; Bell, R E; Tritz, K; Diallo, A; Gerhardt, S P; LeBlanc, B; Kozub, T A; Parker, R R; Stratton, B C

    2014-11-01

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed.

  19. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-Ua)

    Science.gov (United States)

    Faust, I.; Delgado-Aparicio, L.; Bell, R. E.; Tritz, K.; Diallo, A.; Gerhardt, S. P.; LeBlanc, B.; Kozub, T. A.; Parker, R. R.; Stratton, B. C.

    2014-11-01

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed.

  20. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Faust, I.; Parker, R. R. [MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Delgado-Aparicio, L.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; LeBlanc, B.; Kozub, T. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Tritz, K. [The Johns Hopkins University, Baltimore, Maryland 21209 (United States); Stratton, B. C. [MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States)

    2014-11-15

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed.

  1. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-U

    International Nuclear Information System (INIS)

    Faust, I.; Parker, R. R.; Delgado-Aparicio, L.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; LeBlanc, B.; Kozub, T. A.; Tritz, K.; Stratton, B. C.

    2014-01-01

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed

  2. Electronic DC transformer with high power density

    NARCIS (Netherlands)

    Pavlovský, M.

    2006-01-01

    This thesis is concerned with the possibilities of increasing the power density of high-power dc-dc converters with galvanic isolation. Three cornerstones for reaching high power densities are identified as: size reduction of passive components, reduction of losses particularly in active components

  3. The Multi一physics Research on I ron一Core Vibration Noise of Power Reactor

    Directory of Open Access Journals (Sweden)

    LI U Ja

    2017-02-01

    Full Text Available On the basis of theoretical research releted to the magnetostriction and maxwell’.s equations,the fi- nite element coupling in the transient electromagnetic field coupling,structure and sound field coupling has been developed In thts paper by using the flnlte element sOftWare CO}IS01., Whleh establish a serles three-phase COT’e re- actor model, to analyzing the power frequency magnetic field distribution,core magnetostrictive displacement,max- well force displacement and sound pressure level of the three-phase series core reactor under the power frequency working state. According to transient magnetic field distribution in the simulation of the reactor,the magnetic flux density distribution inside the reactor and the vibration displacement distribution are calculated,the acoustic field distribution is measured alao. It is shown that physical field simulation results and measured data are basically in consisent by experiment,it is proved multi-physics coupling is an effective method for forecast of noise.

  4. Core fuelling to produce peaked density profiles in large tokamaks

    International Nuclear Information System (INIS)

    Mikkelsen, D.R.; McGuire, K.M.; Schmidt, G.L.; Zweben, S.J.

    1995-01-01

    Peaking the density profile increases the usable bootstrap current and the average fusion power density; this could reduce the current drive power and increase the net output of power producing tokamaks. The use of neutral beams and pellet injection to produce peaked density profiles is assessed. It is shown that with radially 'hollow' diffusivity profiles (and no particle pinch) moderately peaked density profiles can be produced by particle source profiles that are peaked off-axis. The fuelling penetration requirements can therefore be relaxed and this greatly improves the feasibility of generating peaked density profiles in large tokamaks. In particular, neutral beam fuelling does not require Megavolt particle energies. Even with beam voltages of ∼ 200 keV, however, exceptionally good particle confinement is needed to achieve net electrical power generation. The required ratio of particle to thermal diffusivities is an order of magnitude outside the range reported for tokamaks. In a system with no power production requirement (e.g., neutron sources) neutral beam fuelling should be capable of producing peaked density profiles in devices as large as ITER. Fuelling systems with low energy cost per particle - such as cryogenic pellet injection - must be used in power producing tokamaks when τ P ∼ τ E . Simulations with pellet injection speeds of 7 km/s show that the peaking factor, n e0 / e >, approaches 2. (author). 65 refs, 8 figs

  5. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  6. Adaptive control method for core power control in TRIGA Mark II reactor

    Science.gov (United States)

    Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

    2018-01-01

    The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  7. ac power control in the Core Flow Test Loop

    International Nuclear Information System (INIS)

    McDonald, D.W.

    1980-01-01

    This work represents a status report on a development effort to design an ac power controller for the Core Flow Test Loop. The Core Flow Test Loop will be an engineering test facility which will simulate the thermal environment of a gas-cooled fast-breeder reactor. The problems and limitations of using sinusoidal ac power to simulate the power generated within a nuclear reactor are addressed. The transformer-thyristor configuration chosen for the Core Flow Test Loop power supply is presented. The initial considerations, design, and analysis of a closed-loop controller prototype are detailed. The design is then analyzed for improved performance possibilities and failure modes are investigated at length. A summary of the work completed to date and a proposed outline for continued development completes the report

  8. Sensitivity analysis for reactivity and power density investigations in nuclear reactors

    International Nuclear Information System (INIS)

    Naguib, K.; Morcos, H.N.; Sallam, O.H.; Abdelsamei, SH.

    1993-01-01

    Sensitivity analysis theory based on the variational functional approaches was applied to evaluate sensitivities of eigenvalues and power densities due to variation of the absorber concentration in the reactor core. The practical usefulness of this method is illustrated by considering test cases. The result indicates that this method is as accurate as those obtained from direct calculations, yet it provides an economical means in saving computational time since it requires fewer calculations. The SARC-1/2 code have been written in Fortran-77 to solve this problem.3 tab. 1 fig

  9. ADX: a high field, high power density, advanced divertor and RF tokamak

    Science.gov (United States)

    LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

    2015-05-01

    The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

  10. In-core fuel management for the course on operational physics of power reactors

    International Nuclear Information System (INIS)

    Levine, S.H.

    1982-01-01

    The heart of a nuclear power station is the reactor core producing power from the fissioning of uranium or plutonium fuel. Expertise in many different technical fields is required to provide fuel for continuous economical operation of a nuclear power plant. In general, these various technical disciplines can be dichotomized into ''Out-of-core'' and ''In-core'' fuel management. In-core fuel management is concerned, as the name implies, with the reactor core itself. It entails calculating the core reactivity, power distribution, and isotopic inventory for the first and subsequent cores of a nuclear power plant to maintain adequate safety margins and operating lifetime for each core. In addition, the selection of reloading schemes is made to minimize energy costs

  11. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Energy Technology Data Exchange (ETDEWEB)

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  12. BWR power oscillation evaluation methodologies in core design

    International Nuclear Information System (INIS)

    Hotta, Akitoshi

    1995-01-01

    At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

  13. Study on the impact of the engineering energy gain and the FPC mass power density on the generation cost of fusion power plant

    International Nuclear Information System (INIS)

    Huang Desuo; Wu Yican

    2004-01-01

    The impact of the engineering energy gain and the fusion-power-core (FPC) mass power density (MPD) on the generation cost of fusion power plant are analyzed based on the economic elasticity approach in this paper. From the functions describing the relationship of the generation cost with the engineering energy gain and the MPD, the elasticity coefficients of the generation cost with the engineering energy gain and the MPD have been derived respectively to analyze their sensitivity to the generation cost and the MPD to the generation cost decreases with increasing the engineering energy gain or the MPD. (authors)

  14. Fusion power core engineering for the ARIES-ST power plant

    International Nuclear Information System (INIS)

    Tillack, M.S.; Wang, X.R.; Pulsifer, J.; Malang, S.; Sze, D.K.; Billone, M.; Sviatoslavsky, I.

    2003-01-01

    ARIES-ST is a 1000 MWe fusion power plant based on a low aspect ratio 'spherical torus' (ST) plasma. The ARIES-ST power core was designed to accommodate the unique features of an ST power plant, to meet the top-level requirements of an attractive fusion energy source, and to minimize extrapolation from the fusion technology database under development throughout the world. The result is an advanced helium-cooled ferritic steel blanket with flowing PbLi breeder and tungsten plasma-interactive components. Design improvements, such as the use of SiC inserts in the blanket to extend the outlet coolant temperature range were explored and the results are reported here. In the final design point, the power and particle loads found in ARIES-ST are relatively similar to other advanced tokamak power plants (e.g. ARIES-RS [Fusion Eng. Des. 38 (1997) 3; Fusion Eng. Des. 38 (1997) 87]) such that exotic technologies were not required in order to satisfy all of the design criteria. Najmabadi and the ARIES Team [Fusion Eng. Des. (this issue)] provide an overview of ARIES-ST design. In this article, the details of the power core design are presented together with analysis of the thermal-hydraulic, thermomechanical and materials behavior of in-vessel components. Detailed engineering analysis of ARIES-ST TF and PF systems, nuclear analysis, and safety are given in the companion papers

  15. Environment-based pin-power reconstruction method for homogeneous core calculations

    International Nuclear Information System (INIS)

    Leroyer, H.; Brosselard, C.; Girardi, E.

    2012-01-01

    Core calculation schemes are usually based on a classical two-step approach associated with assembly and core calculations. During the first step, infinite lattice assemblies calculations relying on a fundamental mode approach are used to generate cross-sections libraries for PWRs core calculations. This fundamental mode hypothesis may be questioned when dealing with loading patterns involving several types of assemblies (UOX, MOX), burnable poisons, control rods and burn-up gradients. This paper proposes a calculation method able to take into account the heterogeneous environment of the assemblies when using homogeneous core calculations and an appropriate pin-power reconstruction. This methodology is applied to MOX assemblies, computed within an environment of UOX assemblies. The new environment-based pin-power reconstruction is then used on various clusters of 3x3 assemblies showing burn-up gradients and UOX/MOX interfaces, and compared to reference calculations performed with APOLLO-2. The results show that UOX/MOX interfaces are much better calculated with the environment-based calculation scheme when compared to the usual pin-power reconstruction method. The power peak is always better located and calculated with the environment-based pin-power reconstruction method on every cluster configuration studied. This study shows that taking into account the environment in transport calculations can significantly improve the pin-power reconstruction so far as it is consistent with the core loading pattern. (authors)

  16. A benefit assessment of using in-core neutron detector signals in core protection calculator system (CPCS)

    International Nuclear Information System (INIS)

    Han, S.; Park, S.J.; Seong, P.H.

    1997-01-01

    A Core Protection Calculator System (CPCS) is a digital computer based safety system generating trip signals based on the calculation of Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). Currently, CPCS uses ex-core detector signals for core power calculation and it has some uncertainties. In this study, in-core detector signals which directly measure inside flux of core are applied to CPCS to get more accurate power distribution profile, DNBR and LPD. In order to improve axial power distribution calculation, piece-wise cubic Spline method is applied; from the 40 nodes of expanded signals, more accurate and detailed core information can be provided. Simulation is carried out to verify its applicability to power distribution calculation. Simulation result shows that the improved method reduces the calculational uncertainties significantly and it allows larger operational margin. It is also expected that no power reduction is required while Core Operating Limit Supervisory System (COLSS) is out-of-service due to reduced uncertainties when the improved method is applied. In this study, a quantitative economic benefit assessment of using in-core neutron detector signals is also carried out. (authors)

  17. A benefit assessment of using in-core neutron detector signals in core protection calculator system(CPCS)

    International Nuclear Information System (INIS)

    Han, Seung

    1996-02-01

    A Core Protection Calculator System(CPCS) is a digital computer based safety system generating trip signals based on the calculation of Departure from Nucleate Boiling Ratio(DNBR) and Local Power Density(LPD). Currently, CPCS uses ex-core detector signals for core power calculation and it has some uncertainties. In this study, In-core detector signals which directly measure inside flux of core are applied to CPCS to get more accurate power distribution profile, DNBR and LPD. In order to improve axial power distribution calculation, piecewise cubic spline method is applied: From the 40 nodes of expanded signals, more accurate and detailed core information can be provided. Simulation is carried out to verify its applicability to power distribution calculation. Simulation result shows that the improved method reduces the calculational uncertainties significantly and it allows larger operational margin. It is also expected that no power reduction is required while Core Operating Limit Supervisory System(COLSS) is out-of-service due to reduced uncertainties when the improved method is applied. In this study, a quantitative economic benefit assessment of using in-core neutron detector signals is also carried out

  18. High power density yeast catalyzed microbial fuel cells

    Science.gov (United States)

    Ganguli, Rahul

    Microbial fuel cells leverage whole cell biocatalysis to convert the energy stored in energy-rich renewable biomolecules such as sugar, directly to electrical energy at high efficiencies. Advantages of the process include ambient temperature operation, operation in natural streams such as wastewater without the need to clean electrodes, minimal balance-of-plant requirements compared to conventional fuel cells, and environmentally friendly operation. These make the technology very attractive as portable power sources and waste-to-energy converters. The principal problem facing the technology is the low power densities compared to other conventional portable power sources such as batteries and traditional fuel cells. In this work we examined the yeast catalyzed microbial fuel cell and developed methods to increase the power density from such fuel cells. A combination of cyclic voltammetry and optical absorption measurements were used to establish significant adsorption of electron mediators by the microbes. Mediator adsorption was demonstrated to be an important limitation in achieving high power densities in yeast-catalyzed microbial fuel cells. Specifically, the power densities are low for the length of time mediator adsorption continues to occur. Once the mediator adsorption stops, the power densities increase. Rotating disk chronoamperometry was used to extract reaction rate information, and a simple kinetic expression was developed for the current observed in the anodic half-cell. Since the rate expression showed that the current was directly related to microbe concentration close to the electrode, methods to increase cell mass attached to the anode was investigated. Electrically biased electrodes were demonstrated to develop biofilm-like layers of the Baker's yeast with a high concentration of cells directly connected to the electrode. The increased cell mass did increase the power density 2 times compared to a non biofilm fuel cell, but the power density

  19. Some aspects of optimising the reactor core for a 600 MW(e) high temperature helium turbine power plant

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U; Presser, W

    1972-04-24

    For the HHT 600 MW(e) power plant a core design with Triagonal blocks containing 24 channels with directly cooled fuel pins was considered. The design was found to require a low HM loading in the fuel zone to achieve favourable economic merits. For low HM densities a strong incentive exists to aim for burn-ups between 80 and 100 GWd/t. At the present an average discharge irradiation of 80 GWd/t was thought feasible and a reference design with a HM density of 0.6 g/cm {sup 3} in both core zones was chosen. The optimisation is not likely to be upset by local hot channel effects as a special investigation into the influence of safety margins found no changes in fuel cycle economics.

  20. Fuzzy logic based power-efficient real-time multi-core system

    CERN Document Server

    Ahmed, Jameel; Najam, Shaheryar; Najam, Zohaib

    2017-01-01

    This book focuses on identifying the performance challenges involved in computer architectures, optimal configuration settings and analysing their impact on the performance of multi-core architectures. Proposing a power and throughput-aware fuzzy-logic-based reconfiguration for Multi-Processor Systems on Chip (MPSoCs) in both simulation and real-time environments, it is divided into two major parts. The first part deals with the simulation-based power and throughput-aware fuzzy logic reconfiguration for multi-core architectures, presenting the results of a detailed analysis on the factors impacting the power consumption and performance of MPSoCs. In turn, the second part highlights the real-time implementation of fuzzy-logic-based power-efficient reconfigurable multi-core architectures for Intel and Leone3 processors. .

  1. The online simulation of core physics in nuclear power plant

    International Nuclear Information System (INIS)

    Zhao Qiang

    2005-01-01

    The three-dimensional power distribution in core is one of the most important status variables of nuclear reactor. In order to monitor the 3-D in core power distribution timely and accurately, the online simulation system of core physics was designed in the paper. This system combines core physics simulation with the data, which is from the plant and reactor instrumentation. The design of the system consists of the hardware part and the software part. The online simulation system consists of a main simulation computer and a simulation operation station. The online simulation system software includes of the real-time simulation support software, the system communication software, the simulation program and the simulation interface software. Two-group and three-dimensional neutron kinetics model with six groups delayed neutrons was used in the real-time simulation of nuclear reactor core physics. According to the characteristics of the nuclear reactor, the core was divided into many nodes. Resolving the neutron equation, the method of separate variables was used. The input data from the plant and reactor instrumentation system consist of core thermal power, loop temperatures and pressure, control rod positions, boron concentration, core exit thermocouple data, Excore detector signals, in core flux detectors signals. There are two purposes using the data, one is to ensure that the model is as close as the current actual reactor condition, and the other is to calibrate the calculated power distribution. In this paper, the scheme of the online simulation system was introduced. Under the real-time simulation support system, the simulation program is being compiled. Compared with the actual operational data, the elementary simulation results were reasonable and correct. (author)

  2. Core Power Control of the fast nuclear reactors with estimation of the delayed neutron precursor density using Sliding Mode method

    International Nuclear Information System (INIS)

    Ansarifar, G.R.; Nasrabadi, M.N.; Hassanvand, R.

    2016-01-01

    Highlights: • We present a S.M.C. system based on the S.M.O for control of a fast reactor power. • A S.M.O has been developed to estimate the density of delayed neutron precursor. • The stability analysis has been given by means Lyapunov approach. • The control system is guaranteed to be stable within a large range. • The comparison between S.M.C. and the conventional PID controller has been done. - Abstract: In this paper, a nonlinear controller using sliding mode method which is a robust nonlinear controller is designed to control a fast nuclear reactor. The reactor core is simulated based on the point kinetics equations and one delayed neutron group. Considering the limitations of the delayed neutron precursor density measurement, a sliding mode observer is designed to estimate it and finally a sliding mode control based on the sliding mode observer is presented. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. Sliding Mode Control (SMC) is one of the robust and nonlinear methods which have several advantages such as robustness against matched external disturbances and parameter uncertainties. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability.

  3. Fast core prediction simulator for load follow control

    International Nuclear Information System (INIS)

    Yim, Man Sung; Lee, Sang Hoon; Lee, Un Chul

    1990-01-01

    An operator-assisting system for the reactor core control under power changing operating condition was developed. The system is consisted of core simulator routine and Xenon and Iodine initial condition generation routine. The initial condition generation routine, without exactly knowing the core status, is capable of providing accurate number densities and axial offset conditions of Xenon and Iodine after several hours of predictor- corrector calculations using the plant instrumentation signals of power level and power axial offset. The core simulator routine, even with the two node core model, gives equivalently accurate results as the one-dimensional model for the core behaviour simulation under power changing condition and can provide proper control strategies for load follow operation. The core simulator can also be used by the operator to develop remedial actions to restore the distorted power distribution by using its prediction capability

  4. UNVEILING THE DETAILED DENSITY AND VELOCITY STRUCTURES OF THE PROTOSTELLAR CORE B335

    Energy Technology Data Exchange (ETDEWEB)

    Kurono, Yasutaka; Saito, Masao; Kamazaki, Takeshi; Morita, Koh-Ichiro; Kawabe, Ryohei, E-mail: yasutaka.kurono@nao.ac.jp [Chile Observatory, National Astronomical Observatory of Japan, Osawa 2-21-1, Mitaka, Tokyo 181-8588 (Japan)

    2013-03-10

    We present an observational study of the protostellar core B335 harboring a low-mass Class 0 source. The observations of the H{sup 13}CO{sup +}(J = 1-0) line emission were carried out using the Nobeyama 45 m telescope and Nobeyama Millimeter Array. Our combined image of the interferometer and single-dish data depicts detailed structures of the dense envelope within the core. We found that the core has a radial density profile of n(r){proportional_to}r {sup -p} and a reliable difference in the power-law indices between the outer and inner regions of the core: p Almost-Equal-To 2 for r {approx}> 4000 AU and p Almost-Equal-To 1.5 for r {approx}< 4000 AU. The dense core shows a slight overall velocity gradient of {approx}1.0 km s{sup -1} over the scale of 20, 000 AU across the outflow axis. We believe that this velocity gradient represents a solid-body-like rotation of the core. The dense envelope has a quite symmetrical velocity structure with a remarkable line broadening toward the core center, which is especially prominent in the position-velocity diagram across the outflow axis. The model calculations of position-velocity diagrams do a good job of reproducing observational results using the collapse model of an isothermal sphere in which the core has an inner free-fall region and an outer region conserving the conditions at the formation stage of a central stellar object. We derived a central stellar mass of {approx}0.1 M{sub Sun }, and suggest a small inward velocity, v{sub r{>=}r{sub i{sub n{sub f}}}}{approx}0 km s{sup -1} in the outer core at {approx}> 4000 AU. We concluded that our data can be well explained by gravitational collapse with a quasi-static initial condition, such as Shu's model, or by the isothermal collapse of a marginally critical Bonnor-Ebert sphere.

  5. JOYO MK-II core characteristics database

    International Nuclear Information System (INIS)

    Ohkawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi; Aoyama, Takafumi

    2003-04-01

    The 'JOYO' MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. Total control rod worth, reactor kinetic parameters and the MK-II core performance test results were included per user's requests. The core characteristics obtained from the 32 nd to 35 th operational cycles, which were conducted in the MK-III transition core, were newly added in this revised version. The MK-II core management data and core characteristics data were recorded to CD-ROM for user convenience. The Configuration Data' include the core arrangement and refueling record for each operational cycle. The 'Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The 'Output Data' contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The 'Core Characteristics Data' include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. (author)

  6. Development of JOYO MK-II core characteristics database

    International Nuclear Information System (INIS)

    Tabuchi, Shiro; Aoyama, Takafumi

    2000-01-01

    The MK-II core of the experimental fast reactor JOYO served as the irradiation bed for testing fuels and materials for FBR development since 1982 for 15 years. During the MK-II operation, extensive data were accumulated from the core management calculations and characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database recorded on CD-ROM for user convenience. The calculated core management data are the text style data. The 'Configuration Data' include the history of the fuel exchange and core arrangement for each cycle. The Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of about 300 fuel subassemblies, and 60 irradiation subassemblies. The 'Output Data' include the neutron fluxes, gamma fluxes, power density, linear heat rates, coolant and fuel temperature distributions of each core position at the beginning and end of each cycle. The measured core characteristics data, such as the excess reactivity, control rod worths, temperature coefficient, power coefficient, and burn-up coefficient are also included along with the measurement conditions. (J.P.N.)

  7. Error Analysis of High Frequency Core Loss Measurement for Low-Permeability Low-Loss Magnetic Cores

    DEFF Research Database (Denmark)

    Niroumand, Farideh Javidi; Nymand, Morten

    2016-01-01

    in magnetic cores is B-H loop measurement where two windings are placed on the core under test. However, this method is highly vulnerable to phase shift error, especially for low-permeability, low-loss cores. Due to soft saturation and very low core loss, low-permeability low-loss magnetic cores are favorable...... in many of the high-efficiency high power-density power converters. Magnetic powder cores, among the low-permeability low-loss cores, are very attractive since they possess lower magnetic losses in compared to gapped ferrites. This paper presents an analytical study of the phase shift error in the core...... loss measuring of low-permeability, low-loss magnetic cores. Furthermore, the susceptibility of this measurement approach has been analytically investigated under different excitations. It has been shown that this method, under square-wave excitation, is more accurate compared to sinusoidal excitation...

  8. Long time-scale density peaking in JET

    International Nuclear Information System (INIS)

    Sartori, R.; Saibene, G.; Becoulet, M.

    2002-01-01

    This paper discusses how the proximity to the L-H threshold affects the confinement of ELMy H-modes at high density. The largest reduction in confinement at high density is observed at the transition from the Type I to the Type III ELMy regime. At medium plasma triangularity, δ≅0.3 (where δ is the average triangularity at the separatrix), JET experiments show that, by increasing the margin above the L-H threshold power and maintaining the edge temperature above the critical temperature for the transition to Type III ELMs, it is possible to avoid the degradation of the pedestal pressure with density, normally observed at lower power. As a result, the range of achievable densities (both in the core and in the pedestal) is increased. At high power above the L-H threshold power the core density was equal to the Greenwald limit with H97≅0.9. There is evidence that a mixed regime of Type I and Type II ELMs has been obtained at this intermediate triangularity, possibly as a result of this increase in density. At higher triangularity, δ≅0.5, the power required to achieve similar results is lower. (author)

  9. High-Power-Density, High-Energy-Density Fluorinated Graphene for Primary Lithium Batteries

    Directory of Open Access Journals (Sweden)

    Guiming Zhong

    2018-03-01

    Full Text Available Li/CFx is one of the highest-energy-density primary batteries; however, poor rate capability hinders its practical applications in high-power devices. Here we report a preparation of fluorinated graphene (GFx with superior performance through a direct gas fluorination method. We find that the so-called “semi-ionic” C-F bond content in all C-F bonds presents a more critical impact on rate performance of the GFx in comparison with sp2 C content in the GFx, morphology, structure, and specific surface area of the materials. The rate capability remains excellent before the semi-ionic C-F bond proportion in the GFx decreases. Thus, by optimizing semi-ionic C-F content in our GFx, we obtain the optimal x of 0.8, with which the GF0.8 exhibits a very high energy density of 1,073 Wh kg−1 and an excellent power density of 21,460 W kg−1 at a high current density of 10 A g−1. More importantly, our approach opens a new avenue to obtain fluorinated carbon with high energy densities without compromising high power densities.

  10. Coupled neutronics and thermal hydraulics of high density cores for FRM II

    Energy Technology Data Exchange (ETDEWEB)

    Breitkreutz, Harald

    2011-03-04

    % enriched disperse UMo core with different densities at two reactor power levels, and two 40% enriched monolithic UMo cores, one with flat plates and one with plates that have a thickness gradient. Again, two different power levels were regarded. To estimate the uncertainty of the performed calculations, a sensitivity analysis was conducted. This includes results of neutronics, burn-up and thermal hydraulics. The identified uncertainties are generally quite small. (orig.)

  11. Coupled neutronics and thermal hydraulics of high density cores for FRM II

    International Nuclear Information System (INIS)

    Breitkreutz, Harald

    2011-01-01

    different densities at two reactor power levels, and two 40% enriched monolithic UMo cores, one with flat plates and one with plates that have a thickness gradient. Again, two different power levels were regarded. To estimate the uncertainty of the performed calculations, a sensitivity analysis was conducted. This includes results of neutronics, burn-up and thermal hydraulics. The identified uncertainties are generally quite small. (orig.)

  12. Cost-based optimizations of power density and target-blanket modularity for 232Th/233U-based ADEP

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1995-01-01

    A cost-based parametric systems model is developed for an Accelerator-Driven Energy Production (ADEP) system based on a 232 Th/ 233 U fuel cycle and a molten-salt (LiF/BeF 2 /ThF 3 ) fluid-fuel primary system. Simplified neutron-balance, accelerator, reactor-core, chemical-processing, and balance-of-plant models are combined parametrically with a simplified costing model. The main focus of this model is to examine trade offs related to fission power density, reactor-core modularity, 233 U breeding rate, and fission product transmutation capacity

  13. Modeling the Power Variability of Core Speed Scaling on Homogeneous Multicore Systems

    Directory of Open Access Journals (Sweden)

    Zhihui Du

    2017-01-01

    Full Text Available We describe a family of power models that can capture the nonuniform power effects of speed scaling among homogeneous cores on multicore processors. These models depart from traditional ones, which assume that individual cores contribute to power consumption as independent entities. In our approach, we remove this independence assumption and employ statistical variables of core speed (average speed and the dispersion of the core speeds to capture the comprehensive heterogeneous impact of subtle interactions among the underlying hardware. We systematically explore the model family, deriving basic and refined models that give progressively better fits, and analyze them in detail. The proposed methodology provides an easy way to build power models to reflect the realistic workings of current multicore processors more accurately. Moreover, unlike the existing lower-level power models that require knowledge of microarchitectural details of the CPU cores and the last level cache to capture core interdependency, ours are easier to use and scalable to emerging and future multicore architectures with more cores. These attributes make the models particularly useful to system users or algorithm designers who need a quick way to estimate power consumption. We evaluate the family of models on contemporary x86 multicore processors using the SPEC2006 benchmarks. Our best model yields an average predicted error as low as 5%.

  14. Thermal-Hydraulic Performance of Cross-Shaped Spiral Fuel in High-Power-Density BWRs

    International Nuclear Information System (INIS)

    Conboy, Thomas; Hejzlar, Pavel

    2006-01-01

    Power up-rating of existing nuclear reactors promises to be an area of great study for years to come. One of the major approaches to efficiently increasing power density is by way of advanced fuel design, and cross-shaped spiral-fuel has shown such potential in previous studies. Our work aims to model the thermal-hydraulic consequences of filling a BWR core with these spiral-shaped pins. The helically-wound pins have a cross-section resembling a 4-petaled flower. They fill an assembly in a tight bundle, their dimensions chosen carefully such that the petals of neighboring pins contact each other at their outer-most extent in a self-supporting lattice, absent of grid spacers. Potential advantages of this design raise much optimism from a thermal-hydraulic perspective. These spiral rods possess about 40% larger surface area than traditional rods, resulting in increased cooling and a proportional reduction in average surface heat flux. The thin petal-like extensions help by lowering thermal resistance between the hot central region of the pin and the bulk coolant flow, decreasing the maximum fuel temperature by 200 deg. C according to Finite Element (COSMOS) models. However, COSMOS models also predict a potential problem area at the 'elbow' region of two adjoining petals, where heat flux peaking is twice that along the extensions. Preliminary VIPRE models, which account only for the surface area increase, predict a 22% increase in critical power. It is also anticipated that the spiral twist would provide the flowing coolant with an additional radial velocity component, and likely promote turbulence and mixing within an assembly. These factors are expected to provide further margin for increased power density, and are currently being incorporated into the VIPRE model. The reduction in pressure drop inherent in any core without grid-spacers is also expected to be significant in aiding core stability, though this has not yet been quantified. Spiral-fuel seems to be a

  15. Power distribution gradients in WWER type cores and fuel failure root causes

    Energy Technology Data Exchange (ETDEWEB)

    Mikuš, Ján M., E-mail: JanMikus.nrc@hotmail.com

    2014-02-15

    Highlights: • Power (fission rate) distribution gradients can represent fuel failure root causes. • Positions with above gradients were investigated in WWER type cores on reactor LR-0. • Above gradients were evaluated near core heterogeneities and construction materials. • Results can be used for code validation and fuel failure occurrence investigation. - Abstract: Neutron flux non-uniformity and gradients of neutron current resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plant (NPP), results of some benchmark type experiments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the spatial power distribution changes in fuel assemblies (FAs): Containing fuel rods (FRs) with Gd burnable absorber in WWER-440 and WWER-1000 type cores, Neighboring the core blanket and dummy steel assembly simulators on the periphery of the WWER-440 standard and low leakage type cores, resp., Neighboring baffle in WWER-1000 type cores, and Neighboring control rod (CR) in WWER-440 type cores, namely (a) power peak in axial power distribution in periphery FRs of the adjacent FAs near the area between CR fuel part and butt joint to the CR absorbing part and (b) decrease in radial power distribution in FRs near CR absorbing part. An overview of relevant experimental results from reactor LR-0 and some information concerning leaking FAs on NPP Temelín are presented. Obtained data can be used for code validation and subsequently for the fuel failure occurrence investigation.

  16. Monitoring device for the power distribution within a nuclear reactor core

    International Nuclear Information System (INIS)

    Tanzawa, Tomio; Kumanomido, Hironori; Toyoshi, Isamu.

    1986-01-01

    Purpose: To provide a monitoring device for the power distribution in the reactor core that calculates the power distribution based on the measurement by instruments disposed within the reactor core of BWR type reactors. Constitution: The power distribution monitoring device in a reactor core comprises a signal correcting device, a signal normalizing device and a power distribution calculating device, in which the power distribution calculating device is constituted with an average power calculating device for four fuel assemblies and an average power calculating device for fuel assemblies. Gamma-ray signals corrected by the signal correcting device and signals from neutron detectors are inputted to the signal normalizing device, both of which are calibrated to determine the axial gamma-ray signal distribution in the central water gap region with the four fuel assemblies being as the unit. The average power from the four fuel assemblies are inputted to the fuel assembly average power calculating device to allocate to each of the fuel assembly average power thereby attaining the purpose. Further, thermal restriction values are calculated thereby enabling to secure the fuel integrity. (Kamimura, M.)

  17. Fuel pin failure root causes and power distribution gradients in WWER cores

    International Nuclear Information System (INIS)

    Mikus, J.

    2008-01-01

    The purpose of this work is to investigate the influence of some core heterogeneities and reactor construction materials on space power distribution in WWER type cores, especially from viewpoint of the values and gradient occurrence that could result in static loads with some consequences, e.g., fuel pin (FP) or fuel assembly (FA) bowing and possible contribution to the FP failure root causes. Presented information were obtained by means of experiments on research reactor LR-0 concerning the: 1) Power distribution estimation on pellet surface of the FPs neighbouring a FP containing gadolinium (Gd 2 O 3 ) burnable absorber integrated into fuel in WWER-440 and -1000 type cores; 2) Power distribution measurement in periphery FAs neighbouring the baffle in WWER-1000 type cores and 3) Power distribution in FAs neighbouring the control rod absorbing part in a WWER-440 type core. (author)

  18. Preliminary concept of a zero power nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

    2011-01-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  19. Preliminary concept of a zero power nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  20. Gas density fluctuations in the Perseus Cluster: clumping factor and velocity power spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Zhuravleva, I.; Churazov, E.; Arevalo, P.; Schekochihin, A. A.; Allen, S. W.; Fabian, A. C.; Forman, W. R.; Sanders, J. S.; Simionescu, A.; Sunyaev, R.; Vikhlinin, A.; Werner, N.

    2015-05-20

    X-ray surface brightness fluctuations in the core of the Perseus Cluster are analysed, using deep observations with the Chandra observatory. The amplitude of gas density fluctuations on different scales is measured in a set of radial annuli. It varies from 7 to 12 per cent on scales of ~10–30 kpc within radii of 30–220 kpc from the cluster centre. Using a statistical linear relation between the observed amplitude of density fluctuations and predicted velocity, the characteristic velocity of gas motions on each scale is calculated. The typical amplitudes of the velocity outside the central 30 kpc region are 90–140 km s-1 on ~20–30 kpc scales and 70–100 km s-1 on smaller scales ~7–10 kpc. The velocity power spectrum (PS) is consistent with cascade of turbulence and its slope is in a broad agreement with the slope for canonical Kolmogorov turbulence. The gas clumping factor estimated from the PS of the density fluctuations is lower than 7–8 per cent for radii ~30–220 kpc from the centre, leading to a density bias of less than 3–4 per cent in the cluster core. Uncertainties of the analysis are examined and discussed. Future measurements of the gas velocities with the Astro-H, Athena and Smart-X observatories will directly measure the gas density–velocity perturbation relation and further reduce systematic uncertainties in this analysis.

  1. On-line reconstruction of in-core power distribution by harmonics expansion method

    International Nuclear Information System (INIS)

    Wang Changhui; Wu Hongchun; Cao Liangzhi; Yang Ping

    2011-01-01

    Highlights: → A harmonics expansion method for the on-line in-core power reconstruction is proposed. → A harmonics data library is pre-generated off-line and a code named COMS is developed. → Numerical results show that the maximum relative error of the reconstruction is less than 5.5%. → This method has a high computational speed compared to traditional methods. - Abstract: Fixed in-core detectors are most suitable in real-time response to in-core power distributions in pressurized water reactors (PWRs). In this paper, a harmonics expansion method is used to reconstruct the in-core power distribution of a PWR on-line. In this method, the in-core power distribution is expanded by the harmonics of one reference case. The expansion coefficients are calculated using signals provided by fixed in-core detectors. To conserve computing time and improve reconstruction precision, a harmonics data library containing the harmonics of different reference cases is constructed. Upon reconstruction of the in-core power distribution on-line, the two closest reference cases are searched from the harmonics data library to produce expanded harmonics by interpolation. The Unit 1 reactor of DayaBay Nuclear Power Plant (DayaBay NPP) in China is considered for verification. The maximum relative error between the measurement and reconstruction results is less than 5.5%, and the computing time is about 0.53 s for a single reconstruction, indicating that this method is suitable for the on-line monitoring of PWRs.

  2. Metamorphic core complex formation by density inversion and lower-crust extrusion.

    Science.gov (United States)

    Martinez, F; Goodliffe, A M; Taylor, B

    2001-06-21

    Metamorphic core complexes are domal uplifts of metamorphic and plutonic rocks bounded by shear zones that separate them from unmetamorphosed cover rocks. Interpretations of how these features form are varied and controversial, and include models involving extension on low-angle normal faults, plutonic intrusions and flexural rotation of initially high-angle normal faults. The D'Entrecasteaux islands of Papua New Guinea are actively forming metamorphic core complexes located within a continental rift that laterally evolves to sea-floor spreading. The continental rifting is recent (since approximately 6 Myr ago), seismogenic and occurring at a rapid rate ( approximately 25 mm yr-1). Here we present evidence-based on isostatic modelling, geological data and heat-flow measurements-that the D'Entrecasteaux core complexes accommodate extension through the vertical extrusion of ductile lower-crust material, driven by a crustal density inversion. Although buoyant extrusion is accentuated in this region by the geological structure present-which consists of dense ophiolite overlaying less-dense continental crust-this mechanism may be generally applicable to regions where thermal expansion lowers crustal density with depth.

  3. Identifying functions for ex-core neutron noise analysis

    International Nuclear Information System (INIS)

    Avila, J.M.; Oliveira, J.C.

    1987-01-01

    A method of performing the phase analysis of signals arising from neutron detectors placed in the periphery of a pressurized water reactor is proposed. It consists in the definition of several identifying functions, based on the phases of cross power spectral densities corresponding to four ex-core neutron detectors. Each of these functions enhances the appearance of different sources of noise. The method, applied to the ex-core neutron fluctuation analysis of a French PWR, proved to be very useful as it allows quick recognition of various patterns in the power spectral densities. (orig.) [de

  4. Modeling and analysis of the disk MHD generator component of a gas core reactor/MHD Rankine cycle space power system

    International Nuclear Information System (INIS)

    Welch, G.E.; Dugan, E.T.; Lear, W.E. Jr.; Appelbaum, J.G.

    1990-01-01

    A gas core nuclear reactor (GCR)/disk magnetohydrodynamic (MHD) generator direct closed Rankine space power system concept is described. The GCR/disk MHD generator marriage facilitates efficient high electric power density system performance at relatively high operating temperatures. The system concept promises high specific power levels, on the order of 1 kW e /kg. An overview of the disk MHD generator component magnetofluiddynamic and plasma physics theoretical modeling is provided. Results from a parametric design analysis of the disk MHD generator are presented and discussed

  5. The whiteStar development project: Westinghouse's next generation core design simulator and core monitoring software to power the nuclear renaissance

    International Nuclear Information System (INIS)

    Boyd, W. A.; Mayhue, L. T.; Penkrot, V. S.; Zhang, B.

    2009-01-01

    The WhiteStar project has undertaken the development of the next generation core analysis and monitoring system for Westinghouse Electric Company. This on-going project focuses on the development of the ANC core simulator, BEACON core monitoring system and NEXUS nuclear data generation system. This system contains many functional upgrades to the ANC core simulator and BEACON core monitoring products as well as the release of the NEXUS family of codes. The NEXUS family of codes is an automated once-through cross section generation system designed for use in both PWR and BWR applications. ANC is a multi-dimensional nodal code for all nuclear core design calculations at a given condition. ANC predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. BEACON is an advanced core monitoring and support system which uses existing instrumentation data in conjunction with an analytical methodology for on-line generation and evaluation of 3D core power distributions. This new system is needed to design and monitor the Westinghouse AP1000 PWR. This paper describes provides an overview of the software system, software development methodologies used as well some initial results. (authors)

  6. Reconstruction of core axial power shapes using the alternating conditional expectation algorithm

    International Nuclear Information System (INIS)

    Lee, Eun Ki; Kim, Yong Hee; Cha, Kune Ho; Park, Moon Ghu

    1999-01-01

    We have introduced the alternating conditional expectation (ACE) algorithm in reconstructing 20-node axial core power shapes from five-level in-core detector powers. The core design code, Reactor Operation and Control Simulation (ROCS), calculates 3-dimensional power distributions for various core states, and the reference core-averaged axial power shapes and corresponding simulated detector powers are utilized to synthesize the axial power shape. By using the ACE algorithm, the optimal relationship between a dependent variable, the plane power, and independent variables, five detector powers, is determined without any preprocessing. A total of ∼3490 data sets per each cycle of YongGwang Nuclear (YGN) power plant units 3 and 4 is used for the regression. Continuous analytic function corresponding to each optimal transformation is calculated by simple regression model. The reconstructed axial power shapes of ∼21,200 cases are compared to the original ROCS axial power shapes. Also, to test the validity and accuracy of the new method, its performance is compared with that of the Fourier fitting method (FFM), a typical method of the deterministic approach. For a total of 21,204 data cases, the averages of root mean square (rms) error, axial peak error (ΔF z ), and axial shape index error (ΔASI) of new method are calculated as 0.81%, 0.51% and 0.00204, while those of FFM are 2.29%, 2.37% and 0.00264, respectively. We also evaluated the wide range of axial power profiles from the xenon-oscillation. The results show that the newly developed method is far superior to FFM; average rms and axial peak error are just ∼35 and ∼20% of those of FFM, respectively

  7. ICRF power limitation relation to density limit in ASDEX

    International Nuclear Information System (INIS)

    Ryter, F.

    1992-01-01

    Launching high ICRF power into ASDEX plasmas required good antenna-plasma coupling. This could be achieved by sufficient electron density in front of the antennas i.e. small antenna-plasma distance (1-2 cm) and moderate to high line-averaged electron density compared to the density window in ASDEX. These are conditions eventually close to the density limit. ICRF heated discharges terminated by plasma disruptions caused by the RF pulse limited the maximum RF power which can be injected into the plasma. The disruptions occurring in these cases have clear phenomenological similarities with those observed in density limit discharges. We show in this paper that the ICRF-power limitation by plasma disruptions in ASDEX was due to reaching the density limit. (orig.)

  8. ICRF power limitation relation to density limit in ASDEX

    International Nuclear Information System (INIS)

    Ryter, F.

    1992-01-01

    Launching high ICRF power into ASDEX plasmas required good antenna-plasma coupling. This could be achieved by sufficient electron density in front of the antennas i.e. small antenna-plasma distance (1-2 cm) and moderate to high line-averaged electron density compared to the density window in ASDEX. These are conditions eventually close to the density limit. ICRF heated discharges terminated by plasma disruptions caused by the RF pulse limited the maximum RF power which can be injected into the plasma. The disruptions occurring in these cases have clear phenomenological similarities with those observed in density limit discharges. We show in this paper that the ICRF-power limitation by plasma disruptions in ASDEX was due to reaching the density limit. (author) 3 refs., 3 figs

  9. Density and temperature profile modifications with electron cyclotron power injection in quiescent double barrier discharges on DIII-D

    International Nuclear Information System (INIS)

    Casper, T A; Burrell, K H; Doyle, E J; Gohil, P; Lasnier, C J; Leonard, A W; Moller, J M; Osborne, T H; Snyder, P B; Thomas, D M; Weiland, J; West, W P

    2006-01-01

    Quiescent double barrier (QDB) conditions often form when an internal transport barrier is created with high-power neutral-beam injection into a quiescent H-mode (QH) plasma. These QH-modes offer an attractive, high-performance operating scenario for burning plasma experiments due to their quasi-stationarity and lack of edge localized modes. Our initial experiments and modelling using ECH/ECCD in QDB shots were designed to control the current profile and, indeed, we have observed a strong dependence on the q-profile when EC-power is used inside the core transport barrier region. While strong electron heating is observed with EC power injection, we also observe a drop in the other core parameters, namely ion temperature and rotation, electron density and impurity concentration. At onset and termination of the EC pulse, dynamically changing conditions are induced that provide a rapid evolution of T e /T i profiles accessible with 0.3 e /T i ) axis e /T i ratio as the ion temperature and density profiles flatten with this change in transport. The change in transport is consistent with a destabilization of ITG turbulence as inferred from the reduction of the stability threshold due to the change in T e /T i

  10. Research on Power Factor Correction Boost Inductor Design Optimization – Efficiency vs. Power Density

    DEFF Research Database (Denmark)

    Li, Qingnan; Andersen, Michael A. E.; Thomsen, Ole Cornelius

    2011-01-01

    Nowadays, efficiency and power density are the most important issues for Power Factor Correction (PFC) converters development. However, it is a challenge to reach both high efficiency and power density in a system at the same time. In this paper, taking a Bridgeless PFC (BPFC) as an example......, a useful compromise between efficiency and power density of the Boost inductors on 3.2kW is achieved using an optimized design procedure. The experimental verifications based on the optimized inductor are carried out from 300W to 3.2kW at 220Vac input....

  11. Core power distribution measurement and data processing in Daya Bay Nuclear Power Station

    International Nuclear Information System (INIS)

    Zhang Hong

    1997-01-01

    For the first time in China, Daya Bay Nuclear Power Station applied the advanced technology of worldwide commercial pressurized reactors to the in-core detectors, the leading excore six-chamber instrumentation for precise axial power distribution, and the related data processing. Described in this article are the neutron flux measurement in Daya Bay Nuclear Power Station, and the detailed data processing

  12. Calculation of power density with MCNP in TRIGA reactor

    International Nuclear Information System (INIS)

    Snoj, L.; Ravnik, M.

    2006-01-01

    Modern Monte Carlo codes (e.g. MCNP) allow calculation of power density distribution in 3-D geometry assuming detailed geometry without unit-cell homogenization. To normalize MCNP calculation by the steady-state thermal power of a reactor, one must use appropriate scaling factors. The description of the scaling factors is not adequately described in the MCNP manual and requires detailed knowledge of the code model. As the application of MCNP for power density calculation in TRIGA reactors has not been reported in open literature, the procedure of calculating power density with MCNP and its normalization to the power level of a reactor is described in the paper. (author)

  13. Heterogeneous gas core reactor

    International Nuclear Information System (INIS)

    Han, K.I.

    1977-01-01

    Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant

  14. In-core neutron flux measurements at PARR using self powered neutron detector

    International Nuclear Information System (INIS)

    Hussain, A.; Ansari, S.A.

    1989-10-01

    This report describes experimental reactor physics measure ments at PARR using the in-core neutron detectors. Rhodium self powered neutron detectors (SPND) were used in the PARR core and several measurements were made aimed at detector calibration, response time determination and neutron flux measurements. The detectors were calibrated at low power using gold foils and full power by the thermal channel. Based on this calibration it was observed that the detector response remains almost linear throughout the power range. The self powered detectors were used for on-line determination of absolute neutron flux in the core as well as the spatial distribution of neutron flux or reactor power. The experimental, axial and horizontal flux mapping results at certain locations in the core are presented. The total response time of rhodium detector was experimentally determined to be about 5 minutes, which agree well with the theoretical results. Because of longer response time of SPND of the detectors it is not possible to use them in the reactor protection system. (author). 10 figs

  15. Determination of neutron flux densities in WWR-S reactor core

    International Nuclear Information System (INIS)

    Tomasek, F.

    1989-04-01

    The method is described of determining neutron flux densities and neutron fluences using activation detectors. The basic definitions and relations for determining reaction rates, fluence and neutron flux as well as the characteristics of some reactions and of sitable activation detectors are reported. The flux densities were determined of thermal and fast neutrons and of gamma quanta in the WWR-S reactor core. The data measured in the period 1984-1987 are tabulated. Cross sections for the individual reactions were determined from spectra measurements processed using program SAND-II and cross section library ENDF-B IV. Neutron flux densities were also measured for the WWR-S reactor vertical channels. (E.J.). 10 figs., 8 tabs., 111 refs

  16. Modeling of the core of Atucha II nuclear power plant

    International Nuclear Information System (INIS)

    Blanco, Anibal

    2007-01-01

    This work is part of a Nuclear Engineer degree thesis of the Instituto Balseiro and it is carried out under the development of an Argentinean Nuclear Power Plant Simulator. To obtain the best representation of the reactor physical behavior using the state of the art tools this Simulator should couple a 3D neutronics core calculation code with a thermal-hydraulics system code. Focused in the neutronic nature of this job, using PARCS, we modeled and performed calculations of the nuclear power plant Atucha 2 core. Whenever it is possible, we compare our results against results obtained with PUMA (the official core code for Atucha 2). (author) [es

  17. Improving Free-Piston Stirling Engine Power Density

    Science.gov (United States)

    Briggs, Maxwell H.

    2016-01-01

    Analyses and experiments demonstrate the potential benefits of optimizing piston and displacer motion in a free piston Stirling Engine. Isothermal analysis shows the theoretical limits of power density improvement due to ideal motion in ideal Stirling engines. More realistic models based on nodal analysis show that ideal piston and displacer waveforms are not optimal, often producing less power than engines that use sinusoidal piston and displacer motion. Constrained optimization using nodal analysis predicts that Stirling engine power density can be increased by as much as 58% using optimized higher harmonic piston and displacer motion. An experiment is conducted in which an engine designed for sinusoidal motion is forced to operate with both second and third harmonics, resulting in a maximum piston power increase of 14%. Analytical predictions are compared to experimental data showing close agreement with indirect thermodynamic power calculations, but poor agreement with direct electrical power measurements.

  18. Density-based and transport-based core-periphery structures in networks.

    Science.gov (United States)

    Lee, Sang Hoon; Cucuringu, Mihai; Porter, Mason A

    2014-03-01

    Networks often possess mesoscale structures, and studying them can yield insights into both structure and function. It is most common to study community structure, but numerous other types of mesoscale structures also exist. In this paper, we examine core-periphery structures based on both density and transport. In such structures, core network components are well-connected both among themselves and to peripheral components, which are not well-connected to anything. We examine core-periphery structures in a wide range of examples of transportation, social, and financial networks-including road networks in large urban areas, a rabbit warren, a dolphin social network, a European interbank network, and a migration network between counties in the United States. We illustrate that a recently developed transport-based notion of node coreness is very useful for characterizing transportation networks. We also generalize this notion to examine core versus peripheral edges, and we show that the resulting diagnostic is also useful for transportation networks. To examine the properties of transportation networks further, we develop a family of generative models of roadlike networks. We illustrate the effect of the dimensionality of the embedding space on transportation networks, and we demonstrate that the correlations between different measures of coreness can be very different for different types of networks.

  19. In core system mapping reactor power distribution

    International Nuclear Information System (INIS)

    Yoriyaz, H.; Moreira, J.M.L.

    1989-01-01

    Based on the signals of SPND'S (Self Powered Neutron Detectors) distributed inside of a core, the spatial power distribution is obtained using the MAP program, developed in this work. The methodology applied in MAP program uses a least mean square technique to calculate expansion coefficients that depend on the SPND'S signals. The final power or neutron flux distribution is obtained by a combination of certains functions or expansion modes that are provided from diffusion calculation with the CITATION code. The MAP program is written in PASCAL language and will be used in IEA-R1 reactor for assisting its operation. (author) [pt

  20. Evaluation report on CCTF Core-II reflood test second shakedown test, C2-SH2 (Run 54)

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Sugimoto, Jun; Akimoto, Hajime; Okubo, Tsutomu; Murao, Yoshio

    1985-03-01

    A low power test (the initial averaged linear power density = 1.18 kW/m) and the base case test (1.4 kW/m) were performed with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute, in order to study the effect of the power on the reflood phenomena. The former linear power density corresponds nearly to the scaled linear power density based on the current safety evaluation criterio. During the early period of the reflood ( 200s) the heat transfer coefficient became higher and resultantly the quench front advanced faster in the low power test. The core flooding rate was nearly identical between both tests, independently of the different power. The insensitiveness of the power to the core flooding rate was also observed in FLECHT-SET performed in the USA. A significatn large differential pressure oscillation at ECC ports was experienced in the low power test, and it may be important for the long term core cooling although it has not been taken note on the previous studies. (author)

  1. Industrialization of nanocrystalline Fe–Si–B–P–Cu alloys for high magnetic flux density cores

    International Nuclear Information System (INIS)

    Takenaka, Kana; Setyawan, Albertus D.; Sharma, Parmanand; Nishiyama, Nobuyuki; Makino, Akihiro

    2016-01-01

    Nanocrystalline Fe–Si–B–P–Cu alloys exhibit high saturation magnetic flux density (B s ) and extremely low magnetic core loss (W), simultaneously. Low amorphous-forming ability of these alloys hinders their application potential in power transformers and motors. Here we report a solution to this problem. Minor addition of C is found to be effective in increasing the amorphous-forming ability of Fe–Si–B–P–Cu alloys. It allows fabrication of 120 mm wide ribbons (which was limited to less than 40 mm) without noticeable degradation in magnetic properties. The nanocrystalline (Fe 85.7 Si 0.5 B 9.5 P 3.5 Cu 0.8 ) 99 C 1 ribbons exhibit low coercivity (H c )~4.5 A/m, high B s ~1.83 T and low W~0.27 W/kg (@ 1.5 T and 50 Hz). Success in fabrication of long (60–100 m) and wide (~120 mm) ribbons, which are made up of low cost elements is promising for mass production of energy efficient high power transformers and motors - Highlights: • Minor addition of C in FeSiBPCu alloy increases amorphous-forming ability. • The FeSiBPCuC alloy exhibits B s close to Si-steel and Core loss lower than it. • Excellent soft magnetic properties were obtained for 120 mm wide ribbons. • Nanocrystalline FeSiBPCuC alloy can be produced at industrial scale with low cost. • The alloy is suitable for making low energy loss power transformers and motors.

  2. Improved core protection calculator system algorithm

    International Nuclear Information System (INIS)

    Yoon, Tae Young; Park, Young Ho; In, Wang Kee; Bae, Jong Sik; Baeg, Seung Yeob

    2009-01-01

    Core Protection Calculator System (CPCS) is a digitized core protection system which provides core protection functions based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels which adapted a two out of four trip logic. CPCS algorithm improvement for the newly designed core protection calculator system, RCOPS (Reactor COre Protection System), is described in this paper. New features include the improvement of DNBR algorithm for thermal margin, the addition of pre trip alarm generation for auxiliary trip function, VOPT (Variable Over Power Trip) prevention during RPCS (Reactor Power Cutback System) actuation and the improvement of CEA (Control Element Assembly) signal checking algorithm. To verify the improved CPCS algorithm, CPCS algorithm verification tests, 'Module Test' and 'Unit Test', would be performed on RCOPS single channel facility. It is expected that the improved CPCS algorithm will increase DNBR margin and enhance the plant availability by reducing unnecessary reactor trips

  3. Bridge density functional approximation for non-uniform hard core repulsive Yukawa fluid

    International Nuclear Information System (INIS)

    Zhou Shiqi

    2008-01-01

    In this work, a bridge density functional approximation (BDFA) (J. Chem. Phys. 112, 8079 (2000)) for a non-uniform hard-sphere fluid is extended to a non-uniform hard-core repulsive Yukawa (HCRY) fluid. It is found that the choice of a bulk bridge functional approximation is crucial for both a uniform HCRY fluid and a non-uniform HCRY fluid. A new bridge functional approximation is proposed, which can accurately predict the radial distribution function of the bulk HCRY fluid. With the new bridge functional approximation and its associated bulk second order direct correlation function as input, the BDFA can be used to well calculate the density profile of the HCRY fluid subjected to the influence of varying external fields, and the theoretical predictions are in good agreement with the corresponding simulation data. The calculated results indicate that the present BDFA captures quantitatively the phenomena such as the coexistence of solid-like high density phase and low density gas phase, and the adsorption properties of the HCRY fluid, which qualitatively differ from those of the fluids combining both hard-core repulsion and an attractive tail. (condensed matter: structure, thermal and mechanical properties)

  4. Self-powered neutron detector

    International Nuclear Information System (INIS)

    Goldstein, N.P.; Todt, W.H.

    1976-01-01

    A self-powered neutron detector is detailed wherein a thin conductive layer of low neutron cross section, high density material is disposed about an emitter core of material which spontaneously emits radiation on neutron capture. The high density material is absorptive of beta radiation emitted by decay of the emitter core activation product, but is substantially transmissive to the high average energy prompt electrons emitted by the emitter core material. (author)

  5. Comparative assessment of out-of-core nuclear thermionic power systems

    International Nuclear Information System (INIS)

    Estabrook, W.C.; Koenig, D.R.; Prickett, W.Z.

    1975-01-01

    The hardware selections available for fabrication of a nuclear electric propulsion stage for planetary exploration were explored. The investigation was centered around a heat-pipe-cooled, fast-spectrum nuclear reactor for an out-of-core power conversion system with sufficient detail for comparison with the in-core system studies completed previously. A survey of competing power conversion systems still indicated that the modular reliability of thermionic converters makes them the desirable choice to provide the 240-kWe end-of-life power for at least 20,000 full power hours. The electrical energy will be used to operate a number of mercury ion bombardment thrusters with a specific impulse in the range of about 4,000-5,000 seconds. (Author)

  6. Cu₂-xSe@mSiO₂-PEG core-shell nanoparticles: a low-toxic and efficient difunctional nanoplatform for chemo-photothermal therapy under near infrared light radiation with a safe power density.

    Science.gov (United States)

    Liu, Xijian; Wang, Qian; Li, Chun; Zou, Rujia; Li, Bo; Song, Guosheng; Xu, Kaibing; Zheng, Yun; Hu, Junqing

    2014-04-21

    A low-toxic difunctional nanoplatform integrating both photothermal therapy and chemotherapy for killing cancer cells using Cu₂-xSe@mSiO₂-PEG core-shell nanoparticles is reported. Silica coating and further PEG modification improve the hydrophilicity and biocompatibility of copper selenide nanoparticles. As-prepared Cu₂-xSe@mSiO₂-PEG nanoparticles not only display strong near infrared (NIR) region absorption and good photothermal effect, but also exhibit excellent biocompatibility. The mesoporous silica shell is provided as the carrier for loading the anticancer drug, doxorubicin (DOX). Moreover, the release of DOX from Cu₂-xSe@mSiO₂-PEG core-shell nanoparticles can be triggered by pH and NIR light, resulting in a synergistic effect for killing cancer cells. Importantly, the combination of photothermal therapy and chemotherapy driven by NIR radiation with safe power density significantly improves the therapeutic efficacy, and demonstrates better therapeutic effects for cancer treatment than individual therapy.

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

    International Nuclear Information System (INIS)

    Mancini, R. C.

    2007-01-01

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

  8. Power reactor core safety research

    International Nuclear Information System (INIS)

    Rim, C.S.; Kim, W.C.; Shon, D.S.; Kim, J.

    1981-01-01

    As a part of nuclear safety research program, a project was launched to develop a model to predict fuel failure, to produce the data required for the localizaton of fuel design and fabrication technology, to establish safety limits for regulation of nuclear power plants and to develop reactor operation method to minimize fuel failure through the study of fuel failure mechanisms. During 1980, the first year of this project, various fuel failure mechanisms were analyzed, an experimental method for out-of-pile tests to study the stress corrosion cracking (SCC) behaviour of Zircaloy cladding underiodine environment was established, and characteristics of PWR and CANDU Zircaloy specimens were examined. Also developed during 1980 were the methods and correlations to evaluate fuel failures in the reactor core based on operating data from power reactors

  9. Density of basalt core from Hilo drill hole, Hawaii

    Science.gov (United States)

    Moore, J.G.

    2001-01-01

    Density measurements of 1600 samples of core from 889 to 3097 m depth below sea level in the Hawaii Scientific Drilling Program hole near Hilo, Hawaii show marked differences between the basaltic rock types and help define stratigraphy in the hole. Water-saturated densities of subaerial lava flows (occurring above 1079 m depth) have the broadest range because of the large density variation within a single lava flow. Water-saturated densities commonly range from 2.0 to 3.0 with an average of 2.55 ?? 0.24 g/cc. Dikes and sills range from 2.8 to 3.1 g/cc). Densities of hyaloclastite commonly range from 2.3 to 2.7, with an overall average of about 2.5 g/cc. The low-density of most hyaloclastite is due primarily to palagonitization of abundant glass and presence of secondary minerals in the interstices between fragments. Four principal zones of pillow lava, separated by hyaloclastite, occur in the drill core. The shallowest (1983-2136 m) is paradoxically the densest, averaging 3.01 ?? 0.10 g/cc. The second (2234-2470 m) is decidedly the lightest, averaging 2.67 ?? 0.13 g/cc. The third (2640-2790 m) and fourth (2918-bottom at 3097 m) are high, averaging 2.89 ?? 0.17 and 2.97 ?? 0.08 g/cc, respectively. The first pillow zone includes degassed pillows i.e. lava erupted on land that flowed into the sea. These pillows are poor in vesicles, because the subaerial, one-atmosphere vesicles were compressed when the flow descended to deeper water and higher pressure. The second (low-density, non-degassed) pillow zone is the most vesicle-rich, apparently because it was erupted subaqueously at a shallow depth. The higher densities of the third and fourth zones result from a low vesicularity of only a few percent and an olivine content averaging more than 5% for the third zone and about 10% for the fourth zone. The uppermost hyaloclastite extending about 400 m below the bottom of the subaerial basalt is poorly cemented and absorbs up to 6 wt% of water when immersed. Progressing

  10. Cause and countermeasure for heat up of HTTR core support plate at power rise tests

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Nozomu; Takada, Eiji; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-01-01

    HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate reached higher than expected at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in the core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses. (author)

  11. Simulating Valence-to-Core X-ray Emission Spectroscopy of Transition Metal Complexes with Time-Dependent Density Functional Theory.

    Science.gov (United States)

    Zhang, Yu; Mukamel, Shaul; Khalil, Munira; Govind, Niranjan

    2015-12-08

    Valence-to-core (VtC) X-ray emission spectroscopy (XES) has emerged as a powerful technique for the structural characterization of complex organometallic compounds in realistic environments. Since the spectrum represents electronic transitions from the ligand molecular orbitals to the core holes of the metal centers, the approach is more chemically sensitive to the metal-ligand bonding character compared with conventional X-ray absorption techniques. In this paper we study how linear-response time-dependent density functional theory (LR-TDDFT) can be harnessed to simulate K-edge VtC X-ray emission spectra reliably. LR-TDDFT allows one to go beyond the single-particle picture that has been extensively used to simulate VtC-XES. We consider seven low- and high-spin model complexes involving chromium, manganese, and iron transition metal centers. Our results are in good agreement with experiment.

  12. Core support structure for nuclear power plants

    International Nuclear Information System (INIS)

    Steinkamp, E.; Tautz, J.; Ries, H.

    1979-01-01

    A core support structure for nuclear power plants includes a grid of mutually crossing bridges and a support ring surrounding the grid and connected to ends of the outer bridges of the grid, the grid being formed of profile rod crosses having legs of given length, respective legs of pairs of adjacent crosses abutting one another endwise to form together a side of the smallest mesh opening of the grid, and weld means for securing the profile rod crosses to one another at the mutually abutting ends of the legs thereof; and method of producing the foregoing core support structure

  13. Ultrasonic density detector for vessel and reactor core two-phase flow measurements

    International Nuclear Information System (INIS)

    Arave, A.E.

    1979-01-01

    A local ultrasonic density (LUD) detector has been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory for the Loss-of-Fluid Test (LOFT) reactor vessel and core two-phase flow density measurements. The principle of operating the sensor is the change in propagation time of a torsional ultrasonic wave in a metal transmission line as a function of the density of the surrounding media. A theoretical physics model is presented which represents the total propagation time as a function of the sensor modulus of elasticity and polar moment of inertia

  14. Fuel management service for Tarapur Atomic Power Station core thermal hydraulics

    International Nuclear Information System (INIS)

    Saha, D.; Venkat Raj, V.; Markandeya, S.G.

    1977-01-01

    Core thermal hydraulic analysis forms an integral part of the fuel management service for the Tarapur reactors. A distinguishing feature of boiling water reactors is the dependence of core flow distribution on the power distribution. Because of the changes in the axial and radial power distribution from cycle to cycle as well as during the cycle and also the variations in leakage flow, it is necessary to evaluate the core thermal hydraulic parameters for every cycle. Some of the typical results obtained in the course of analysis for different cycles of both the units at Tarapur are presented. The use of MCPR (Minimum Critical Power Ratio), instead of MCHFR (Minimum Critical Heat Flux Ratio) as a figure of merit for fuel cladding integrity is also discussed. (K.B.)

  15. Safety aspects of core power distribution surveillance and control

    International Nuclear Information System (INIS)

    Beraha, D.; Grumbach, R.; Hoeld, A.; Werner, W.

    1978-01-01

    The incentives for improved core surveillance and core control systems are outlined. An efficient code for evaluating the power distribution is indispensable for designing and testing such a system. The characteristics of the core simulator QUABOX/CUBBOX and the features required for off-line and on-line applications are described. The important role of the simulator for the safety assessment of a digital core control system is underlined. With regard to the safety aspects of core control, possible disturbances are classified. Simulation results are given concerning the failure of a control actuator. It is shown that means can be devised to prevent unstable behaviour of the control system and, furthermore, to contribute to a safe reactor operation by accounting for process disturbances. (author)

  16. Microfabricated Air-core Toroidal Inductor In Very High Frequency Power Converters

    DEFF Research Database (Denmark)

    Lê Thanh, Hoà; Nour, Yasser; Han, Anpan

    2018-01-01

    Miniaturization of power supplies is required for future intelligent electronic systems e.g. internet of things devices. Inductors play an essential role, and they are by far the most bulky and expensive components in power supplies. This paper presents a miniaturized microelectromechanical systems...... (MEMS) inductor and its performance in a very high frequency (VHF) power converter. The MEMS inductor is a siliconembedded air-core toroidal inductor, and it is constructed with through-silicon vias, suspended copper windings, silicon fixtures, and a silicon support die. The air-core inductors...

  17. Measurement of the relative power density distribution of the IPEN/MB-01 reactor, using a fuel rod gamma scanning technique

    International Nuclear Information System (INIS)

    Carneiro, Alvaro Luiz Guimaraes

    1996-01-01

    This work presents a measurement methodology for determination of radial and axial relative power density distribution of the IPEN/MB-01 Reactor core by means of the fuel rod gamma scanning. The methodology is based on the proportionality between gamma activity emitted by the radioactive decay of the fission products and power density. The scanning technique consists of counting gamma radiation above 0,6 MeV along the active area of the fuel rod, getting a distribution profile. The experimental results will be used as a benchmark for qualification and to establish possible deviations for the calculational methodology currently used at IPEN. The comparison of the calculated and measured results showed good agreement. (author)

  18. Effect of Ambipolar Diffusion on Ion Abundances in Contracting Protostellar Cores

    Science.gov (United States)

    Ciolek, Glenn E.; Mouschovias, Telemachos Ch.

    1998-09-01

    Numerical simulations and analytical solutions have established that ambipolar diffusion can reduce the dust-to-gas ratio in magnetically and thermally supercritical cores during the epoch of core formation. We study the effect that this has on the ion chemistry in contracting protostellar cores and present a simplified analytical method that allows one to calculate the ion power-law exponent k (≡d ln ni/d ln nn, where ni and nn are the ion and neutral densities, respectively) as a function of core density. We find that, as in earlier numerical simulations, no single value of k can adequately describe the ion abundance for nn 1/2 during the core formation epoch (densities principle, to determine whether ambipolar diffusion is responsible for core formation in interstellar molecular clouds. For densities >>105 cm-3, k is generally <<1/2.

  19. Nitrogen-doped amorphous carbon-silicon core-shell structures for high-power supercapacitor electrodes.

    Science.gov (United States)

    Tali, S A Safiabadi; Soleimani-Amiri, S; Sanaee, Z; Mohajerzadeh, S

    2017-02-10

    We report successful deposition of nitrogen-doped amorphous carbon films to realize high-power core-shell supercapacitor electrodes. A catalyst-free method is proposed to deposit large-area stable, highly conformal and highly conductive nitrogen-doped amorphous carbon (a-C:N) films by means of a direct-current plasma enhanced chemical vapor deposition technique (DC-PECVD). This approach exploits C 2 H 2 and N 2 gases as the sources of carbon and nitrogen constituents and can be applied to various micro and nanostructures. Although as-deposited a-C:N films have a porous surface, their porosity can be significantly improved through a modification process consisting of Ni-assisted annealing and etching steps. The electrochemical analyses demonstrated the superior performance of the modified a-C:N as a supercapacitor active material, where specific capacitance densities as high as 42 F/g and 8.5 mF/cm 2 (45 F/cm 3 ) on silicon microrod arrays were achieved. Furthermore, this supercapacitor electrode showed less than 6% degradation of capacitance over 5000 cycles of a galvanostatic charge-discharge test. It also exhibited a relatively high energy density of 2.3 × 10 3  Wh/m 3 (8.3 × 10 6  J/m 3 ) and ultra-high power density of 2.6 × 10 8  W/m 3 which is among the highest reported values.

  20. Research on intelligent monitor for 3D power distribution of reactor core

    International Nuclear Information System (INIS)

    Xia, Hong; Li, Bin; Liu, Jianxin

    2014-01-01

    Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

  1. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Verma, V., E-mail: vasudha.verma@physics.uu.se [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Barbot, L.; Filliatre, P. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Hellesen, C. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Jammes, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Svärd, S. Jacobsson [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden)

    2017-07-11

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment. - Highlights: • Studied possibility of using SPNDs as in-core detectors in SFRs. • Study done to detect local power profile changes when reactor is at nominal power. • SPND with a Pt-emitter gives measurable prompt current of the order of 600 nA/m. • Dominant proportion of prompt response is maintained throughout the operation. • Detector signal gives dynamic information on the power fluctuations.

  2. High volumetric power density, non-enzymatic, glucose fuel cells.

    Science.gov (United States)

    Oncescu, Vlad; Erickson, David

    2013-01-01

    The development of new implantable medical devices has been limited in the past by slow advances in lithium battery technology. Non-enzymatic glucose fuel cells are promising replacement candidates for lithium batteries because of good long-term stability and adequate power density. The devices developed to date however use an "oxygen depletion design" whereby the electrodes are stacked on top of each other leading to low volumetric power density and complicated fabrication protocols. Here we have developed a novel single-layer fuel cell with good performance (2 μW cm⁻²) and stability that can be integrated directly as a coating layer on large implantable devices, or stacked to obtain a high volumetric power density (over 16 μW cm⁻³). This represents the first demonstration of a low volume non-enzymatic fuel cell stack with high power density, greatly increasing the range of applications for non-enzymatic glucose fuel cells.

  3. In-situ Density and Thermal Expansion Measurements of Fe and Fe-S Alloying Liquids Under Planetary Core Conditions

    Science.gov (United States)

    Jing, Z.; Chantel, J.; Yu, T.; Sakamaki, T.; Wang, Y.

    2015-12-01

    Liquid iron is likely the dominant constituent in the cores of terrestrial planets and icy satellites such as Earth, Mars, Mercury, the Moon, Ganymede, and Io. Suggested by geophysical and geochemical observations, light elements such as S, C, Si, etc., are likely present in planetary cores. These light elements can significantly reduce the density and melting temperature of the Fe cores, and hence their abundances are crucial to our understanding of the structure and thermal history of planetary cores, as well as the generation of intrinsic magnetic fields. Knowledge on the density of Fe-light element alloying liquids at high pressures is critical to place constraints on the composition of planetary cores. However, density data on liquid Fe-light element alloys at core pressures are very limited in pressure and composition and are sometimes controversial. In this study, we extend the density dataset for Fe-rich liquids by measuring the density of Fe, Fe-10wt%S, Fe-20wt%S, Fe-27wt%S, and FeS liquids using the X-ray absorption technique in a DIA-type multianvil apparatus up to 7 GPa and 2173 K. An ion chamber (1D-detector) and a CCD camera (2D-detector) were used to measure intensities of transmitted monochromatic X-rays through molten samples, with the photon energy optimized at 40 keV. The densities were then determined from the Beer-Lambert law using the mass absorption coefficients, calibrated by solid standards using X-ray diffraction. At each pressure, density measurements were conducted at a range of temperatures above the liquidus of the samples, enabling the determination of thermal expansion. Combined with our previous results on the sound velocity of Fe and Fe-S liquids at high pressures (Jing et al., 2014, Earth Planet. Sci. Lett. 396, 78-87), these data provide tight constraints on the equation of state and thermodynamic properties such as the adiabatic temperature gradient for Fe-S liquids. We will discuss these results with implications to planetary

  4. High-power picosecond pulse delivery through hollow core photonic band gap fibers

    DEFF Research Database (Denmark)

    Michieletto, Mattia; Johansen, Mette Marie; Lyngsø, Jens Kristian

    2015-01-01

    We demonstrated robust and bend insensitive fiber delivery of high power pulsed laser with diffraction limited beam quality for two different kind of hollow core photonic band gap fibers......We demonstrated robust and bend insensitive fiber delivery of high power pulsed laser with diffraction limited beam quality for two different kind of hollow core photonic band gap fibers...

  5. Effects of sintering temperature on properties of toroid cores using NiZnCu ferrites for power applications at >1 MHz

    Science.gov (United States)

    Liu, Junchang; Mei, Yunhui; Liu, Wen; Li, Xin; Hou, Feng; Lu, Guo-Quan

    2018-05-01

    The microstructures, magnetic and electronic performance of NiZnCu ferrites have been investigated at temperature from 850 °C to 1000 °C. X-ray diffraction (XRD) patterns showed that only single phase with spinel structure existed. Scanning electron microscopy (SEM) results showed that grain size increased with enhancement of sintering temperature and the most homogeneous, compact microstructure was obtained at 950 °C. Magnetic properties measurements revealed that both complex permeability and saturation magnetization increased with increasing of sintering temperature. The initial permeability was approximately linear within the scope of 850-1000 °C as well as the resonance frequency decreased from 70 MHz to 30 MHz. Power loss density tests demonstrated that the core sintered at 950 °C instead of the one sintered at 1000 °C had the lower power loss density at both 5 mT and 10 mT and the higher inductance under a certain exciting direct current at 1 MHz. Also the inductance of the sample sintered at the higher temperature dropped faster than that at the lower temperature. The results showed that the core sintered at 950 °C had better electrical performance and was suitable for wide usage.

  6. The Density Jump at the Inner Core Boundary in the Eastern and Western Hemispheres

    Science.gov (United States)

    Krasnoshchekov, D. N.; Ovtchinnikov, V. M.

    2018-02-01

    The results of analysis of more than 1300 new PKiKP/PcP amplitude ratios measured in Southeast Asia and South America at the epicentral distances of 3.2°-35.2° are presented. The density jump in the Eastern Hemisphere of the Earth's inner core (IC) is 0.3 g/cm3, and it is 0.9 g/cm3 in the Western one. Taking the large discrepancy in the obtained estimates into consideration, maintenance of such large lateral variations in the mosaic properties of the IC reflecting surface requires considerable variations in the thermodynamic parameters (mostly temperature) of the inner-outer core transition. However, if the observed asymmetry in the density jump distribution is of a global character, the data presented support the translation model of the IC dynamics. This model implies IC crystallization in the Western Hemisphere and melting in the Eastern one, not vice versa, as suggested by another geodynamic model based on thermochemical convection in the outer core and the thermal balance of the core-mantle system.

  7. Estimation of a Reactor Core Power Peaking Factor Using Support Vector Regression and Uncertainty Analysis

    International Nuclear Information System (INIS)

    Bae, In Ho; Naa, Man Gyun; Lee, Yoon Joon; Park, Goon Cherl

    2009-01-01

    The monitoring of detailed 3-dimensional (3D) reactor core power distribution is a prerequisite in the operation of nuclear power reactors to ensure that various safety limits imposed on the LPD and DNBR, are not violated during nuclear power reactor operation. The LPD and DNBR should be calculated in order to perform the two major functions of the core protection calculator system (CPCS) and the core operation limit supervisory system (COLSS). The LPD at the hottest part of a hot fuel rod, which is related to the power peaking factor (PPF, F q ), is more important than the LPD at any other position in a reactor core. The LPD needs to be estimated accurately to prevent nuclear fuel rods from melting. In this study, support vector regression (SVR) and uncertainty analysis have been applied to estimation of reactor core power peaking factor

  8. Power peaking nuclear reliability factors

    International Nuclear Information System (INIS)

    Hassan, H.A.; Pegram, J.W.; Mays, C.W.; Romano, J.J.; Woods, J.J.; Warren, H.D.

    1977-11-01

    The Calculational Nuclear Reliability Factor (CNRF) assigned to the limiting power density calculated in reactor design has been determined. The CNRF is presented as a function of the relative power density of the fuel assembly and its radial local. In addition, the Measurement Nuclear Reliability Factor (MNRF) for the measured peak hot pellet power in the core has been evaluated. This MNRF is also presented as a function of the relative power density and radial local within the fuel assembly

  9. Low-Power Embedded DSP Core for Communication Systems

    Science.gov (United States)

    Tsao, Ya-Lan; Chen, Wei-Hao; Tan, Ming Hsuan; Lin, Maw-Ching; Jou, Shyh-Jye

    2003-12-01

    This paper proposes a parameterized digital signal processor (DSP) core for an embedded digital signal processing system designed to achieve demodulation/synchronization with better performance and flexibility. The features of this DSP core include parameterized data path, dual MAC unit, subword MAC, and optional function-specific blocks for accelerating communication system modulation operations. This DSP core also has a low-power structure, which includes the gray-code addressing mode, pipeline sharing, and advanced hardware looping. Users can select the parameters and special functional blocks based on the character of their applications and then generating a DSP core. The DSP core has been implemented via a cell-based design method using a synthesizable Verilog code with TSMC 0.35[InlineEquation not available: see fulltext.]m SPQM and 0.25[InlineEquation not available: see fulltext.]m 1P5M library. The equivalent gate count of the core area without memory is approximately 50 k. Moreover, the maximum operating frequency of a[InlineEquation not available: see fulltext.] version is 100 MHz (0.35[InlineEquation not available: see fulltext.]m) and 140 MHz (0.25[InlineEquation not available: see fulltext.]m).

  10. Study of core support barrel vibration monitoring using ex-core neutron noise analysis and fuzzy logic algorithm

    International Nuclear Information System (INIS)

    Christian, Robby; Song, Seon Ho; Kang, Hyun Gook

    2015-01-01

    The application of neutron noise analysis (NNA) to the ex-core neutron detector signal for monitoring the vibration characteristics of a reactor core support barrel (CSB) was investigated. Ex-core flux data were generated by using a nonanalog Monte Carlo neutron transport method in a simulated CSB model where the implicit capture and Russian roulette technique were utilized. First and third order beam and shell modes of CSB vibration were modeled based on parallel processing simulation. A NNA module was developed to analyze the ex-core flux data based on its time variation, normalized power spectral density, normalized cross-power spectral density, coherence, and phase differences. The data were then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core neutron signal fluctuation was directly proportional to the CSB's vibration observed at 8Hz and15Hzin the beam mode vibration, and at 8Hz in the shell mode vibration. The coherence result between flux pairs was unity at the vibration peak frequencies. A distinct pattern of phase differences was observed for each of the vibration models. The developed fuzzy logic module demonstrated successful recognition of the vibration frequencies, modes, orders, directions, and phase differences within 0.4 ms for the beam and shell mode vibrations.

  11. Evaluation of neutron flux density and power density with SPN-detectors and micro calorimeters

    International Nuclear Information System (INIS)

    Gehre, G.; Rindelhardt, U.; Seidenkranz, T.; Hogel, J.; Jirousek, V.; Vazek, J.

    1983-02-01

    During investigations with a special equipped fuel assembly in the Rheinsberg nuclear power station the neutron flux and the power density were evaluated from measurements with SPN-detectors and micro calorimeters. The reliability of both detector types, their measurement accuracy under different physical conditions and the usefulness of the developed calculation models are discussed in detail. The thermal flux and the power density evaluated with SPND's agree well with theoretical results. The values obtained through micro calorimeter measurements are systematic lower by about 18%. This deviation is probably a result of differences in the used calculation models. (author)

  12. CASPER: Embedding Power Estimation and Hardware-Controlled Power Management in a Cycle-Accurate Micro-Architecture Simulation Platform for Many-Core Multi-Threading Heterogeneous Processors

    Directory of Open Access Journals (Sweden)

    Arun Ravindran

    2012-02-01

    Full Text Available Despite the promising performance improvement observed in emerging many-core architectures in high performance processors, high power consumption prohibitively affects their use and marketability in the low-energy sectors, such as embedded processors, network processors and application specific instruction processors (ASIPs. While most chip architects design power-efficient processors by finding an optimal power-performance balance in their design, some use sophisticated on-chip autonomous power management units, which dynamically reduce the voltage or frequencies of idle cores and hence extend battery life and reduce operating costs. For large scale designs of many-core processors, a holistic approach integrating both these techniques at different levels of abstraction can potentially achieve maximal power savings. In this paper we present CASPER, a robust instruction trace driven cycle-accurate many-core multi-threading micro-architecture simulation platform where we have incorporated power estimation models of a wide variety of tunable many-core micro-architectural design parameters, thus enabling processor architects to explore a sufficiently large design space and achieve power-efficient designs. Additionally CASPER is designed to accommodate cycle-accurate models of hardware controlled power management units, enabling architects to experiment with and evaluate different autonomous power-saving mechanisms to study the run-time power-performance trade-offs in embedded many-core processors. We have implemented two such techniques in CASPER–Chipwide Dynamic Voltage and Frequency Scaling, and Performance Aware Core-Specific Frequency Scaling, which show average power savings of 35.9% and 26.2% on a baseline 4-core SPARC based architecture respectively. This power saving data accounts for the power consumption of the power management units themselves. The CASPER simulation platform also provides users with complete support of SPARCV9

  13. Power Requirements Determined for High-Power-Density Electric Motors for Electric Aircraft Propulsion

    Science.gov (United States)

    Johnson, Dexter; Brown, Gerald V.

    2005-01-01

    Future advanced aircraft fueled by hydrogen are being developed to use electric drive systems instead of gas turbine engines for propulsion. Current conventional electric motor power densities cannot match those of today s gas turbine aircraft engines. However, if significant technological advances could be made in high-power-density motor development, the benefits of an electric propulsion system, such as the reduction of harmful emissions, could be realized.

  14. A Novel Method To On-Line Monitor Reactor Nuclear Power And In-Core Thermal Environments

    International Nuclear Information System (INIS)

    Liu, Hanying; Miller, Don W.; Li, Dongxu; Radcliff, Thomas D.

    2002-01-01

    For current nuclear power plants, nuclear power can not be directly measured and in-core fuel thermal environments can not be monitored due to the unavailability of an appropriate measurement technology and the inaccessibility of the fuel. If the nuclear deposited power and the in-core thermal conditions (i.e. fuel or coolant temperature and heat transfer coefficient) can be monitored in-situ, then it would play a valuable and critical role in increasing nuclear power, predicting abnormal reactor operation, improving core physical models and reducing core thermal margin so as to implement higher fuel burn-up. Furthermore, the management of core thermal margin and fuel operation may be easier during reactor operation, post-accident or spent fuel storage. On the other hand, for some advanced Generation IV reactors, the sealed and long-lived reactor core design challenges traditional measurement techniques while conventional ex-core detectors and current in-core detectors can not monitor details of the in-core fuel conditions. A method is introduced in this paper that responds to the challenge to measure nuclear power and to monitor the in-core thermal environments, for example, local fuel pin or coolant heat convection coefficient and temperature. In summary, the method, which has been designed for online in-core measurement and surveillance, will be beneficial to advanced plant safety, efficiency and economics by decreasing thermal margin or increasing nuclear power. The method was originally developed for a constant temperature power sensor (CTPS). The CTPS is undergoing design and development for an advanced reactor core to measure in-core nuclear power in measurement mode and to monitor thermal environments in compensation mode. The sensor dynamics was analyzed in compensation mode to determine the environmental temperature and the heat transfer coefficient. Previous research demonstrated that a first order dynamic model is not sufficient to simulate sensor

  15. Study on HANARO core conversion using U-Mo fuel

    International Nuclear Information System (INIS)

    Lee, K.H.; Lee, C.S.; Seo, C.G.; Park, S.J.; Kim, H.; Kim, C.K.

    2002-01-01

    Two types of fuel rods with different fuel meat diameter and uranium density are considered for HANARO core conversion with high density U-Mo fuel. Arranging standard fuels of 5.0 g U/cc and 6.35 mm in diameter at the inner ring of an assembly and reduced fuels of 4.3 g U/cc and 5.49 mm in diameter at the outer ring of an assembly flattens the assembly power distribution and avoids the increase of linear heat generation rate due to using higher uranium density and less number of fuel rods. The maximum linear heat generation rate is similar with the current reference core and four fuel sites at the outer core in the reflector tank is converted to the irradiation sites to suit more demand on fuel tests and radioisotope production at outer core sites. This new core has 32% longer fuel cycle than the current reference core. (author)

  16. Metal Amorphous Nanocomposite (MANC) Alloy Cores with Spatially Tuned Permeability for Advanced Power Magnetics Applications

    Science.gov (United States)

    Byerly, K.; Ohodnicki, P. R.; Moon, S. R.; Leary, A. M.; Keylin, V.; McHenry, M. E.; Simizu, S.; Beddingfield, R.; Yu, Y.; Feichter, G.; Noebe, R.; Bowman, R.; Bhattacharya, S.

    2018-06-01

    Metal amorphous nanocomposite (MANC) alloys are an emerging class of soft magnetic materials showing promise for a range of inductive components targeted for higher power density and higher efficiency power conversion applications including inductors, transformers, and rotating electrical machinery. Magnetization reversal mechanisms within these alloys are typically determined by composition optimization as well as controlled annealing treatments to generate a nanocomposite structure composed of nanocrystals embedded in an amorphous precursor. Here we demonstrate the concept of spatially varying the permeability within a given component for optimization of performance by using the strain annealing process. The concept is realized experimentally through the smoothing of the flux profile from the inner to outer core radius achieved by a monotonic variation in tension during the strain annealing process. Great potential exists for an extension of this concept to a wide range of other power magnetic components and more complex spatially varying permeability profiles through advances in strain annealing techniques and controls.

  17. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  18. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  19. Xenon oscillation tests in four-loop PWR cores

    International Nuclear Information System (INIS)

    Aoki, Norihiko; Osaka, Kenichi; Shimada, Shoichiro; Tochihara, Hiroshi; Machii, Seigo

    1980-01-01

    The Kansai Electric Power Co.'s OHI Unit 1 and 2 are the first 4-loop PWRs in Japan which use 17 x 17 fuel assemblies and have essentially the same plant parameters. A 4-loop core has larger core radius and higher power density than those of 2- or 3-loop cores, and is less stable for Xe oscillation. It is therefore important to confirm that Xe oscillations in radial direction are sufficiently stable in a 4-loop core. Radial and axial Xe oscillation tests were performed during the startup physics tests of OHI Unit 1 and 2; Xe oscillation was induced by perturbation of control rods and the Xe effect on power distribution observed periodically. The test results show that the transverse Xe oscillation in the 4-loop core is sufficiently stable and that the agreement between the measurement and the calculated prediction is good. (author)

  20. Improving Power Density of Free-Piston Stirling Engines

    Science.gov (United States)

    Briggs, Maxwell H.; Prahl, Joseph M.; Loparo, Kenneth A.

    2016-01-01

    Analyses and experiments demonstrate the potential benefits of optimizing piston and displacer motion in a free-piston Stirling Engine. Isothermal analysis shows the theoretical limits of power density improvement due to ideal motion in ideal Stirling engines. More realistic models based on nodal analysis show that ideal piston and displacer waveforms are not optimal, often producing less power than engines that use sinusoidal piston and displacer motion. Constrained optimization using nodal analysis predicts that Stirling engine power density can be increased by as much as 58 percent using optimized higher harmonic piston and displacer motion. An experiment is conducted in which an engine designed for sinusoidal motion is forced to operate with both second and third harmonics, resulting in a piston power increase of as much as 14 percent. Analytical predictions are compared to experimental data and show close agreement with indirect thermodynamic power calculations, but poor agreement with direct electrical power measurements.

  1. Large Core Three Branch Polymer Power Splitters

    Directory of Open Access Journals (Sweden)

    V. Prajzler

    2015-12-01

    Full Text Available We report about three branch large core polymer power splitters optimized for connecting standard plastic optical fibers. A new point of the design is insertion of a rectangle-shaped spacing between the input and the central part of the splitter, which will ensure more even distribution of the output optical power. The splitters were designed by beam propagation method using BeamPROP software. Acrylic-based polymers were used as optical waveguides being poured into the Y-grooves realized by computer numerical controlled engraving on poly(methyl methacrylate substrate. Measurement of the optical insertion losses proved that the insertion optical loss could be lowered to 2.1 dB at 650 nm and optical power coupling ratio could reach 31.8% : 37.3% : 30.9%.

  2. Implementation of 252Cf-source-driven power spectrum density measurement system

    International Nuclear Information System (INIS)

    Ren Yong; Wei Biao; Feng Peng; Li Jiansheng; Ye Cenming

    2012-01-01

    The principle of 252 Cf-source-driven power spectrum density measurement method is introduced. A measurement system and platform is realized accordingly, which is a combination of hardware and software, for measuring nuclear parameters. The detection method of neutron pulses based on an ultra-high-speed data acquisition card (three channels, 1 GHz sampling rate, 1 ns synchronization) is described, and the data processing process and the power spectrum density algorithm on PC are designed. This 252 Cf-source-driven power spectrum density measurement system can effectively obtain the nuclear tag parameters of nuclear random processes, such as correlation function and power spectrum density. (authors)

  3. Whole core calculations of power reactors by Monte Carlo method

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Mori, Takamasa

    1993-01-01

    Whole core calculations have been performed for a commercial size PWR and a prototype LMFBR by using vectorized Monte Carlo codes. Geometries of cores were precisely represented in a pin by pin model. The calculated parameters were k eff , control rod worth, power distribution and so on. Both multigroup and continuous energy models were used and the accuracy of multigroup approximation was evaluated through the comparison of both results. One million neutron histories were tracked to considerably reduce variances. It was demonstrated that the high speed vectorized codes could calculate k eff , assembly power and some reactivity worths within practical computation time. For pin power and small reactivity worth calculations, the order of 10 million histories would be necessary. Required number of histories to achieve target design accuracy were estimated for those neutronic parameters. (orig.)

  4. Core Design Concept and Core Structural Material Development for a Prototype SFR

    International Nuclear Information System (INIS)

    Chang, Jinwook

    2013-01-01

    Core design Concept: – Initial core is Uranium metal fueled core, then it will evolve into TRU core; – Tight pressure drop constraint lowers power density; – Trade-off studies with relaxed pressure drop constraint (~0.4MPa) are on-going; – Major feature will be finalized this year. • KAERI is developing advanced cladding for high burnup fuel in Ptototype SFR: – Advanced cladding materials are now developing, which shows superior high temperature mechanical property to the conventional material; – Processing technologies related to tube making process are now developed to enhance high temperature mechanical propertyl – Preliminary HT9 cladding tube was manufactured and out-of pile mechanical properties were evaluated. Advanced cladding tube is now being developed and being prepared for irradiation test

  5. High-field, high-density tokamak power reactor

    International Nuclear Information System (INIS)

    Cohn, D.R.; Cook, D.L.; Hay, R.D.; Kaplan, D.; Kreischer, K.; Lidskii, L.M.; Stephany, W.; Williams, J.E.C.; Jassby, D.L.; Okabayashi, M.

    1977-11-01

    A conceptual design of a compact (R 0 = 6.0 m) high power density (average P/sub f/ = 7.7 MW/m 3 ) tokamak demonstration power reactor has been developed. High magnetic field (B/sub t/ = 7.4 T) and moderate elongation (b/a = 1.6) permit operation at the high density (n(0) approximately 5 x 10 14 cm -3 ) needed for ignition in a relatively small plasma, with a spatially-averaged toroidal beta of only 4%. A unique design for the Nb 3 Sn toroidal-field magnet system reduces the stress in the high-field trunk region, and allows modularization for simpler disassembly. The modest value of toroidal beta permits a simple, modularized plasma-shaping coil system, located inside the TF coil trunk. Heating of the dense central plasma is attained by the use of ripple-assisted injection of 120-keV D 0 beams. The ripple-coil system also affords dynamic control of the plasma temperature during the burn period. A FLIBE-lithium blanket is designed especially for high-power-density operation in a high-field environment, and gives an overall tritium breeding ratio of 1.05 in the slowly pumped lithium

  6. Development of Liquid-Vapor Core Reactors with MHD Generator for Space Power and Propulsion Applications

    International Nuclear Information System (INIS)

    Samim Anghaie

    2002-01-01

    Any reactor that utilizes fuel consisting of a fissile material in a gaseous state may be referred to as a gaseous core reactor (GCR). Studies on GCRs have primarily been limited to the conceptual phase, mostly due to budget cuts and program cancellations in the early 1970's. A few scientific experiments have been conducted on candidate concepts, primarily of static pressure fissile gas filling a cylindrical or spherical cavity surrounded by a moderating shell, such as beryllium, heavy water, or graphite. The main interest in this area of nuclear power generation is for space applications. The interest in space applications has developed due to the promise of significant enhancement in fuel utilization, safety, plant efficiency, special high-performance features, load-following capabilities, power conversion optimization, and other key aspects of nuclear power generation. The design of a successful GCR adapted for use in space is complicated. The fissile material studied in the pa st has been in a fluorine compound, either a tetrafluoride or a hexafluoride. Both of these molecules have an impact on the structural material used in the making of a GCR. Uranium hexafluoride as a fuel allows for a lower operating temperature, but at temperatures greater than 900K becomes essentially impossible to contain. This difficulty with the use of UF6 has caused engineers and scientists to use uranium tetrafluoride, which is a more stable molecule but has the disadvantage of requiring significantly higher operating temperatures. Gas core reactors have traditionally been studied in a steady state configuration. In this manner a fissile gas and working fluid are introduced into the core, called a cavity, that is surrounded by a reflector constructed of materials such as Be or BeO. These reactors have often been described as cavity reactors because the density of the fissile gas is low and criticality is achieved only by means of the reflector to reduce neutron leakage from the core

  7. Atmospheric turbulence profiling with unknown power spectral density

    Science.gov (United States)

    Helin, Tapio; Kindermann, Stefan; Lehtonen, Jonatan; Ramlau, Ronny

    2018-04-01

    Adaptive optics (AO) is a technology in modern ground-based optical telescopes to compensate for the wavefront distortions caused by atmospheric turbulence. One method that allows to retrieve information about the atmosphere from telescope data is so-called SLODAR, where the atmospheric turbulence profile is estimated based on correlation data of Shack-Hartmann wavefront measurements. This approach relies on a layered Kolmogorov turbulence model. In this article, we propose a novel extension of the SLODAR concept by including a general non-Kolmogorov turbulence layer close to the ground with an unknown power spectral density. We prove that the joint estimation problem of the turbulence profile above ground simultaneously with the unknown power spectral density at the ground is ill-posed and propose three numerical reconstruction methods. We demonstrate by numerical simulations that our methods lead to substantial improvements in the turbulence profile reconstruction compared to the standard SLODAR-type approach. Also, our methods can accurately locate local perturbations in non-Kolmogorov power spectral densities.

  8. Thermodynamic, energy efficiency, and power density analysis of reverse electrodialysis power generation with natural salinity gradients.

    Science.gov (United States)

    Yip, Ngai Yin; Vermaas, David A; Nijmeijer, Kitty; Elimelech, Menachem

    2014-05-06

    Reverse electrodialysis (RED) can harness the Gibbs free energy of mixing when fresh river water flows into the sea for sustainable power generation. In this study, we carry out a thermodynamic and energy efficiency analysis of RED power generation, and assess the membrane power density. First, we present a reversible thermodynamic model for RED and verify that the theoretical maximum extractable work in a reversible RED process is identical to the Gibbs free energy of mixing. Work extraction in an irreversible process with maximized power density using a constant-resistance load is then examined to assess the energy conversion efficiency and power density. With equal volumes of seawater and river water, energy conversion efficiency of ∼ 33-44% can be obtained in RED, while the rest is lost through dissipation in the internal resistance of the ion-exchange membrane stack. We show that imperfections in the selectivity of typical ion exchange membranes (namely, co-ion transport, osmosis, and electro-osmosis) can detrimentally lower efficiency by up to 26%, with co-ion leakage being the dominant effect. Further inspection of the power density profile during RED revealed inherent ineffectiveness toward the end of the process. By judicious early discontinuation of the controlled mixing process, the overall power density performance can be considerably enhanced by up to 7-fold, without significant compromise to the energy efficiency. Additionally, membrane resistance was found to be an important factor in determining the power densities attainable. Lastly, the performance of an RED stack was examined for different membrane conductivities and intermembrane distances simulating high performance membranes and stack design. By thoughtful selection of the operating parameters, an efficiency of ∼ 37% and an overall gross power density of 3.5 W/m(2) represent the maximum performance that can potentially be achieved in a seawater-river water RED system with low

  9. 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

  10. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

    2017-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

  11. Measurement of power distribution in FCA-HCLWR core (Phase-1)

    International Nuclear Information System (INIS)

    Ohno, Akio; Osugi, Toshitaka; Satoh, Kunio

    1991-11-01

    In the report are described experiments with zone-type mock-up cores (FCA XIV) which consisted of uranium fuels and polystilene plates, assembled at FCA (Fast Critical Assembly), to study a series of physical characteristics for high conversion light water reactors. As one example of those characteristics, power distributions were measured by a γ-counting method in the mock-up cores by changing parametrically voidage states of the moderator, volume ratio of moderator to fuel and fuel enrichment. Fine structures of fission rate in a plate-array cell having strong heterogeneities were obtained at the center cell of the core to investigate the validity of the SRAC code for the analysis of the high conversion light water reactor. Furthermore infinite multiplication factors K ∞ which are an important physical parameter were derived from calculated migration areas and bucklings of each direction. This was obtained by fitting the measured power distributions into a cosine function. Calculated power distributions of radial directions overestimate largely the measure ones, while those of axial directions agree well with the measured values. Calculations on fine structure of fission ratio in the cell follow generally the measured values, but it is recognized that the calculation underestimates the measurement in a soft neutron spectrum core. As for infinite multiplication factors K ∞ , calculated values by the SRAC code agree within experimental errors with measured ones. No trend is observed on different voidage state of moderator and fuel enrichment in the limit of this experiment. (author)

  12. Design of 50 MWe HTR-PBMR reactor core and nuclear power plant fuel using SRAC2006 programme

    International Nuclear Information System (INIS)

    Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono

    2014-01-01

    This research aims to assess the design of core and fuel of nuclear power plant type High Temperature Reactor-Pebble Bed Modular Reactor 50 MWe from the Beginning of Life (BOL) to Ending of life (EOL) with eight years operating life. The parameters that need to be analyzed in this research are the temperature distribution inside the core, quantity enrichment of U 235 , fuel composition, criticality, and temperature reactivity coefficient of the core. The research was conducted with a data set of core design parameters such as nuclides density, core and fuel dimensions, and the axial temperature distribution inside the core. Using SRAC2006 program package, the effective multiplication factor (k eff ) values obtained from the input data that has been prepared. The results show the value of the criticality of core is proportional to the addition of U 235 enrichment. The optimum enrichment obtained at 10.125% without the use of burnable poison with an excess reactivity of 3.1 2% at BOL. The addition Gd 2O3 obtained an optimum value of 12 ppm burnable poison with an excess reactivity 0.38 %. The use of Er 2O3 with an optimum value 290 ppm has an excess reactivity 1.24 % at BOL. The core temperature reactivity coefficient with and without the use of burnable poison has a negative values that indicates the nature of its inherent safety. (author)

  13. Constraining the supersaturation density equation of state from core-collapse supernova simulations? Excluded volume extension of the baryons

    International Nuclear Information System (INIS)

    Fischer, Tobias

    2016-01-01

    In this article the role of the supersaturation density equation of state (EOS) is explored in simulations of failed core-collapse supernova explosions. Therefore the nuclear EOS is extended via a one-parameter excluded-volume description for baryons, taking into account their finite and increasing volume with increasing density in excess of saturation density. Parameters are selected such that the resulting supernova EOS represent extreme cases, with high pressure variations at supersaturation density which feature extreme stiff and soft EOS variants of the reference case, i.e. without excluded-volume corrections. Unlike in the interior of neutron stars with central densities in excess of several times saturation density, central densities of core-collapse supernovae reach only slightly above saturation density. Hence, the impact of the supersaturation density EOS on the supernova dynamics as well as the neutrino signal is found to be negligible. It is mainly determined from the low- and intermediate-density domain, which is left unmodified within this generalized excluded volume approach. (orig.)

  14. Comparison of the fractional power motor with cores made of various magnetic materials

    Science.gov (United States)

    Gmyrek, Zbigniew; Lefik, Marcin; Cavagnino, Andrea; Ferraris, Luca

    2017-12-01

    The optimization of the motor cores, coupled with new core shapes as well as powering the motor at high frequency are the primary reasons for the use of new materials. The utilization of new materials, like SMC (soft magnetic composite), reduce the core loss and/or provide quasi-isotropic core's properties in any magnetization direction. Moreover, the use of SMC materials allows for avoiding degradation of the material portions, resulting from punching process, thereby preventing the deterioration of operating parameters of the motor. The authors examine the impact of technological parameters on the properties of a new type of SMC material and analyze the possibility of its use as the core of the fractional power motor. The result of the work is an indication of the shape of the rotor core made of a new SMC material to achieve operational parameters similar to those that have a motor with a core made of laminations.

  15. Precessing vortex core in a swirling wake with heat release

    International Nuclear Information System (INIS)

    Gorbunova, A.; Klimov, A.; Molevich, N.; Moralev, I.; Porfiriev, D.; Sugak, S.; Zavershinskii, I.

    2016-01-01

    Highlights: • Precessing vortex core is left-handed co-rotated bending single-vortex structure. • The precession frequency grows with the heat-source power. • Growth of the heat-source power decreases vortex core oscillations. • The left-handed bending mode is the most unstable mode in the low-density wake. - Abstract: Numerical simulation of the non-stationary three-dimensional swirling flow is presented for an open tube with a paraxial heat source. In the considered type of swirling flows, it is shown that a precessing vortex core (PVC) appears. The obtained PVC is a left-handed co-rotated bending single-vortex structure. The influence of the heat release enhancement on parameters of PVC is investigated. Using various turbulence models (the Spalart–Allmaras, k–ω and SST models), it is shown that an increase in the heat-source power leads to an increase in the PVC frequency and to a decrease in the amplitude of PVC oscillations. Moreover, we conduct the linear stability analysis of the simplified flow model with paraxial heating (the Rankine vortex with the piecewise axial flow and density) and demonstrate that its results correspond to the results of numerical simulations rather well. In particular, we prove that the left-handed bending mode (m = +1) is the most unstable one in the low-density wake and its frequency increases with a decrease of density ratio that is similar to the behavior of precession frequency with an increase of heat-source power.

  16. ADX: a high field, high power density, Advanced Divertor test eXperiment

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Shiraiwa, S.; Terry, J.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; ADX Team

    2014-10-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment (ADX) - a tokamak specifically designed to address critical gaps in the world fusion research program on the pathway to FNSF/DEMO. This high field (6.5 tesla, 1.5 MA), high power density (P/S ~ 1.5 MW/m2) facility would utilize Alcator magnet technology to test innovative divertor concepts for next-step DT fusion devices (FNSF, DEMO) at reactor-level boundary plasma pressures and parallel heat flux densities while producing high performance core plasma conditions. The experimental platform would also test advanced lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators and wave physics at the plasma densities and magnetic field strengths of a DEMO, with the unique ability to deploy launcher structures both on the low-magnetic-field side and the high-field side - a location where energetic plasma-material interactions can be controlled and wave physics is most favorable for efficient current drive, heating and flow drive. This innovative experiment would perform plasma science and technology R&D necessary to inform the conceptual development and accelerate the readiness-for-deployment of FNSF/DEMO - in a timely manner, on a cost-effective research platform. Supported by DE-FC02-99ER54512.

  17. High power density carbonate fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Yuh, C.; Johnsen, R.; Doyon, J.; Allen, J. [Energy Research Corp., Danbury, CT (United States)

    1996-12-31

    Carbonate fuel cell is a highly efficient and environmentally clean source of power generation. Many organizations worldwide are actively pursuing the development of the technology. Field demonstration of multi-MW size power plant has been initiated in 1996, a step toward commercialization before the turn of the century, Energy Research Corporation (ERC) is planning to introduce a 2.85MW commercial fuel cell power plant with an efficiency of 58%, which is quite attractive for distributed power generation. However, to further expand competitive edge over alternative systems and to achieve wider market penetration, ERC is exploring advanced carbonate fuel cells having significantly higher power densities. A more compact power plant would also stimulate interest in new markets such as ships and submarines where space limitations exist. The activities focused on reducing cell polarization and internal resistance as well as on advanced thin cell components.

  18. Dependence of Core and Extended Flux on Core Dominance ...

    Indian Academy of Sciences (India)

    Abstract. Based on two extragalactic radio source samples, the core dominance parameter is calculated, and the correlations between the core/extended flux density and core dominance parameter are investi- gated. When the core dominance parameter is lower than unity, it is linearly correlated with the core flux density, ...

  19. Power distribution investigation in the transition phase of the low moderation type MOX fueled LWR from the high conversion core to the breeding core

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

    2011-01-01

    The key concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a core transition from a high conversion (HC) type to a plutonium breeding (BR) type in a same reactor system only by replacing fuel assemblies. Consequently in this transition phase, there are two types of assemblies in the same core. Due to the differences of the two assembly types, region-wise soft to hard neutron spectra appears and result in a large power peaking. Therefore, power distribution of FLWR in the HC to BR transition phase was studied by performing assembly and core calculations. For the whole core calculation, a new 14-group energy structure is developed to better represent the power distribution obtained with the fine 107-group structure than the 9-group structure in the previous evaluations. Calculations on few assemblies geometries show large local power peakings can be effectively reduced by considering plutonium enrichment distribution in an assembly. In the whole core calculation, there is a power level mismatch between HC and BR assemblies, but overall power distribution flattening is possible by optimizing fuel assemblies loading. Although the fuel loading should be decided also taking into account the void coefficient, transition from HC to BR type FLWR seems feasible without difficulty. (author)

  20. PIII Plasma Density Enhancement by a New DC Power Source

    International Nuclear Information System (INIS)

    Lopez-Callejas, R.; Godoy-Cabrera, O. G.; Granda-Gutierrez, E. E.; Piedad-Beneitez, A. de la; Munoz-Castro, A. E.; Valencia A, R.; Barocio, S. R.; Mercado-Cabrera, A.; Pena-Eguiluz, R.

    2006-01-01

    In practical terms, those plasmas produced by a DC voltage power supply do not attain densities above the 108 to 109 cm-3 band. Here we present a power supply, controlled in current and voltage, which has been successfully designed and constructed delivering plasma densities in the orders of 109 - 1010 cm-3. Its experimental performance test was conducted within one toroidal and one cylindrical chambers capable of 29 and 35 litres, respectively, using nitrogen gas. The DC plasma was characterized by a double electric probe. Several physical phenomena present in the PIII process have been keenly investigated including plasma sheath dynamics, interaction of plasma and surface, etc. In this paper we analyze the effect of the implantation voltage, plasma density and pulse time in the PIII average heating power and fluence density

  1. Advanced in-core monitoring system for high-power reactors

    International Nuclear Information System (INIS)

    Mitin, V.I.; Alekseev, A.N.; Golovanov, M.N.; Zorin, A.V.; Kalinushkin, A.E.; Kovel, A.I.; Milto, N.V.; Musikhin, A.M.; Tikhonova, N.V.; Filatov, V.P.

    2006-01-01

    This paper encompasses such section as objective, conception and engineering solution for construction of advanced in-core instrumentation system for high power reactor, including WWER-1000. The ICIS main task is known to be an on-line monitoring of power distribution and functionals independently of design programs to avoid a common cause error. This paper shows in what way the recovery of power distribution has been carried out using the signals from in-core neutron detectors or temperature sensors. On the basis of both measured and processed data, the signals of preventive and emergency protection on local parameters (linear power of the maximum intensive fuel rods, departure from nucleate boiling ratio peaking factor) have been automatically generated. The paper presents a detection technology and processing methods for signals from SPNDs and TCs, ICIS composition and structure, computer hardware, system and applied software. Structure, composition and the taken decisions allow combining class IE and class B and C tasks in accordance with international standards of separation and safety category realization. Nowadays, ICIS-M is a system that is capable to ensure: monitoring, safety, information display and diagnostics function, which allow securing actual increase of quality, reliability and safety in operation of nuclear fuel and power units. Meanwhile, it reduce negative influence of human factor on thermal technical reliability in the operational process (Authors)

  2. Ambient RF energy scavenging: GSM and WLAN power density measurements

    NARCIS (Netherlands)

    Visser, H.J.; Reniers, A.C.F.; Theeuwes, J.A.C.

    2009-01-01

    To assess the feasibility of ambient RF energy scavenging, a survey of expected power density levels distant from GSM-900 and GSM-1800 base stations has been conducted and power density measurements have been performed in a WLAN environment. It appears that for distances ranging from 25 m to 100 m

  3. Fundamental design bases for independent core cooling in Swedish nuclear power reactors

    International Nuclear Information System (INIS)

    Jelinek, Tomas

    2015-01-01

    New regulations on design and construction of nuclear power plants came into force in 2005. The need of an independent core cooling system and if the regulations should include such a requirement was discussed. The Swedish Radiation Safety authority (SSM) decided to not include such a requirement because of open questions about the water balance and started to investigate the consequences of an independent core cooling system. The investigation is now finished and SSM is also looking at the lessons learned from the accident in Fukushima 2011. One of the most important measures in the Swedish national action plan is the implementation of an independent core cooling function for all Swedish power plants. SSM has investigated the basic design criteria for such a function where some important questions are the level of defence in depth and the acceptance criteria. There is also a question about independence between the levels of defence in depth that SSM have included in the criteria. Another issue that has to be taken into account is the complexity of the system and the need of automation where independence and simplicity are very strong criteria. In the beginning of 2014 a memorandum was finalized regarding fundamental design bases for independent core cooling in Swedish nuclear power reactors. A decision based on this memorandum with an implementation plan will be made in the first half of 2014. Sweden is also investigating the possibility to have armed personnel on site, which is not allowed currently. The result from the investigation will have impact on the possibility to use mobile equipment and the level of protection of permanent equipment. In this paper, SSM will present the memorandum for design bases for independent core cooling in Swedish nuclear power reactors that was finalized in March 20147 that also describe SSM's position regarding independence and automation of the independent core cooling function. This memorandum describes the Swedish

  4. Edge operational space for high density/high confinement ELMY H-modes in JET

    International Nuclear Information System (INIS)

    Sartori, R.; Saibene, G.; Loarte, A.

    2002-01-01

    This paper discusses how the proximity to the L-H threshold affects the confinement of ELMy H-modes at high density. The largest reduction in confinement at high density is observed at the transition from the Type I to the Type III ELMy regime. At medium plasma triangularity, δ≅0.3 (where δ is the average triangularity at the separatrix), JET experiments show that by increasing the margin above the L-H threshold power and maintaining the edge temperature above the critical temperature for the transition to Type III ELMs, it is possible to avoid the degradation of the pedestal pressure with density, normally observed at lower power. As a result, the range of achievable densities (both in the core and in the pedestal) is increased. At high power above the L-H threshold power the core density was equal to the Greenwald limit with H97≅0.9. There is evidence that a mixed regime of Type I and Type II ELMs has been obtained at this intermediate triangularity, possibly as a result of this increase in density. At higher triangularity, δ≅0.5, the power required to achieve similar results is lower. (author)

  5. Reconstruction calculation of pin power for ship reactor core

    International Nuclear Information System (INIS)

    Li Haofeng; Shang Xueli; Chen Wenzhen; Wang Qiao

    2010-01-01

    Aiming at the limitation of the software that pin power distribution for ship reactor core was unavailable, the calculation model and method of the axial and radial pin power distribution were proposed. Reconstruction calculations of pin power along axis and radius was carried out by bicubic and bilinear interpolation and cubic spline interpolation, respectively. The results were compared with those obtained by professional reactor physical soft with fine mesh difference. It is shown that our reconstruction calculation of pin power is simple and reliable as well as accurate, which provides an important theoretic base for the safety analysis and operating administration of the ship nuclear reactor. (authors)

  6. Experimental study of particle transport and density fluctuation in LHD

    International Nuclear Information System (INIS)

    Tanaka, K.; Michael, C.; Sanin, A.

    2005-01-01

    A variety of electron density (n e ) profiles have been observed in Large Helical Device (LHD). The density profiles change dramatically with heating power and toroidal magnetic field (B t ) under the same line averaged density. The particle transport coefficients, i.e., diffusion coefficient (D) and convection velocity (V) are experimentally obtained in the standard configuration from density modulation experiments. The values of D and V are estimated separately in the core and edge. The diffusion coefficients are found to be a strong function of electron temperature (T e ) and are proportional to T e 1.7±0.9 in the core and T e 1.1±0.14 in the edge. Edge diffusion coefficients are proportional to B t -2.08 . It is found that the scaling of D in the edge is close to gyro-Bohm-like in nature. Non-zero V is observed and it is found that the electron temperature gradient can drive particle convection, particularly in the core region. The convection velocity in the core reverses direction from inward to outward as the T e gradient increases. In the edge, convection is inward directed in most cases of the present data set. It shows a modest tendency, being proportional to T e gradient and remaining inward directed. However, the toroidal magnetic field also significantly affects the value and direction of V. The density fluctuation spectrum varies with heating power suggesting that it has an influence on particle transport. The value of K sub(perpendicular) ρ i is around 0.1, as expected for gyro-Bohm diffusion. Fluctuations are localized in both positive and negative density gradient regions of the hollow density profiles. The fluctuation power in each region is clearly distinguished having different phase velocity profiles. (author)

  7. A heterogeneous multi-core platform for low power signal processing in systems-on-chip

    DEFF Research Database (Denmark)

    Paker, Ozgun; Sparsø, Jens; Haandbæk, Niels

    2002-01-01

    is based on message passing. The mini-cores are designed as parameterized soft macros intended for a synthesis based design flow. A 520.000 transistor 0.25µm CMOS prototype chip containing 6 mini-cores has been fabricated and tested. Its power consumption is only 50% higher than a hardwired ASIC and more......This paper presents a low-power and programmable DSP architecture - a heterogeneous multiprocessor platform consisting of standard CPU/DSP cores, and a set of simple instruction set processors called mini-cores each optimized for a particular class of algorithm (FIR, IIR, LMS, etc.). Communication...

  8. Comparison of the fractional power motor with cores made of various magnetic materials

    Directory of Open Access Journals (Sweden)

    Gmyrek Zbigniew

    2017-12-01

    Full Text Available The optimization of the motor cores, coupled with new core shapes as well as powering the motor at high frequency are the primary reasons for the use of new materials. The utilization of new materials, like SMC (soft magnetic composite, reduce the core loss and/or provide quasi-isotropic core’s properties in any magnetization direction. Moreover, the use of SMC materials allows for avoiding degradation of the material portions, resulting from punching process, thereby preventing the deterioration of operating parameters of the motor. The authors examine the impact of technological parameters on the properties of a new type of SMC material and analyze the possibility of its use as the core of the fractional power motor. The result of the work is an indication of the shape of the rotor core made of a new SMC material to achieve operational parameters similar to those that have a motor with a core made of laminations.

  9. Effects of Radial Reflector Composition on Core Reactivity and Peak Power

    International Nuclear Information System (INIS)

    Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

    2007-10-01

    The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future

  10. Effects of Radial Reflector Composition on Core Reactivity and Peak Power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

    2007-10-15

    The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future.

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

    International Nuclear Information System (INIS)

    Khan, L.A.; Qazi, M.K.; Bokhari, I.H.; Fazal, R.

    1989-10-01

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

  12. Liquid-poison type power controlling device for nuclear reactor

    International Nuclear Information System (INIS)

    Horiuchi, Tetsuo; Yamanari, Shozo; Sugisaki, Toshihiko; Goto, Hiroshi.

    1981-01-01

    Purpose: To improve the safety and the operability of a nuclear reactor by adjusting the density of liquid poison. Constitution: The thermal expansion follow-up failure between cladding and a pellet upon abrupt and local variations of the power is avoided by adjusting the density of liquid poison during ordinary operation in combination with a high density liquid poison tank and a filter and smoothly controlling the reactor power through a pipe installed in the reactor core. The high density liquid poison is abruptly charged in to the reactor core under relatively low pressure through the tube installed in the reactor core at the time of control rod insertion failure in an accident, thereby effectively shutting down the reactor and improving the safety and the operability of the reactor. (Yoshihara, H.)

  13. Operation manual for the core flow test loop zone power-supply controller

    Energy Technology Data Exchange (ETDEWEB)

    Harper, R.E.

    1981-11-01

    The core flow test loop, which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at ORNL, is a high-pressure, high-temperature, out-of-reactor helium circulation system that is being constructed to permit study of the performance at steady-state and transient conditions of simulated segments of core assemblies for a GCFR demonstration plant. The simulated core segments, which are divided into zones, contain electrical heating elements to simulate the heat generated by fission. To control the power which is applied to a zone, a novel multitapped transformer and zone power control system have been designed and built which satisfy stringent design criteria. The controller can match power output to demand to within better than +-1% over a 900:1 dynamic range and perform full-power transients within 1 s. The power is applied in such a way as to minimize the electromagnetic interference at the bandwidth of the loop instrumentation, and the controller incorporates several error detection techniques, making it inherently fail-safe. The operation manual describes the specifications, operating instructions, error detection capabilities, error recovery, troubleshooting, calibration and QA procedures, and maintenance requirements. Also included are sections on the theory of operation, circuitry description, and a complete set of schematics.

  14. Upgrading of the Munich reactor with a compact core

    International Nuclear Information System (INIS)

    Boening, K.; Glaeser, W.; Meier, J.; Rau, G.; Roehrmoser, A.; Zhang, L.

    1985-01-01

    An extremely small reactor core has been proposed for the project of substantial modernization of the FRM research reactor at Munich. According to the present status this 'compact core' will be a cylinder with a diameter of about 20 cm and 70 cm high. The new high-density U 3 Si/Al dispersion fuel of about 45% enrichment is contained in 20 concentric fuel plate rings. The compact core is surrounded by a large heavy-water tank which will incorporate the user installations (beam tubes and irradiation channels). However, the primary cooling circuit will contain light water which is not only more economic but also essential for the performance of the small core. An important optimization potential to decrease easily the power density peaks in the core is to reduce further the enrichment in those fuel plate rings where the neutron flux is particularly high. Two-dimensional neutron transport calculations show that such a core, containing about 7.5 kg 235 U, should have an effective multiplication factor of about 1.22 and an unperturbed but realistic maximum thermal neutron flux in the heavy water tank of 7 to 8x10 14 cm -2 .s -1 at 20 MW reactor power. (author)

  15. Investigation of Physical Processes Limiting Plasma Density in DIII--D

    Science.gov (United States)

    Maingi, R.

    1996-11-01

    Understanding the physical processes which limit operating density is crucial in achieving peak performance in confined plasmas. Studies from many of the world's tokamaks have indicated the existence(M. Greenwald, et al., Nucl. Fusion 28) (1988) 2199 of an operational density limit (Greenwald limit, n^GW_max) which is proportional to the plasma current and independent of heating power. Several theories have reproduced the current dependence, but the lack of a heating power dependence in the data has presented an enigma. This limit impacts the International Thermonuclear Experimental Reactor (ITER) because the nominal operating density for ITER is 1.5 × n^GW_max. In DIII-D, experiments are being conducted to understand the physical processes which limit operating density in H-mode discharges; these processes include X-point MARFE formation, high core recycling and neutral pressure, resistive MHD stability, and core radiative collapse. These processes affect plasma properties, i.e. edge/scrape-off layer conduction and radiation, edge pressure gradient and plasma current density profile, and core radiation, which in turn restrict the accessible density regime. With divertor pumping and D2 pellet fueling, core neutral pressure is reduced and X-point MARFE formation is effectively eliminated. Injection of the largest-sized pellets does cause transient formation of divertor MARFEs which occasionally migrate to the X-point, but these are rapidly extinguished in pumped discharges in the time between pellets. In contrast to Greenwald et al., it is found that the density relaxation time after pellets is largely independent of the density relative to the Greenwald limit. Fourier analysis of Mirnov oscillations indicates the de-stabilization and growth of rotating, tearing-type modes (m/n= 2/1) when the injected pellets cause large density perturbations, and these modes often reduce energy confinement back to L-mode levels. We are examining the mechanisms for de

  16. Effect of core configuration on the burnup calculations of MTR research reactors

    International Nuclear Information System (INIS)

    Hussein, H.M.; Amin, E.H.; Sakr, A.M.

    2014-01-01

    Highlights: • 3D burn-up calculations of MTR-type research reactor were performed. Examination of the effect of control rod pattern on power density and neutron flux distributions is presented. • The calculations are performed using the MTR P C package and the programs (WIMS and CITVAP). • An empirical formula was generated for every fuel element type, to correlate irradiation to burn-up. - Abstract: In the present paper, three-dimensional burn-up calculations were performed using different patterns of control rods, in order to examine their effect on power density and neutron flux distributions through out the entire core and hence on the local burn-up distribution. These different cores burn-up calculations are carried out for an operating cycle equivalent to 15 Full Power Days (FPDs), with a power rating of 22 MW. Calculations were performed using an example of a typical research reactor of MTR-type using the internationally known computer codes’ package “MTR P C system”, using the cell calculation transport code WIMS-D4 with 12 energy groups and the core calculation diffusion code CITVAP with 5 energy groups. A depletion study was done and the effects on the research reactor fuel (U-235) were performed. The burn-up percentage (B.U.%) curves for every fuel element type were drawn versus irradiation (MWD/TE). Then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Charts of power density and neutron flux distribution for each core were plotted at different sections of each fuel element of the reactor core. Then a complete discussion and analysis of these curves are performed with comparison between the different core configurations, illustrating the effect of insertion or extraction of either of the four control rods directly on the neutron flux and consequently on the power distribution and burn-up. A detailed study of fuel burn-up gives detailed insight on the different B.U.% calculations

  17. Active core profile and transport modification by application of Ion Bernstein Wave power in PBX-M

    International Nuclear Information System (INIS)

    LeBlanc, B.; Bell, R.

    1995-01-01

    Application of Ion Bernstein Wave Heating (IBWH) into the Princeton Beta Experiment-Modification (PBX-M) tokamak stabilizes sawtooth oscillations and generates peaked density profiles. A transport barrier, spatially correlated with the IBWH power deposition profile, is observed in the core of IBWH assisted neutral beam injection (NBI) discharges. A precursor to the fully developed barrier is seen in the soft x-ray data during edge localized mode (ELM) activity. Sustained IBWH operation is conducive to a regime where the barrier supports large triangledown n e , triangledown T e , triangledown v phi , and triangledown T i , delimiting the confinement zone. This regime is reminiscent of the H(high)-mode but with a confinement zone moved inwards. The core region has better than H-mode confinement while the peripheral region is L(low)-mode-like. The peaked profile enhanced NBI core deposition and increases nuclear reactivity. An increase in central T i results from χ i reduction (compared to H-mode) and better beam penetration. Bootstrap current fractions of up to 0.32--0.35 locally and 0.28 overall were obtained when an additional NBI burst is applied to this plasma

  18. Core-Shell-Yarn-Based Triboelectric Nanogenerator Textiles as Power Cloths.

    Science.gov (United States)

    Yu, Aifang; Pu, Xiong; Wen, Rongmei; Liu, Mengmeng; Zhou, Tao; Zhang, Ke; Zhang, Yang; Zhai, Junyi; Hu, Weiguo; Wang, Zhong Lin

    2017-12-26

    Although textile-based triboelectric nanogenerators (TENGs) are highly promising because they scavenge energy from their working environment to sustainably power wearable/mobile electronics, the challenge of simultaneously possessing the qualities of cloth remains. In this work, we propose a strategy for TENG textiles as power cloths in which core-shell yarns with core conductive fibers as the electrode and artificial polymer fibers or natural fibrous materials tightly twined around core conductive fibers are applied as the building blocks. The resulting TENG textiles are comfortable, flexible, and fashionable, and their production processes are compatible with industrial, large-scale textile manufacturing. More importantly, the comfortable TENG textiles demonstrate excellent washability and tailorability and can be fully applied in further garment processing. TENG textiles worn under the arm or foot have also been demonstrated to scavenge various types of energy from human motion, such as patting, walking, and running. All of these merits of proposed TENG textiles for clothing uses suggest their great potentials for viable applications in wearable electronics or smart textiles in the near future.

  19. The biological properties of iron oxide core high-density lipoprotein in experimental atherosclerosis

    NARCIS (Netherlands)

    Skajaa, Torjus; Cormode, David P.; Jarzyna, Peter A.; Delshad, Amanda; Blachford, Courtney; Barazza, Alessandra; Fisher, Edward A.; Gordon, Ronald E.; Fayad, Zahi A.; Mulder, Willem J. M.

    2011-01-01

    Lipoproteins are a family of plasma nanoparticles responsible for the transportation of lipids throughout the body. High-density lipoprotein (HDL), the smallest of the lipoprotein family, measures 7-13 nm in diameter and consists of a cholesteryl ester and triglyceride core that is covered with a

  20. Effect of Rigid Polyurethane Foam Core Density on Flexural and Compressive Properties of Sandwich Panels with Glass/Epoxy Faces

    Directory of Open Access Journals (Sweden)

    saeed Nemati

    2013-01-01

    Full Text Available Sandwich panels as composite materials have two external walls of either metallic or polymer type. The space between these walls is filled by hard foam or other materials and the thickness of different layers is based on the final application of the panel. In the present work, the extent of variation in core density of polyether urethane foam and subsequent flexural and compressive changes in sandwich panels with glass or epoxy face sheets are tested and investigated. A number of hard polyether urethane foams with different middle panel layers density 80-295 kg/m3 are designed to study the effect of foam density on mechanical properties including flexural and compressive properties. Flexural and compressive test resultsshow that increased core density leads to improved mechanical properties. The slope of the curve decreases beyond density of 235 kg/m3. The reason may be explained on the limitation of shear intensity in increasing the mechanical properties. In this respect an optimum density of 235 kg/m3 is obtained for the system under examinations and for reaching higher strength panels, foams of different core materials should be selected.

  1. Core-powered mass-loss and the radius distribution of small exoplanets

    Science.gov (United States)

    Ginzburg, Sivan; Schlichting, Hilke E.; Sari, Re'em

    2018-05-01

    Recent observations identify a valley in the radius distribution of small exoplanets, with planets in the range 1.5-2.0 R⊕ significantly less common than somewhat smaller or larger planets. This valley may suggest a bimodal population of rocky planets that are either engulfed by massive gas envelopes that significantly enlarge their radius, or do not have detectable atmospheres at all. One explanation of such a bimodal distribution is atmospheric erosion by high-energy stellar photons. We investigate an alternative mechanism: the luminosity of the cooling rocky core, which can completely erode light envelopes while preserving heavy ones, produces a deficit of intermediate sized planets. We evolve planetary populations that are derived from observations using a simple analytical prescription, accounting self-consistently for envelope accretion, cooling and mass-loss, and demonstrate that core-powered mass-loss naturally reproduces the observed radius distribution, regardless of the high-energy incident flux. Observations of planets around different stellar types may distinguish between photoevaporation, which is powered by the high-energy tail of the stellar radiation, and core-powered mass-loss, which depends on the bolometric flux through the planet's equilibrium temperature that sets both its cooling and mass-loss rates.

  2. Power Peaking Effect of OTTO Fuel Scheme Pebble Bed Reactor

    Science.gov (United States)

    Setiadipura, T.; Suwoto; Zuhair; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    Pebble Bed Reactor (PBR) type of Hight Temperature Gas-cooled Reactor (HTGR) is a very interesting nuclear reactor design to fulfill the growing electricity and heat demand with a superior passive safety features. Effort to introduce the PBR design to the market can be strengthen by simplifying its system with the Once-through-then-out (OTTO) cycle PBR in which the pebble fuel only pass the core once. Important challenge in the OTTO fuel scheme is the power peaking effect which limit the maximum nominal power or burnup of the design. Parametric survey is perform in this study to investigate the contribution of different design parameters to power peaking effect of OTTO cycle PBR. PEBBED code is utilized in this study to perform the equilibrium PBR core analysis for different design parameter and fuel scheme. The parameters include its core diameter, height-per-diameter (H/D), power density, and core nominal power. Results of this study show that diameter and H/D effectsare stronger compare to the power density and nominal core power. Results of this study might become an importance guidance for design optimization of OTTO fuel scheme PBR.

  3. Management of radioactive waste from a major core damage in a BWR power plant

    International Nuclear Information System (INIS)

    Elkert, J.; Christensen, H.; Torstenfelt, B.

    1990-01-01

    Large amounts of fission products would be released in case of a major core damage in a nuclear power reactor. In this theoretical study the core damage is caused by a loss of coolant accident followed by a complete loss of all electric power for about 30 minutes resulting in the release of 10% of the core inventory of noble gases. A second case has also been briefly studied, in which the corresponding core damage is supposed to be created merely by the complete loss of electric power during a limited time period. It appears from the study that the radioactive waste generated as a consequence of an accident of the extent can be managed in the reference reactor with only minor modifications required in the waste plant. The detailed results of the study are reactor specific, but many of the findings and recommendations are generally applicable. (author) 28 refs

  4. Core shift effect in blazars

    Science.gov (United States)

    Agarwal, A.; Mohan, P.; Gupta, Alok C.; Mangalam, A.; Volvach, A. E.; Aller, M. F.; Aller, H. D.; Gu, M. F.; Lähteenmäki, A.; Tornikoski, M.; Volvach, L. N.

    2017-07-01

    We studied the pc-scale core shift effect using radio light curves for three blazars, S5 0716+714, 3C 279 and BL Lacertae, which were monitored at five frequencies (ν) between 4.8 and 36.8 GHz using the University of Michigan Radio Astronomical Observatory (UMRAO), the Crimean Astrophysical Observatory (CrAO) and Metsähovi Radio Observatory for over 40 yr. Flares were Gaussian fitted to derive time delays between observed frequencies for each flare (Δt), peak amplitude (A) and their half width. Using A ∝ να, we infer α in the range of -16.67-2.41 and using Δ t ∝ ν ^{1/k_r}, we infer kr ∼ 1, employed in the context of equipartition between magnetic and kinetic energy density for parameter estimation. From the estimated core position offset (Ωrν) and the core radius (rcore), we infer that opacity model may not be valid in all cases. The mean magnetic field strengths at 1 pc (B1) and at the core (Bcore) are in agreement with previous estimates. We apply the magnetically arrested disc model to estimate black hole spins in the range of 0.15-0.9 for these blazars, indicating that the model is consistent with expected accretion mode in such sources. The power-law-shaped power spectral density has slopes -1.3 to -2.3 and is interpreted in terms of multiple shocks or magnetic instabilities.

  5. Core reset system design for linear induction accelerator

    International Nuclear Information System (INIS)

    Durga Praveen Kumar, D.; Mitra, S.; Sharma, Archana; Nagesh, K.V.; Chakravarthy, D.P.

    2006-01-01

    A repetitive pulsed power system based Linear Induction Accelerator (LIA-200) is being developed at BARC to get an electron beam of 200keV, 5kA, 50ns, 10-100 Hz. Amorphous core is the heart of these accelerators. It serves various functions in different subsystems viz. pulse power modulator, pulse transformer, magnetic switches and induction cavities. One of the factors that make the magnetic components compact is utilization of the total flux swing available in the core. In the present system, magnetic switches, pulse transformers, and induction cavity are designed to avail the full flux swing available in the core. For achieving this objective, flux density in the core has to be kept at the reverse saturation, before the main pulse is applied. The electrical circuit which makes it possible is called the core reset system. In this paper the details of core reset system designed for LIA-200 are described. (author)

  6. Design and fabrication of self-powered in-core neutron flux monitor assembly

    International Nuclear Information System (INIS)

    Chung, M.K.; Cho, S.W.; Kang, H.D.; Cho, K.K.; Cho, B.S.; Kang, S.S.

    1980-01-01

    This is the final report on the prototypical fabrication of an in-core neutron flux monitor detector assembly for a specific power reactor conducted by KAERI from July 1, 1978 to December 31, 1979. It is well known that power reactors require a large number of in-core neutron flux detector for reactor regulation and the structures of detector assemblies are different from reactor to reactor. Therefore, from the nature of this project, it should be noted here that the target model of the prototypical farbrication of an in-core neutron flux monitor detector assembly is a VFD-2 System for Wolsung CANDU. It is concluded that fabrication of in-core neutron flux monitor detector assembly for CANDU reactor is technically feasible and will bring economical benefit as much as 50 % of the unit price if they are fabricated in Korea by using partially materials which are available from local market. (author)

  7. Weld defect identification in friction stir welding using power spectral density

    Science.gov (United States)

    Das, Bipul; Pal, Sukhomay; Bag, Swarup

    2018-04-01

    Power spectral density estimates are powerful in extraction of useful information retained in signal. In the current research work classical periodogram and Welch periodogram algorithms are used for the estimation of power spectral density for vertical force signal and transverse force signal acquired during friction stir welding process. The estimated spectral densities reveal notable insight in identification of defects in friction stir welded samples. It was observed that higher spectral density against each process signals is a key indication in identifying the presence of possible internal defects in the welded samples. The developed methodology can offer preliminary information regarding presence of internal defects in friction stir welded samples can be best accepted as first level of safeguard in monitoring the friction stir welding process.

  8. Methods to enhance blanket power density

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Miller, L.G.; Bohn, T.S.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Wessol, D.E.; Abdou, M.A.

    1982-06-01

    The overall objective of this task is to investigate the extent to which the power density in the FED/INTOR breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow advanced reactor blanket modules to be tested on FED/INTOR under representative conditions

  9. Increasing the neutron flux study for the TRR-II core design

    International Nuclear Information System (INIS)

    Chen, C.-H.; Yang, J.-T.; Chou, Y.-C.

    1999-01-01

    The maximum unperturbed thermal flux of the originally proposed core design, which is a 6x6 square arrangement with power level of 20 MW and has been presented at the 6th Meeting of IGORR, for the TRR-II reactor is about 2.0x10 14 n/cm 2 -sec. However, it is no longer satisfied the user's requirement, that is, it must reach at least 2.5x10 14 n/cm 2 -sec. In order to enhance the thermal neutron flux, one of the most effective ways is to increase the average power density. Therefore, two new designs with more compact cores are then proposed and studied. One is 5x6 rectangular arrangement with power of 20 MW; the other one is 5x5 square arrangement with power of 16 MW. It is for sure that both core designs can satisfy thermal hydraulic safety limits. The designed parameters related to neutronics are listed and compared fundamentally. According to our calculation, although both cores have similar average power density, the results show that the 5x6/20 MW design has the maximum unperturbed thermal flux in the D 2 O region about 2.7x10 14 n/cm 2 -sec, and the 5x5/16 MW design has 2.5x10 14 n/cm 2 -sec. The maximum thermal flux in the neighborhood of the longer side of the 5x6 core is about 7% higher than the one in the neighborhood of any side of the 5x5 core. This 'long-side effect' gives the 5x6/20 MW core design an advantage of the utilization of the thermal neutron flux in the D 2 O region. In addition, the 5x5 core is also more sensitive to the reactivity change on account of in-core irradiation test facilities. Therefore, under overall considerations the 5x6/20 MW core design is chosen for further detailed design. (author)

  10. Application of MSHIM core control strategy for westinghouse AP1000 nuclear power plant

    International Nuclear Information System (INIS)

    Onoue, Masaaki; Kawanishi, Tomohiro; Carlson, William R.; Morita, Toshio

    2003-01-01

    Westinghouse has developed a new core control strategy, in which two independently moving Rod Cluster Control Assembly (RCCA) groups are utilized for core control; one group for reactivity/temperature control, the other for axial power distribution (Axial Offset) control. This control procedure eliminates the need for Chemical Shim adjustments during power maneuvers, such as load follow, and is designated MSHIM (Mechanical Shim). This core control strategy is utilized in the AP1000. In the AP1000, it is possible to perform MSHIM load follow maneuvers for up to 95% of cycle life without changing the soluble boron concentration in the moderator. This core control strategy has been evaluated, via computer simulations, to provide appropriate margins to core and fuel design limits during normal operation maneuvers (including load follow operations) and also during anticipated Condition II accident transients. The primary benefits of MSHIM as a control strategy are as follows; Power change operation can be totally automated due to the elimination of boron concentration adjustments. Full load follow capability is achievable for up to more than 95% of cycle life. Load follow operations performed solely by mechanical devices results in a significant reduction in the boron system requirements and a significant reduction in daily effluent to be processed. (author)

  11. Mechanical and Physical Properties of Low Density Kenaf Core Particleboards Bonded with Different Resins

    OpenAIRE

    Mohamad Jani Saad; Izran Kamal

    2012-01-01

    Single layer kenaf (Hibiscus cannabinus) core particleboards bonded with urea formaldehyde (UF), phenol formaldehyde (PF) and polymeric 4,4-methyl phenylmethane di-isocyanate (PMDI) resins were manufactured. The boards were fabricated with three different densities i.e 350 kg/m3, 450 kg/m3 and 550 kg/m3. Each type of the resin used was sprayed at three different resin loadings on the kenaf core particles. The boards produced was evaluated for its modulus of rupture (MOR), modulus of elasticit...

  12. The structure of protostellar dense cores: a millimeter continuum study

    International Nuclear Information System (INIS)

    Motte, Frederique

    1998-01-01

    A comprehensive theoretical scenario explains low-mass star formation and describes the gravitational collapse of an isolated 'ideal' dense core. The major aim of this thesis is to check the standard model predictions on the structure of protostellar dense cores (or envelopes). The earliest stages of star formation remain poorly known because the protostars are still deeply embedded in massive, opaque circumstellar cocoons. On the one hand, sensitive bolometer arrays very recently allow us to measure the millimeter continuum emission arising from dense cores. Such observations are a powerful tool to constrain the density structure of proto-stellar dense cores (on large length scale). In particular, we studied the structure of isolated proto-stellar envelopes in Taurus and protostars in the ρ Ophiuchi cluster. In order to accurately derive their envelope density power law, we simulated the observation of several envelope models. Then we show that most of the Taurus protostars present a density structure consistent with the standard model predictions. In contrast, dense cores in ρ Ophiuchi main cloud are highly fragmented and protostellar envelope have finite size. Moreover fragmentation appears to be essential in determining the final stellar mass of ρ Oph forming stars. In clusters, fragmentation may thus be at the origin of the stellar initial mass function (IMF). On the other hand, our interferometric millimeter continuum observations are tracing (with higher angular resolution) the inner part of protostellar envelopes. Our study show that disks during protostellar stages are not yet massive and thus do not perturb the analysis of envelope density structure. (author) [fr

  13. Dimmable electronic ballasts by variable power density modulation technique

    Science.gov (United States)

    Borekci, Selim; Kesler, Selami

    2014-11-01

    Dimming can be accomplished commonly by switching frequency and pulse density modulation techniques and a variable inductor. In this study, a variable power density modulation (VPDM) control technique is proposed for dimming applications. A fluorescent lamp is operated in several states to meet the desired lamp power in a modulation period. The proposed technique has the same advantages of magnetic dimming topologies have. In addition, a unique and flexible control technique can be achieved. A prototype dimmable electronic ballast is built and experiments related to it have been conducted. As a result, a 36WT8 fluorescent lamp can be driven for a desired lamp power from several alternatives without modulating the switching frequency.

  14. Nonimaging optical designs for maximum-power-density remote irradiation.

    Science.gov (United States)

    Feuermann, D; Gordon, J M; Ries, H

    1998-04-01

    Designs for flexible, high-power-density, remote irradiation systems are presented. Applications include industrial infrared heating such as in semiconductor processing, alternatives to laser light for certain medical procedures, and general remote high-brightness lighting. The high power densities in herent to the small active radiating regions of conventional metal-halide, halogen, xenon, microwave-sulfur, and related lamps can be restored with nonimaging concentrators with little loss of power. These high fluxlevels can then be transported at high transmissivity with light channels such as optical fibers or lightpipes, and reshaped into luminaires that can deliver prescribed angular and spatial flux distributions onto desired targets. Details for nominally two- and three-dimensional systems are developed, along with estimates ofoptical performance.

  15. Calculation of ex-core detector responses

    Energy Technology Data Exchange (ETDEWEB)

    Wouters, R. de; Haedens, M. [Tractebel Engineering, Brussels (Belgium); Baenst, H. de [Electrabel, Brussels (Belgium)

    2005-07-01

    The purpose of this work carried out by Tractebel Engineering, is to develop and validate a method for predicting the ex-core detector responses in the NPPs operated by Electrabel. Practical applications are: prediction of ex-core calibration coefficients for startup power ascension, replacement of xenon transients by theoretical predictions, and analysis of a Rod Drop Accident. The neutron diffusion program PANTHER calculates node-integrated fission sources which are combined with nodal importance representing the contribution of a neutron born in that node to the ex-core response. These importance are computed with the Monte Carlo program MCBEND in adjoint mode, with a model of the whole core at full power. Other core conditions are treated using sensitivities of the ex-core responses to water densities, computed with forward Monte Carlo. The Scaling Factors (SF), or ratios of the measured currents to the calculated response, have been established on a total of 550 in-core flux maps taken in four NPPs. The method has been applied to 15 startup transients, using the average SF obtained from previous cycles, and to 28 xenon transients, using the SF obtained from the in-core map immediately preceding the transient. The values of power (P) and axial offset (AOi) reconstructed with the theoretical calibration agree well with the measured values. The ex-core responses calculated during a rod drop transient have been successfully compared with available measurements, and with theoretical data obtained by alternative methods. In conclusion, the method is adequate for the practical applications previously listed. (authors)

  16. Dark matter halos with cores from hierarchical structure formation

    International Nuclear Information System (INIS)

    Strigari, Louis E.; Kaplinghat, Manoj; Bullock, James S.

    2007-01-01

    We show that dark matter emerging from late decays (z or approx. 0.1 Mpc), and simultaneously generates observable constant-density cores in small dark matter halos. We refer to this class of models as meta-cold dark matter (mCDM), because it is born with nonrelativistic velocities from the decays of cold thermal relics. The constant-density cores are a result of the low phase-space density of mCDM at birth. Warm dark matter cannot produce similar size phase-space limited cores without saturating the Lyα power spectrum bounds. Dark matter-dominated galaxy rotation curves and stellar velocity dispersion profiles may provide the best means to discriminate between mCDM and CDM. mCDM candidates are motivated by the particle spectrum of supersymmetric and extra dimensional extensions to the standard model of particle physics

  17. Accurate core-electron binding energy shifts from density functional theory

    International Nuclear Information System (INIS)

    Takahata, Yuji; Marques, Alberto Dos Santos

    2010-01-01

    Current review covers description of density functional methods of calculation of accurate core-electron binding energy (CEBE) of second and third row atoms; applications of calculated CEBEs and CEBE shifts (ΔCEBEs) in elucidation of topics such as: hydrogen-bonding, peptide bond, polymers, DNA bases, Hammett substituent (σ) constants, inductive and resonance effects, quantitative structure activity relationship (QSAR), and solid state effect (WD). This review limits itself to works of mainly Chong and his coworkers for the period post-2002. It is not a fully comprehensive account of the current state of the art.

  18. Effect of graphite target power density on tribological properties of graphite-like carbon films

    Science.gov (United States)

    Dong, Dan; Jiang, Bailing; Li, Hongtao; Du, Yuzhou; Yang, Chao

    2018-05-01

    In order to improve the tribological performance, a series of graphite-like carbon (GLC) films with different graphite target power densities were prepared by magnetron sputtering. The valence bond and microstructure of films were characterized by AFM, TEM, XPS and Raman spectra. The variation of mechanical and tribological properties with graphite target power density was analyzed. The results showed that with the increase of graphite target power density, the deposition rate and the ratio of sp2 bond increased obviously. The hardness firstly increased and then decreased with the increase of graphite target power density, whilst the friction coefficient and the specific wear rate increased slightly after a decrease with the increasing graphite target power density. The friction coefficient and the specific wear rate were the lowest when the graphite target power density was 23.3 W/cm2.

  19. An improved one-and-a-half group BWR core simulator for a new-generation core management system

    International Nuclear Information System (INIS)

    Iwamoto, Tatsuya; Yamamoto, Munenari

    2000-01-01

    An improved one-and-a-half group core simulator method for a next-generation BWR core management system is presented. In the improved method, intranodal spectral index (thermal to fast flux ratio) is expanded with analytic solutions to the diffusion equation, and the nodal power density and the interface net current are calculated, taking the intranodal flux shape into consideration. A unique method was developed for assembly heterogeneity correction. Thus eliminating the insufficiencies of the conventional one-and-a-half group method, we can have accurate power distributions as well as local peaking factors for cores having large spectral mismatch between fuel assemblies. The historical effects of spectral mismatch are also considered in both nodal power and local peaking calculations. Although reflectors are not solved explicitly, there is essentially no need for core dependent adjustable parameters, since boundary conditions are derived in the same manner as in the interior nodes. Calculation time for nodal solutions is comparable to that for the conventional method, and is less than 1/10 of a few-group nodal simulator. Verifications of the present method were made by comparing the results with those obtained by heterogeneous fine-mesh multi-group core depletion calculations, and the accuracy was shown to be fairly good. (author)

  20. Automated ligand fitting by core-fragment fitting and extension into density

    International Nuclear Information System (INIS)

    Terwilliger, Thomas C.; Klei, Herbert; Adams, Paul D.; Moriarty, Nigel W.; Cohn, Judith D.

    2006-01-01

    An automated ligand-fitting procedure has been developed and tested on 9327 ligands and (F o − F c )exp(iϕ c ) difference density from macromolecular structures in the Protein Data Bank. A procedure for fitting of ligands to electron-density maps by first fitting a core fragment of the ligand to density and then extending the remainder of the ligand into density is presented. The approach was tested by fitting 9327 ligands over a wide range of resolutions (most are in the range 0.8-4.8 Å) from the Protein Data Bank (PDB) into (F o − F c )exp(iϕ c ) difference density calculated using entries from the PDB without these ligands. The procedure was able to place 58% of these 9327 ligands within 2 Å (r.m.s.d.) of the coordinates of the atoms in the original PDB entry for that ligand. The success of the fitting procedure was relatively insensitive to the size of the ligand in the range 10–100 non-H atoms and was only moderately sensitive to resolution, with the percentage of ligands placed near the coordinates of the original PDB entry for fits in the range 58–73% over all resolution ranges tested

  1. Power Burst Facility: power oscillation problem

    International Nuclear Information System (INIS)

    Lussie, W.G.; Wadkins, R.P.; Wells, R.A.

    1975-01-01

    In late 1973 PBF achieved a power level of 15 MW. During this period of operation fluctuations in reactor power were observed. Many possible causes of these fluctuations were considered and a number of nuclear and non-nuclear tests were conducted. Initial instrumentation installed in the core showed coolant outlet temperature variations of 10 0 F for several fuel cannisters and approximately 10 percent power variations at 15 MW. Power spectral density analysis showed a predominant frequency of 0.05 to 0.06 HZ. The testing program to determine the cause of the power oscillations is described

  2. Nonstructural protein 5A is incorporated into hepatitis C virus low-density particle through interaction with core protein and microtubules during intracellular transport.

    Directory of Open Access Journals (Sweden)

    Chao-Kuen Lai

    Full Text Available Nonstructural protein 5A (NS5A of hepatitis C virus (HCV serves dual functions in viral RNA replication and virus assembly. Here, we demonstrate that HCV replication complex along with NS5A and Core protein was transported to the lipid droplet (LD through microtubules, and NS5A-Core complexes were then transported from LD through early-to-late endosomes to the plasma membrane via microtubules. Further studies by cofractionation analysis and immunoelectron microscopy of the released particles showed that NS5A-Core complexes, but not NS4B, were present in the low-density fractions, but not in the high-density fractions, of the HCV RNA-containing virions and associated with the internal virion core. Furthermore, exosomal markers CD63 and CD81 were also detected in the low-density fractions, but not in the high-density fractions. Overall, our results suggest that HCV NS5A is associated with the core of the low-density virus particles which exit the cell through a preexisting endosome/exosome pathway and may contribute to HCV natural infection.

  3. Density dependence of SOL power width in ASDEX upgrade L-Mode

    Directory of Open Access Journals (Sweden)

    B. Sieglin

    2017-08-01

    A recent study [4] with an open divertor configuration found an asymmetry of the power fall-off length between inner and outer target with a smaller power fall-off length λq,i on the inner divertor target. Measurements with a closed divertor configuration find a similar asymmetry for low recycling divertor conditions. It is found, in the experiment, that the in/out asymmetry λq,i/λq,o is strongly increasing with increasing density. Most notably the heat flux density at the inner divertor target is reducing with increasing λq,i whilst the total power onto each divertor target stays constant. It is found that λq,o exhibits no significant density dependence for hydrogen and deuterium but increases with about the square root of the electron density for helium. The difference between H,D and He could be due to the different recycling behaviour in the divertor. These findings may help current modelling attempts to parametrize the density dependence of the widening of the power channel and thus allow for detailed comparison to both divertor effects like recycling or increased upstream SOL cross field transport.

  4. NASA Glenn Research Center Program in High Power Density Motors for Aeropropulsion

    Science.gov (United States)

    Brown, Gerald V.; Kascak, Albert F.; Ebihara, Ben; Johnson, Dexter; Choi, Benjamin; Siebert, Mark; Buccieri, Carl

    2005-01-01

    Electric drive of transport-sized aircraft propulsors, with electric power generated by fuel cells or turbo-generators, will require electric motors with much higher power density than conventional room-temperature machines. Cryogenic cooling of the motor windings by the liquid hydrogen fuel offers a possible solution, enabling motors with higher power density than turbine engines. Some context on weights of various systems, which is required to assess the problem, is presented. This context includes a survey of turbine engine weights over a considerable size range, a correlation of gear box weights and some examples of conventional and advanced electric motor weights. The NASA Glenn Research Center program for high power density motors is outlined and some technical results to date are presented. These results include current densities of 5,000 A per square centimeter current density achieved in cryogenic coils, finite element predictions compared to measurements of torque production in a switched reluctance motor, and initial tests of a cryogenic switched reluctance motor.

  5. The influence of changes in the VVER-1000 fuel assembly shape during operation on the power density distribution

    Energy Technology Data Exchange (ETDEWEB)

    Shishkov, L. K., E-mail: Shishkov-LK@nrcki.ru; Gorodkov, S. S.; Mikailov, E. F.; Sukhino-Homenko, E. A.; Sumarokova, A. S., E-mail: Sumarokova-AS@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2016-12-15

    A new approach to calculation of the coefficients of sensitivity of the fuel pin power to deviations in gap sizes between fuel assemblies of the VVER-1000 reactor during its operation is proposed. It is shown that the calculations by the MCU code should be performed for a full-size model of the core to take the interference of the gap influence into account. In order to reduce the conservatism of calculations, the coolant density and coolant temperature feedbacks should be taken into account, as well as the fuel burnup.

  6. Effective Domain Partitioning for Multi-Clock Domain IP Core Wrapper Design under Power Constraints

    Science.gov (United States)

    Yu, Thomas Edison; Yoneda, Tomokazu; Zhao, Danella; Fujiwara, Hideo

    The rapid advancement of VLSI technology has made it possible for chip designers and manufacturers to embed the components of a whole system onto a single chip, called System-on-Chip or SoC. SoCs make use of pre-designed modules, called IP-cores, which provide faster design time and quicker time-to-market. Furthermore, SoCs that operate at multiple clock domains and very low power requirements are being utilized in the latest communications, networking and signal processing devices. As a result, the testing of SoCs and multi-clock domain embedded cores under power constraints has been rapidly gaining importance. In this research, a novel method for designing power-aware test wrappers for embedded cores with multiple clock domains is presented. By effectively partitioning the various clock domains, we are able to increase the solution space of possible test schedules for the core. Since previous methods were limited to concurrently testing all the clock domains, we effectively remove this limitation by making use of bandwidth conversion, multiple shift frequencies and properly gating the clock signals to control the shift activity of various core logic elements. The combination of the above techniques gains us greater flexibility when determining an optimal test schedule under very tight power constraints. Furthermore, since it is computationally intensive to search the entire expanded solution space for the possible test schedules, we propose a heuristic 3-D bin packing algorithm to determine the optimal wrapper architecture and test schedule while minimizing the test time under power and bandwidth constraints.

  7. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    Science.gov (United States)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  8. Improved water density feedback model for pressurized water reactors

    International Nuclear Information System (INIS)

    Casadei, A.L.

    1976-01-01

    An improved water density feedback model has been developed for neutron diffusion calculations of PWR cores. This work addresses spectral effects on few-group cross sections due to water density changes, and water density predictions considering open channel and subcooled boiling effects. An homogenized spectral model was also derived using the unit assembly diffusion method for employment in a coarse mesh 3D diffusion computer program. The spectral and water density evaluation models described were incorporated in a 3D diffusion code, and neutronic calculations for a typical PWR were completed for both nominal and accident conditions. Comparison of neutronic calculations employing the open versus the closed channel model for accident conditions indicates that significant safety margin increases can be obtained if subcooled boiling and open channel effects are considered in accident calculations. This is attributed to effects on both core reactivity and power distribution, which result in increased margin to fuel degradation limits. For nominal operating conditions, negligible differences in core reactivity and power distribution exist since flow redistribution and subcooled voids are not significant at such conditions. The results serve to confirm the conservatism of currently employed closed channel feedback methods in accident analysis, and indicate that the model developed in this work can contribute to show increased safety margins for certain accidents

  9. Integrated core-edge-divertor modeling studies

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

    An integrated calculation model for simulating the interaction of physics phenomena taking place in the plasma core, in the plasma edge and in the SOL and divertor of tokamaks has been developed and applied to study such interactions. The model synthesises a combination of numerical calculations (1) the power and particle balances for the core plasma, using empirical confinement scaling laws and taking into account radiation losses (2), the particle, momentum and power balances in the SOL and divertor, taking into account the effects of radiation and recycling neutrals, (3) the transport of feeling and recycling neutrals, explicitly representing divertor and pumping geometry, and (4) edge pedestal gradient scale lengths and widths, evaluation of theoretical predictions (5) confinement degradation due to thermal instabilities in the edge pedestals, (6) detachment and divertor MARFE onset, (7) core MARFE onsets leading to a H-L transition, and (8) radiative collapse leading to a disruption and evaluation of empirical fits (9) power thresholds for the L-H and H-L transitions and (10) the width of the edge pedestals. The various components of the calculation model are coupled and must be iterated to a self-consistent convergence. The model was developed over several years for the purpose of interpreting various edge phenomena observed in DIII-D experiments and thereby, to some extent, has been benchmarked against experiment. Because the model treats the interactions of various phenomena in the core, edge and divertor, yet is computationally efficient, it lends itself to the investigation of the effects of different choices of various edge plasma operating conditions on overall divertor and core plasma performance. Studies of the effect of feeling location and rate, divertor geometry, plasma shape, pumping and over 'edge parameters' on core plasma properties (line average density, confinement, density limit, etc.) have been performed for DIII-D model problems. A

  10. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas V, G.; Banfield, J. [GE-Hitachi Nuclear Energy Americas LLC, Global Nuclear Fuel, Americas LLC, 3901 Castle Hayne Road, Wilmingtonm, North Carolina (United States); Avila N, A., E-mail: Gabriel.Cuevas-Vivas@ge.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2016-09-15

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  11. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    International Nuclear Information System (INIS)

    Cuevas V, G.; Banfield, J.; Avila N, A.

    2016-09-01

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  12. Power controlling method for BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1983-01-01

    Purpose: To enable reactor operation exactly following after an aimed curve in the high power resuming and maintaining period without failures in cladding tubes. Method: Upon recovery of the reactor power to a high power level after changing the reactor power from the high power to the low power level, control rod is operated under such conditions that the linear power density after operation of the control rod does not exceed the PC envelope in the low power period, and the core flow rate is coordinated to the control rod operation. The linear power density can be suppressed within an allowable linear power density by the above operation during high power resuming and maintaining period and, as the result, PCI failures can be prevented. (Kamimura, M.)

  13. Monitoring device for the reactor power distribution

    International Nuclear Information System (INIS)

    Uematsu, Hitoshi; Tsuiki, Makoto

    1982-01-01

    Purpose: To enable accurate monitoring for the power distribution in a short time, as well as independent detection for in-core neutron flux detectors in abnormal operation due to failures or like other causes to thereby surely provide reliable substitute values. Constitution: Counted values are inputted from a reactor core present status data detector by a power distribution calculation device to calculate the in-core neutron flux density and the power distribution based on previously stored physical models. While on the other hand, counted value from the in-core neutron detectors and the neutron flux distribution and the power distribution calculated from the power distribution calculation device are inputted from a BCF calculation device to compensate the counting errors incorporated in the counted value from the in-core neutron flux detectors and the calculation errors incorporated in the power distribution calculated in the power distribution calculation device respectively and thereby calculate the power distribution in the reactor core. Further, necessary data are inputted to the power distribution calculation device by an input/output device and the results calculated in the BCF calculation device are displayed. (Aizawa, K.)

  14. COLLAPSE AND FRAGMENTATION OF MAGNETIC MOLECULAR CLOUD CORES WITH THE ENZO AMR MHD CODE. I. UNIFORM DENSITY SPHERES

    International Nuclear Information System (INIS)

    Boss, Alan P.; Keiser, Sandra A.

    2013-01-01

    Magnetic fields are important contributors to the dynamics of collapsing molecular cloud cores, and can have a major effect on whether collapse results in a single protostar or fragmentation into a binary or multiple protostar system. New models are presented of the collapse of magnetic cloud cores using the adaptive mesh refinement code Enzo2.0. The code was used to calculate the ideal magnetohydrodynamics (MHD) of initially spherical, uniform density, and rotation clouds with density perturbations, i.e., the Boss and Bodenheimer standard isothermal test case for three-dimensional (3D) hydrodynamics codes. After first verifying that Enzo reproduces the binary fragmentation expected for the non-magnetic test case, a large set of models was computed with varied initial magnetic field strengths and directions with respect to the cloud core axis of rotation (parallel or perpendicular), density perturbation amplitudes, and equations of state. Three significantly different outcomes resulted: (1) contraction without sustained collapse, forming a denser cloud core; (2) collapse to form a single protostar with significant spiral arms; and (3) collapse and fragmentation into binary or multiple protostar systems, with multiple spiral arms. Comparisons are also made with previous MHD calculations of similar clouds with a barotropic equations of state. These results for the collapse of initially uniform density spheres illustrate the central importance of both magnetic field direction and field strength for determining the outcome of dynamic protostellar collapse.

  15. Solution of the 'MIDICORE' WWER-1000 core periphery power distribution benchmark by KARATE and MCNP

    International Nuclear Information System (INIS)

    Temesvari, E.; Hegyi, G.; Hordosy, G.; Maraczy, C.

    2011-01-01

    The 'MIDICORE' WWER-1000 core periphery power distribution benchmark was proposed by Mr. Mikolas on the twentieth Symposium of AER in Finland in 2010. This MIDICORE benchmark is a two-dimensional calculation benchmark based on the WWER-1000 reactor core cold state geometry with taking into account the geometry of explicit radial reflector. The main task of the benchmark is to test the pin by pin power distribution in selected fuel assemblies at the periphery of the WWER-1000 core. In this paper we present our results (k eff , integral fission power) calculated by MCNP and the KARATE code system in KFKI-AEKI and the comparison to the preliminary reference Monte Carlo calculation results made by NRI, Rez. (Authors)

  16. Effect of low density H-mode operation on edge and divertor plasma parameters

    International Nuclear Information System (INIS)

    Maingi, R.; Mioduszewski, P.K.; Cuthbertson, J.W.

    1994-07-01

    We present a study of the impact of H-mode operation at low density on divertor plasma parameters on the DIII-D tokamak. The line-average density in H-mode was scanned by variation of the particle exhaust rate, using the recently installed divertor cryo-condensation pump. The maximum decrease (50%) in line-average electron density was accompanied by a factor of 2 increase in the edge electron temperature, and 10% and 20% reductions in the measured core and divertor radiated power, respectively. The measured total power to the inboard divertor target increased by a factor of 3, with the major contribution coming from a factor of 5 increase in the peak heat flux very close to the inner strike point. The measured increase in power at the inboard divertor target was approximately equal to the measured decrease in core and divertor radiation

  17. Neutron flux and power in RTP core-15

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis; Bayar, Abi Muttaqin Jalal; Hamzah, Na’im Syauqi Bin [Nuclear and reactor Physics Section, Nuclear Technology Center, Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core with literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.

  18. A validation report for the KALIMER core design computing system by the Monte Carlo transport theory code

    International Nuclear Information System (INIS)

    Lee, Ki Bog; Kim, Yeong Il; Kim, Kang Seok; Kim, Sang Ji; Kim, Young Gyun; Song, Hoon; Lee, Dong Uk; Lee, Byoung Oon; Jang, Jin Wook; Lim, Hyun Jin; Kim, Hak Sung

    2004-05-01

    In this report, the results of KALIMER (Korea Advanced LIquid MEtal Reactor) core design calculated by the K-CORE computing system are compared and analyzed with those of MCDEP calculation. The effective multiplication factor, flux distribution, fission power distribution and the number densities of the important nuclides effected from the depletion calculation for the R-Z model and Hex-Z model of KALIMER core are compared. It is confirmed that the results of K-CORE system compared with those of MCDEP based on the Monte Carlo transport theory method agree well within 700 pcm for the effective multiplication factor estimation and also within 2% in the driver fuel region, within 10% in the radial blanket region for the reaction rate and the fission power density. Thus, the K-CORE system for the core design of KALIMER by treating the lumped fission product and mainly important nuclides can be used as a core design tool keeping the necessary accuracy

  19. Time delays between core power production and external detector response from Monte Carlo calculations

    International Nuclear Information System (INIS)

    Valentine, T.E.; Mihalczo, J.T.

    1996-01-01

    One primary concern for design of safety systems for reactors is the time response of external detectors to changes in the core. This paper describes a way to estimate the time delay between the core power production and the external detector response using Monte Carlo calculations and suggests a technique to measure the time delay. The Monte Carlo code KENO-NR was used to determine the time delay between the core power production and the external detector response for a conceptual design of the Advanced Neutron Source (ANS) reactor. The Monte Carlo estimated time delay was determined to be about 10 ms for this conceptual design of the ANS reactor

  20. Comparison study on in-core neutron detector for online neutron flux mapping of research and power reactor

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Mohd Idris Taib; Izhar Abu Husin; Nurfarhana Ayuni

    2010-01-01

    This paper presents the comparison study on In-Core neutron detector using for online flux mapping of Research and Power reactor. Technical description of in-core neutron also taken into consideration to identify the different characterization of neutron detector and describe on Self Power neutron detector (SPND) for online neutron flux mapping. Able to provide information on the neutron flux distribution and understand how in-core neutron detector are being used in nuclear power plant including to enable to state the principles of neutron detector. (author)

  1. Emergency core cooling systems in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1981-12-01

    This report contains the responses by the Advisory Committee on Nuclear Safety to three questions posed by the Atomic Energy Control Board concerning the need for Emergency Core Cooling Systems (ECCS) in CANDU nuclear power plants, the effectiveness requirement for such systems, and the extent to which experimental evidence should be available to demonstrate compliance with effectiveness standards

  2. Core concept of fast power reactor with zero sodium void reactivity

    International Nuclear Information System (INIS)

    Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

    1991-01-01

    The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

  3. INCA: method of analyzing in-core detector data in power reactors

    International Nuclear Information System (INIS)

    Ober, T.G.; Terney, W.B.; Marks, G.H.

    1975-04-01

    A method (INCA) is described by which signals from fixed in-core detectors are related to estimates of the three dimensional power distribution in an operating reactor core and to the maximum linear heat rate in the core. A description of the large library of data accompanying the method is provided. A detailed examination of the analytical verifications performed using the method is presented, and a summary of the uncertainty associated with the method is given. The uncertainty assigned to the maximum linear heat rate inferred by the method from operating reactor data is found to be 5.8 percent at a 95/95 confidence level. (U.S.)

  4. A Practical Core Loss Model for Filter Inductors of Power Electronic Converters

    DEFF Research Database (Denmark)

    Matsumori, Hiroaki; Shimizu, Toshihisa; Wang, Xiongfei

    2018-01-01

    This paper proposes a core loss model for filter inductors of power electronic converters. The model allows a computationally efficient analysis on the core loss of the inductor under the square voltage excitation and the premagnetization condition. First, the core loss of the filter inductor under...... buck chopper excitation is evaluated with the proposed model and compared with the conventional methods. The comparison shows that the proposed method results in a better core loss prediction under the premagnetized condition than that of conventional alternatives. Then, the core loss of the filter...... inductor with the pulsewidth modulated inverter excitation is evaluated, which shows that the proposed model not only accurately predicts the core loss but also identifies the hysteresis loss part. These results demonstrate that the approach can further be used for the development of magnetic materials...

  5. Neutronic design of the RSG-GAS silicide core

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, T.M.; Kuntoro, I.; Hastowo, H. [Center for Development of Research Reactor Technology National Nuclear Energy Agency BATAN, PUSPIPTEK Serpong Tangerang, 15310 (Indonesia)

    2002-07-01

    The objective of core conversion program of the RSG-GAS multipurpose reactor is to convert the fuel from oxide, U{sub 3}O{sub 8}-Al to silicide, U{sub 3}Si{sub 2}-Al. The aim of the program is to gain longer operation cycle by having, which is technically possible for silicide fuel, a higher density. Upon constraints of the existing reactor system and utilization, an optimal fuel density in amount of 3.55 g U/cc was found. This paper describes the neutronic parameter design of the silicide equilibrium core and the design of its transition cores as well. From reactivity control point of view, a modification of control rod system is also discussed. All calculations are carried out by means of diffusion codes, Batan-EQUIL-2D, Batan-2DIFF and -3DIFF. The silicide core shows that longer operation cycle of 32 full power days can be achieved without decreasing the safety criteria and utilization capabilities. (author)

  6. High-power picosecond pulse delivery through hollow core photonic band gap fibers

    DEFF Research Database (Denmark)

    Michieletto, Mattia; Johansen, Mette Marie; Lyngsø, Jens Kristian

    2016-01-01

    We demonstrated robust and bend insensitive fiber delivery of high power laser with diffraction limited beam quality for two different kinds of hollow core band gap fibers. The light source for this experiment consists of ytterbium-doped double clad fiber aeroGAIN-ROD-PM85 in a high power amplifier...

  7. Examining real-time time-dependent density functional theory nonequilibrium simulations for the calculation of electronic stopping power

    Science.gov (United States)

    Yost, Dillon C.; Yao, Yi; Kanai, Yosuke

    2017-09-01

    In ion irradiation processes, electronic stopping power describes the energy transfer rate from the irradiating ion to the target material's electrons. Due to the scarcity and significant uncertainties in experimental electronic stopping power data for materials beyond simple solids, there has been growing interest in the use of first-principles theory for calculating electronic stopping power. In recent years, advances in high-performance computing have opened the door to fully first-principles nonequilibrium simulations based on real-time time-dependent density functional theory (RT-TDDFT). While it has been demonstrated that the RT-TDDFT approach is capable of predicting electronic stopping power for a wide range of condensed matter systems, there has yet to be an exhaustive examination of the physical and numerical approximations involved and their effects on the calculated stopping power. We discuss the results of such a study for crystalline silicon with protons as irradiating ions. We examine the influences of key approximations in RT-TDDFT nonequilibrium simulations on the calculated electronic stopping power, including approximations related to basis sets, finite size effects, exchange-correlation approximation, pseudopotentials, and more. Finally, we propose a simple and efficient correction scheme to account for the contribution from core-electron excitations to the stopping power, as it was found to be significant for large proton velocities.

  8. VG-400 atomic power and technological installation. Possible core design

    International Nuclear Information System (INIS)

    Komarov, E.V.; Laptev, F.V.; Lyubivyj, A.G.; Mitenkov, F.M.; Samojlov, O.B.; Sukhachevskij, Yu.B.

    1979-01-01

    The main characteristics, basic circuit and configuration of equipment of the VG-400 atomic power and technological installation are considered. This installation is intended for supplying with highly-potential heat of thermal electrochemical hydrogen production and for power generation in the steam-turbine cycle. The main installation characteristics: HTGR reactor heat power 1100 MW, electric power 300 MW, helium coolant pressure 50 atm, output temperature 950 deg C, steam pressure in the second contour 175 atm, temperature 535 deg C, core diameter and height 6.4 m and 4 m, respectively, number of spherical fuel elements 8.5x10 5 . The installation can ensure hydrogen production of 10 5 Nxm 3 /h. For the VG-400 reactor block the integral arrangement of the first circuit equipment in the reinforced concrete is chosen. Two versions of the reactor core with prismatic and spherical fuel elements are compared. It is shown that taking into account great potentialities of the spherical zone in a case of further temperature increase and its positive qualities with respect to construction and processing of fuel elements and graphite blocks, the utilization of simplier units and mechanisms in the overloading system and in the process of profiling of energy distribution the choice of the spherical configuration for the VG-400 pilot plant installation seems to be valid

  9. Feedback control of plasma density and heating power for steady state operation in LHD

    Energy Technology Data Exchange (ETDEWEB)

    Kamio, Shuji, E-mail: kamio@nifs.ac.jp; Kasahara, Hiroshi; Seki, Tetsuo; Saito, Kenji; Seki, Ryosuke; Nomura, Goro; Mutoh, Takashi

    2015-12-15

    Highlights: • We upgraded a control system for steady state operation in LHD. • This system contains gas fueling system and ICRF power control system. • Automatic power boost system is also attached for stable operation. • As a result, we achieved the long pulse up to 48 min in the electron density of more than 1 × 10{sup 19} m{sup −3}. - Abstract: For steady state operation, the feedback control of plasma density and heating power system was developed in the Large Helical Device (LHD). In order to achieve a record of the long pulse discharge, stable plasma density and heating power are needed. This system contains the radio frequency (RF) heating power control, interlocks, gas fueling, automatic RF phase control, ion cyclotron range of frequency (ICRF) antenna position control, and graphical user interface (GUI). Using the density control system, the electron density was controlled to the target density and using the RF heating power control system, the RF power injection could be stable. As a result of using this system, we achieved the long pulse up to 48 min in the electron density of more than 1 × 10{sup 19} m{sup −3}. Further, the ICRF hardware experienced no critical accidents during the 17th LHD experiment campaign in 2013.

  10. Limit cycle analysis of nuclear coupled density wave oscillations

    International Nuclear Information System (INIS)

    Ward, M.E.

    1985-01-01

    An investigation of limit cycle behavior for the nuclear-coupled density wave oscillation (NCDWO) in a boiling water reactor (BWR) was performed. A simplified nonlinear model of BWR core behavior was developed using a two-region flow channel representation, coupled with a form of the point-kinetics equation. This model has been used to investigate the behavior of large amplitude NCDWO's through conventional time-integration solutions and through application of a direct relaxation-oscillation limit cycle solution in phase space. The numerical solutions demonstrate the potential for severe global power and flow oscillations in a BWR core at off-normal conditions, such as might occur during Anticipated Transients without Scram. Because of the many simplifying assumptions used, it is felt that the results should not be interpreted as an absolute prediction of core behavior, but as an indication of the potential for large oscillations and a demonstration of the corresponding limit cycle mechanisms. The oscillations in channel density drive the core power variations, and are reinforced by heat flux variations due to the changing fuel temperature. A global temperature increase occurs as energy is accumulated in the fuel, and limits the magnitude of the oscillations because as the average channel density decreases, the amplitude and duration of positive void reactivity at a given oscillation amplitude is lessened

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

    International Nuclear Information System (INIS)

    Kascak, A.F.

    1972-01-01

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

  12. Results of an analysis of in-core measurements during the first core cycle of the Greifswald nuclear power plant, unit 3

    International Nuclear Information System (INIS)

    Gehre, G.

    1982-01-01

    First results of an analysis of flux and temperature values obtained from the in-core system in the third unit of the Greifswald nuclear power plant during the first core cycle are presented. The analysis has been performed with the aid of the computer code INCA. Possibilities and limits of this code are shown. (author)

  13. Solid-Core, Gas-Cooled Reactor for Space and Surface Power

    International Nuclear Information System (INIS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

  14. Evaluation of plasma-wave spectral density from cross-power spectra

    International Nuclear Information System (INIS)

    Ilic, D.B.; Harker, K.J.

    1975-01-01

    The plasma-wave spectral density is evaluated by performing a spatial Fourier transform on experimental cross-power spectra of ion acoustic waves. The cross-power spectra are recorded on analog magnetic tape, converted to digital form, transferred to digital magnetic tape, and Fourier transformed on a digital computer. The important effects of sampling, finite data strings, and data smoothing on the end results are discussed and illustrated. The results indicate the usefulness of the spectral density method for the study of nonlinear wave phenomena. (auth)

  15. Flexible asymmetric supercapacitors with high energy and high power density in aqueous electrolytes

    Science.gov (United States)

    Cheng, Yingwen; Zhang, Hongbo; Lu, Songtao; Varanasi, Chakrapani V.; Liu, Jie

    2013-01-01

    Supercapacitors with both high energy and high power densities are critical for many practical applications. In this paper, we discuss the design and demonstrate the fabrication of flexible asymmetric supercapacitors based on nanocomposite electrodes of MnO2, activated carbon, carbon nanotubes and graphene. The combined unique properties of each of these components enable highly flexible and mechanically strong films that can serve as electrodes directly without using any current collectors or binders. Using these flexible electrodes and a roll-up approach, asymmetric supercapacitors with 2 V working voltage were successfully fabricated. The fabricated device showed excellent rate capability, with 78% of the original capacitance retained when the scan rate was increased from 2 mV s-1 to 500 mV s-1. Owing to the unique composite structure, these supercapacitors were able to deliver high energy density (24 W h kg-1) under high power density (7.8 kW kg-1) conditions. These features could enable supercapacitor based energy storage systems to be very attractive for a variety of critical applications, such as the power sources in hybrid electric vehicles and the back-up powers for wind and solar energy, where both high energy density and high power density are required.Supercapacitors with both high energy and high power densities are critical for many practical applications. In this paper, we discuss the design and demonstrate the fabrication of flexible asymmetric supercapacitors based on nanocomposite electrodes of MnO2, activated carbon, carbon nanotubes and graphene. The combined unique properties of each of these components enable highly flexible and mechanically strong films that can serve as electrodes directly without using any current collectors or binders. Using these flexible electrodes and a roll-up approach, asymmetric supercapacitors with 2 V working voltage were successfully fabricated. The fabricated device showed excellent rate capability, with 78% of

  16. Stable formation of ultrahigh power-density 248 nm channels in Xe cluster targets

    Energy Technology Data Exchange (ETDEWEB)

    Borisov, Alex B.; Racz, Ervin; Khan, Shahab F.; Poopalasingam, Sankar; McCorkindale, John C.; Boguta, John; Longworth, James W.; Rhodes, Charles K. [Laboratory for X-ray Microimaging and Bioinformatics, Department of Physics, University of Illinois at Chicago, Chicago, IL 60607-7059 (United States); KFKI Research Institute for Particle and Nuclear Physics, EURATOM Association, P.O. Box 49, 1525 Budapest (Hungary)

    2012-07-11

    The optimization of relativistic and ponderomotive self-channeling of ultra-powerful 248 nm laser pulses launched in underdense plasmas with an appropriate longitudinal gradient in the electron density profile located at the initial stage of the self-channeling leads to (1) stable channel formation and (2) highly efficient power compression producing power densities in the 10{sup 19}-10{sup 20} W/cm{sup 3} range. The comparison of theoretical studies with experimental results involving the correlation of (a) Thomson images of the electron density with (b) x-ray images of the channel morphology demonstrates that more than 90% of the incident 248 nm power can be trapped in stable channels and that this stable propagation can be extended to power levels significantly exceeding the critical power of the self-channeling process.

  17. Analysis of in-core coolant temperatures of FFTF instrumented fuels tests at full power

    International Nuclear Information System (INIS)

    Hoth, C.W.

    1981-01-01

    Two full size highly instrumented fuel assemblies were inserted into the core of the Fast Flux Test Facility in December of 1979. The major objectives of these instrumented tests are to provide verification of the FFTF core conditions and to characterize temperature patterns within FFTF driver fuel assemblies. A review is presented of the results obtained during the power ascents and during irradiation at a constant reactor power of 400 MWt. The results obtained from these instrumented tests verify the conservative nature of the design methods used to establish core conditions in FFTF. The success of these tests also demonstrates the ability to design, fabricate, install and irradiate complex, instrumented fuel tests in FFTF using commercially procured components

  18. Thermodynamic, energy efficiency, and power density analysis of reverse electrodialysis power generation with natural salinity gradients

    NARCIS (Netherlands)

    Yip, N.Y.; Vermaas, D.A.; Nijmeijer, K.; Elimelech, M.

    2014-01-01

    Reverse electrodialysis (RED) can harness the Gibbs free energy of mixing when fresh river water flows into the sea for sustainable power generation. In this study, we carry out a thermodynamic and energy efficiency analysis of RED power generation, and assess the membrane power density. First, we

  19. High-power density miniscale power generation and energy harvesting systems

    International Nuclear Information System (INIS)

    Lyshevski, Sergey Edward

    2011-01-01

    This paper reports design, analysis, evaluations and characterization of miniscale self-sustained power generation systems. Our ultimate objective is to guarantee highly-efficient mechanical-to-electrical energy conversion, ensure premier wind- or hydro-energy harvesting capabilities, enable electric machinery and power electronics solutions, stabilize output voltage, etc. By performing the advanced scalable power generation system design, we enable miniscale energy sources and energy harvesting technologies. The proposed systems integrate: (1) turbine which rotates a radial- or axial-topology permanent-magnet synchronous generator at variable angular velocity depending on flow rate, speed and load, and, (2) power electronic module with controllable rectifier, soft-switching converter and energy storage stages. These scalable energy systems can be utilized as miniscale auxiliary and self-sustained power units in various applications, such as, aerospace, automotive, biotechnology, biomedical, and marine. The proposed systems uniquely suit various submersible and harsh environment applications. Due to operation in dynamic rapidly-changing envelopes (variable speed, load changes, etc.), sound solutions are researched, proposed and verified. We focus on enabling system organizations utilizing advanced developments for various components, such as generators, converters, and energy storage. Basic, applied and experimental findings are reported. The prototypes of integrated power generation systems were tested, characterized and evaluated. It is documented that high-power density, high efficiency, robustness and other enabling capabilities are achieved. The results and solutions are scalable from micro (∼100 μW) to medium (∼100 kW) and heavy-duty (sub-megawatt) auxiliary and power systems.

  20. Control of plasma density distribution via wireless power transfer in an inductively coupled plasma

    International Nuclear Information System (INIS)

    Lee, Hee-Jin; Lee, Hyo-Chang; Kim, Young-Cheol; Chung, Chin-Wook

    2013-01-01

    With an enlargement of the wafer size, development of large-area plasma sources and control of plasma density distribution are required. To control the spatial distribution of the plasma density, wireless power transfer is applied to an inductively coupled plasma for the first time. An inner powered antenna and an outer resonant coil connected to a variable capacitor are placed on the top of the chamber. As the self-resonance frequency ω r of the resonant coil is adjusted, the power transfer rate from the inner powered coil to the outer resonant coil is changed and the dramatic evolution of the plasma density profile is measured. As ω r of the outer resonant coil changes from the non-resonant condition (where ω r is not the driving angular frequency ω rf ) to the resonant condition (where ω r = ω rf ), the plasma density profile evolves from a convex shape with maximal plasma density at the radial center into a concave shape with maximal plasma density in the vicinity of the resonant antenna coil. This result shows that the plasma density distribution can be successfully controlled via wireless resonance power transfer. (fast track communication)

  1. Gas core reactor power plants designed for low proliferation potential

    International Nuclear Information System (INIS)

    Lowry, L.L.

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

  2. Computation of the Mutual Inductance between Air-Cored Coils of Wireless Power Transformer

    International Nuclear Information System (INIS)

    Anele, A O; Hamam, Y; Djouani, K; Chassagne, L; Alayli, Y; Linares, J

    2015-01-01

    Wireless power transfer system is a modern technology which allows the transfer of electric power between the air-cored coils of its transformer via high frequency magnetic fields. However, due to its coil separation distance and misalignment, maximum power transfer is not guaranteed. Based on a more efficient and general model available in the literature, rederived mathematical models for evaluating the mutual inductance between circular coils with and without lateral and angular misalignment are presented. Rather than presenting results numerically, the computed results are graphically implemented using MATLAB codes. The results are compared with the published ones and clarification regarding the errors made are presented. In conclusion, this study shows that power transfer efficiency of the system can be improved if a higher frequency alternating current is supplied to the primary coil, the reactive parts of the coils are compensated with capacitors and ferrite cores are added to the coils. (paper)

  3. Study of variations of radiofrequency power density from mobile phone base stations with distance

    International Nuclear Information System (INIS)

    Ayinmode, B. O.; Farai, I. P.

    2013-01-01

    The variations of radiofrequency (RF) radiation power density with distance around some mobile phone base stations (BTSs), in ten randomly selected locations in Ibadan, western Nigeria, were studied. Measurements were made with a calibrated hand-held spectrum analyser. The maximum Global System of Mobile (GSM) communication 1800 signal power density was 323.91 μW m -2 at 250 m radius of a BTS and that of GSM 900 was 1119.00 μW m -2 at 200 m radius of another BTS. The estimated total maximum power density was 2972.00 μW m -2 at 50 m radius of a different BTS. This study shows that the maximum carrier signal power density and the total maximum power density from a BTS may be observed averagely at 200 and 50 m of its radius, respectively. The result of this study demonstrates that exposure of people to RF radiation from phone BTSs in Ibadan city is far less than the recommended limits by International scientific bodies. (authors)

  4. 40 CFR 1042.140 - Maximum engine power, displacement, power density, and maximum in-use engine speed.

    Science.gov (United States)

    2010-07-01

    ... cylinders having an internal diameter of 13.0 cm and a 15.5 cm stroke length, the rounded displacement would... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Maximum engine power, displacement... Maximum engine power, displacement, power density, and maximum in-use engine speed. This section describes...

  5. Design and Performance Test of Axial Halbach Brushless DC Motor with Power Density 1.5 Kw/Kg

    Directory of Open Access Journals (Sweden)

    Kevin Dwi Prasetio

    2017-01-01

    Full Text Available Progress of technology on electric vehicle component sector is one reason the emergence of electric vehicles at the moment. Starting from battery which has a great current density up to the automatic control systems on electric vehicles. But there are still some shortcomings of this electric vehicle components, one of which is the low value of power density of existing electric motor in the market today.On vehicles such as electric cars when Race Car Contest, energy saving problems about power density of the driving motor is very vital. This is because the total weight of the vehicle has a huge influence on the vehicle efficiency is against it. The issue is one of the reasons of the research task. In this final task is done making the design, simulation, and architecture of the Axial Halbach Brushless DC Motor. Use of system configuration on the halbach magnet to avoid the use of iron as a material cantilever rotor. By changing the material of the cantilever rotor with lighter materials such as aluminum or even carbon fibre, the value of power density electric motors can be increased. Then using the litz wire on coil stator to reduce loss-power loss due to the barriers on the coil. Coreless stator on the system and to avoid the phenomenon of cogging at the time due to low rpm style attraction magnet with iron in the core material. While the creation process begins by determining the specifications of the Axial Halbach Brushless DC motors. Then go into the design phase of the mechanical and electrical design. Who then conducted simulations to help determine other parameters such as air gap, slot turn, and magnetic orientation. The process of making a component of stator and rotor after the simulation is completed. After all the components of the rotor and stator on assembly, mounting the hall sensor is carried out to the right to position obtained by reading the signals. After the motor can spin with good motor performance, testing can be done

  6. Design and Fabrication of Air-core Inductors for Power Conversion

    DEFF Research Database (Denmark)

    Lê Thanh, Hoà; Mizushima, Io; Tang, Peter Torben

    supply on chip (PwrSoC) [1]. Examples of PwrSoC applications are power adaptors for LED illumination and the “Internet of Things”. We report an air-core MEMS inductor. Our process is scalable and universal for making inductors with versatile geometries e.g. spiral, solenoid, toroid, and advanced...

  7. In-core Instrument Subcritical Verification (INCISV) - Core Design Verification Method - 358

    International Nuclear Information System (INIS)

    Prible, M.C.; Heibel, M.D.; Conner, S.L.; Sebastiani, P.J.; Kistler, D.P.

    2010-01-01

    According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed in-core self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of confirming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at various rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the ex-core detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities. (authors)

  8. Improvement of the Cubic Spline Function Sets for a Synthesis of the Axial Power Distribution of a Core Protection System

    International Nuclear Information System (INIS)

    Koo, Bon-Seung; Lee, Chung-Chan; Zee, Sung-Quun

    2006-01-01

    Online digital core protection system(SCOPS) for a system-integrated modular reactor is being developed as a part of a plant protection system at KAERI. SCOPS calculates the minimum CHFR and maximum LPD based on several online measured system parameters including 3-level ex-core detector signals. In conventional ABB-CE digital power plants, cubic spline synthesis technique has been used in online calculations of the core axial power distributions using ex-core detector signals once every 1 second in CPC. In CPC, pre-determined cubic spline function sets are used depending on the characteristics of the ex-core detector responses. But this method shows an unnegligible power distribution error for the extremely skewed axial shapes by using restrictive function sets. Therefore, this paper describes the cubic spline method for the synthesis of an axial power distribution and it generates several new cubic spline function sets for the application of the core protection system, especially for the severely distorted power shapes needed reactor type

  9. Characterization of neutron leakage probability, k /SUB eff/ , and critical core surface mass density of small reactor assemblies through the Trombay criticality formula

    International Nuclear Information System (INIS)

    Kumar, A.; Rao, K.S.; Srinivasan, M.

    1983-01-01

    The Trombay criticality formula (TCF) has been derived by incorporating a number of well-known concepts of criticality physics to enable prediction of changes in critical size or k /SUB eff/ following alterations in geometrical and physical parameters of uniformly reflected small reactor assemblies characterized by large neutron leakage from the core. The variant parameters considered are size, shape, density and diluent concentration of the core, and density and thickness of the reflector. The effect of these changes (except core size) manifests, through sigma /SUB c/ the critical surface mass density of the ''corresponding critical core,'' that sigma, the massto-surface-area ratio of the core,'' is essentially a measure of the product /rho/ extended to nonspherical systems and plays a dominant role in the TCF. The functional dependence of k /SUB eff/ on sigma/sigma /SUB c/ , the system size relative to critical, is expressed in the TCF through two alternative representations, namely the modified Wigner rational form and, an exponential form, which is given

  10. Reactor power control device

    International Nuclear Information System (INIS)

    Kobayashi, Akira.

    1980-01-01

    Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

  11. Neutronic design of mixed oxide-silicide cores for the core conversion of rsg-gas reactor

    International Nuclear Information System (INIS)

    Sembiring, Tagor Malem; Tukiran; Pinem surian; Febrianto

    2001-01-01

    The core conversion of rsg-gas reactor from an all-oxide (U 3 O 8 -Al) core, through a series of mixed oxide-silicide core, to an all-silicide (U 3 Si 2 -Al) core for the same meat density of 2.96 g U/cc is in progress. The conversion is first step of the step-wise conversion and will be followed by the second step that is the core conversion from low meat density of silicide core, through a series of mixed lower-higher density of silicide core, to an all-higher meat density of 3.55 g/cc core. Therefore, the objectives of this work is to design the mixed cores on the neutronic performance to achieve safety a first full-silicide core for the reactor with the low uranium meat density of 2.96gU/cc. The neutronic design of the mixed cores was performed by means of Batan-EQUIL-2D and Batan-3DIFF computer codes for 2 and 3 dimension diffusion calculation, respectively. The result shows that all mixed oxide-silicide cores will be feasible to achieve safety a fist full-silicide core. The core performs the same neutronic core parameters as those of the equilibrium silicide core. Therefore, the reactor availability and utilization during the core conversion is not changed

  12. Interactions between canopy cover density and regeneration cores of older saplings in Scots pine (Pinus sylvestris L.) stands

    Energy Technology Data Exchange (ETDEWEB)

    Marchi, M.; Nocentini, S.; Ducci, F.

    2016-07-01

    Aim of the study: This paper provides an analysis of growth and survival of twenty–year–old Scots pine saplings in relation to canopy cover density (CCD) gradients, from dense (D–CCD), sparse (S–CCD), and gap (G–CCD) situations. Area of study: Aladag (Bolu) in northern Turkey. Material and methods: Sparse canopy cover density (S–CCD), dense canopy cover density (D–CCD) and gap canopy (G–CCD) were chosen within ten different strip sample plots (10 × 50 m) with sapling regeneration cores. Those regeneration cores were divided into two portions (individuals at the edge and middle of the regeneration cores) and from each portion three individuals was were obtained from a sample. The growth relationships of individual saplings were calculated with stem analyses. Honowski Light Factor (HLF) (ratio of Terminal sprout length (T) to Lateral sprout length (L)) was used to present growth potential measure of seedlings. Main results: The largest sapling regeneration cores were found in the G–CCD followed by S–CCD, and finally D–CCD, all tested for significance with Kruskal–Wallis Test. Compared with saplings in the middle of regeneration cores (crop saplings), those at the edge were always reduced in terms of mean height. Significant difference was only found between the ‘Main Crop’ and the ‘Edge 1’ of the regeneration cores for G–CCD suggesting that sapling regeneration cores are more typical under G–CCD conditions. HLF ratios were greater than 1 with high growth potentials for both CCD gradients (G–CCD and S–CCD) and there were no significant variations between G–CCD and S–CCD for main crop and edges. The thinning after 12–14 years increased sapling growth. However, under D–CCD, growth had virtually ceased. Research highlights: Naturally occurring Scots pine saplings are suppressed by a dense canopy. However, they are tolerant of shade to the extent that they can survive over relatively long time–periods (10–12 years) and

  13. Nuclear piston engine and pulsed gaseous core reactor power systems

    International Nuclear Information System (INIS)

    Dugan, E.T.

    1976-01-01

    The investigated nuclear piston engines consist of a pulsed, gaseous core reactor enclosed by a moderating-reflecting cylinder and piston assembly and operate on a thermodynamic cycle similar to the internal combustion engine. The primary working fluid is a mixture of uranium hexafluoride, UF 6 , and helium, He, gases. Highly enriched UF 6 gas is the reactor fuel. The helium is added to enhance the thermodynamic and heat transfer characteristics of the primary working fluid and also to provide a neutron flux flattening effect in the cylindrical core. Two and four-stroke engines have been studied in which a neutron source is the counterpart of the sparkplug in the internal combustion engine. The piston motions which have been investigated include pure simple harmonic, simple harmonic with dwell periods, and simple harmonic in combination with non-simple harmonic motion. The results of the conducted investigations indicate good performance potential for the nuclear piston engine with overall efficiencies of as high as 50 percent for nuclear piston engine power generating units of from 10 to 50 Mw(e) capacity. Larger plants can be conceptually designed by increasing the number of pistons, with the mechanical complexity and physical size as the probable limiting factors. The primary uses for such power systems would be for small mobile and fixed ground-based power generation (especially for peaking units for electrical utilities) and also for nautical propulsion and ship power

  14. Influence of FA pin power minimisation on neutron-physical core characteristics

    International Nuclear Information System (INIS)

    Mikolas, P.

    2005-01-01

    The influence of optimisation of radial fuel enrichment profilation in fuel assembly of Gd-2 type on lowering its pin power non-uniformity and subsequent lowering of F dH in WWER-440 cores are described in this paper (Author)

  15. High-power density miniscale power generation and energy harvesting systems

    Energy Technology Data Exchange (ETDEWEB)

    Lyshevski, Sergey Edward [Department of Electrical and Microelectronics Engineering, Rochester Institute of Technology, Rochester, NY 14623-5603 (United States)

    2011-01-15

    This paper reports design, analysis, evaluations and characterization of miniscale self-sustained power generation systems. Our ultimate objective is to guarantee highly-efficient mechanical-to-electrical energy conversion, ensure premier wind- or hydro-energy harvesting capabilities, enable electric machinery and power electronics solutions, stabilize output voltage, etc. By performing the advanced scalable power generation system design, we enable miniscale energy sources and energy harvesting technologies. The proposed systems integrate: (1) turbine which rotates a radial- or axial-topology permanent-magnet synchronous generator at variable angular velocity depending on flow rate, speed and load, and, (2) power electronic module with controllable rectifier, soft-switching converter and energy storage stages. These scalable energy systems can be utilized as miniscale auxiliary and self-sustained power units in various applications, such as, aerospace, automotive, biotechnology, biomedical, and marine. The proposed systems uniquely suit various submersible and harsh environment applications. Due to operation in dynamic rapidly-changing envelopes (variable speed, load changes, etc.), sound solutions are researched, proposed and verified. We focus on enabling system organizations utilizing advanced developments for various components, such as generators, converters, and energy storage. Basic, applied and experimental findings are reported. The prototypes of integrated power generation systems were tested, characterized and evaluated. It is documented that high-power density, high efficiency, robustness and other enabling capabilities are achieved. The results and solutions are scalable from micro ({proportional_to}100 {mu}W) to medium ({proportional_to}100 kW) and heavy-duty (sub-megawatt) auxiliary and power systems. (author)

  16. Physicians' Professionally Responsible Power: A Core Concept of Clinical Ethics.

    Science.gov (United States)

    McCullough, Laurence B

    2016-02-01

    The gathering of power unto themselves by physicians, a process supported by evidence-based practice, clinical guidelines, licensure, organizational culture, and other social factors, makes the ethics of power--the legitimation of physicians' power--a core concept of clinical ethics. In the absence of legitimation, the physician's power over patients becomes problematic, even predatory. As has occurred in previous issues of the Journal, the papers in the 2016 clinical ethics issue bear on the professionally responsible deployment of power by physicians. This introduction explores themes of physicians' power in papers from an international group of authors who address autonomy and trust, the virtues of perinatal hospice, conjoined twins in ethics and law, addiction and autonomy in clinical research on addicting substances, euthanasia of patients with dementia in Belgium, and a pragmatic approach to clinical futility. © The Author 2015. Published by Oxford University Press, on behalf of the Journal of Medicine and Philosophy Inc. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  17. Automated in-core image generation from video to aid visual inspection of nuclear power plant cores

    Energy Technology Data Exchange (ETDEWEB)

    Murray, Paul, E-mail: paul.murray@strath.ac.uk [Department of Electronic and Electrical Engineering, University of Strathclyde, Technology and Innovation Centre, 99 George Street, Glasgow, G1 1RD (United Kingdom); West, Graeme; Marshall, Stephen; McArthur, Stephen [Dept. Electronic and Electrical Engineering, University of Strathclyde, Royal College Building, 204 George Street, Glasgow G1 1XW (United Kingdom)

    2016-04-15

    Highlights: • A method is presented which improves visual inspection of reactor cores. • Significant time savings are made to activities on the critical outage path. • New information is extracted from existing data sources without additional overhead. • Examples from industrial case studies across the UK fleet of AGR stations. - Abstract: Inspection and monitoring of key components of nuclear power plant reactors is an essential activity for understanding the current health of the power plant and ensuring that they continue to remain safe to operate. As the power plants age, and the components degrade from their initial start-of-life conditions, the requirement for more and more detailed inspection and monitoring information increases. Deployment of new monitoring and inspection equipment on existing operational plant is complex and expensive, as the effect of introducing new sensing and imaging equipment to the existing operational functions needs to be fully understood. Where existing sources of data can be leveraged, the need for new equipment development and installation can be offset by the development of advanced data processing techniques. This paper introduces a novel technique for creating full 360° panoramic images of the inside surface of fuel channels from in-core inspection footage. Through the development of this technique, a number of technical challenges associated with the constraints of using existing equipment have been addressed. These include: the inability to calibrate the camera specifically for image stitching; dealing with additional data not relevant to the panorama construction; dealing with noisy images; and generalising the approach to work with two different capture devices deployed at seven different Advanced Gas Cooled Reactor nuclear power plants. The resulting data processing system is currently under formal assessment with a view to replacing the existing manual assembly of in-core defect montages. Deployment of the

  18. Experimental study of particle transport and density fluctuation in LHD

    International Nuclear Information System (INIS)

    Tanaka, K.; Morita, S.; Sanin, A.; Michael, C.; Kawahata, K.; Yamada, H.; Miyazawa, J.; Tokuzawa, T.; Akiyama, T.; Goto, M.; Ida, K.; Yoshinuma, M.; Narihara, K.; Yamada, I.; Yokoyama, M.; Masuzaki, S.; Morisaki, T.; Sakamoto, R.; Funaba, H.; Komori, A.; Vyacheslavov, L.N.; Murakami, S.; Wakasa, A.

    2005-01-01

    A variety of electron density (n e ) profiles have been observed in Large Helical Device (LHD). The density profiles change dramatically with heating power and toroidal magnetic field (B t ) under the same line averaged density. The particle transport coefficients, i.e., diffusion coefficient (D) and convection velocity (V) are experimentally obtained from density modulation experiments in the standard configuration. The values of D and V are estimated separately at the core and edge. The diffusion coefficients are strong function of electron temperature (T e ) and are proportional to T e 1.7±0.9 in core and T e 1.1±0.14 in edge. And edge diffusion coefficients are proportional to B t -2.08 . It is found that the scaling of D in edge is close to gyro-Bohm-like in nature. The existence of non-zero V is observed. It is observed that the electron temperature (T e ) gradient can drive particle convection. This is particularly clear in the core region. The convection velocity in the core region reverses direction from inward to outward as the T e gradient increases. In the edge, the convection is inward directed in the most of the case of the present data set. And it shows modest tendency, whose value is proportional to T e gradient keeping inward direction. However, the toroidal magnetic field also significantly affects value and direction of V. The spectrum of density fluctuation changes at different heating power suggesting that it has an influence on particle transport. The peak wavenumber is around 0.1 times the inversed ion Larmor radius, as is expected from gyro-Bohm diffusion. The peaks of fluctuation intensity are localized at the plasma edge, where density gradient becomes negative and diffusion contributes most to the particle flux. These results suggest a qualitative correlation of fluctuations with particle diffusion. (author)

  19. Power Production Analysis of the OE Buoy WEC for the CORES Project

    DEFF Research Database (Denmark)

    Lavelle, John; Kofoed, Jens Peter

    This report describes the analysis performed on the OE Buoy for the CORES project by the wave energy group at Aalborg University, Denmark. OE Buoy is a type of Oscillating Water Column (OWC) wave energy converter as part of the CORES project. This type of device is one of the most developed...... to extract energy from the ocean (1). Typically, a Wells turbine is used for the Power Take Off (PTO) for OWCs. The Wells turbine has the advantage that it is self-rectifying – with the ability to operate with either direction of airflow, which changes during each cycle of the wave. This type of turbine...... which a total of 39 hours of power production data was collected. A data acquisition system was used to sample the sensors on board and the generator shaft power time-series data was used in the analysis here. A wave-rider buoy, located at the site of OE Buoy and operated by the Marine Institute Ireland...

  20. Toward Low-Cost, High-Energy Density, and High-Power Density Lithium-Ion Batteries

    Science.gov (United States)

    Li, Jianlin; Du, Zhijia; Ruther, Rose E.; AN, Seong Jin; David, Lamuel Abraham; Hays, Kevin; Wood, Marissa; Phillip, Nathan D.; Sheng, Yangping; Mao, Chengyu; Kalnaus, Sergiy; Daniel, Claus; Wood, David L.

    2017-09-01

    Reducing cost and increasing energy density are two barriers for widespread application of lithium-ion batteries in electric vehicles. Although the cost of electric vehicle batteries has been reduced by 70% from 2008 to 2015, the current battery pack cost (268/kWh in 2015) is still >2 times what the USABC targets (125/kWh). Even though many advancements in cell chemistry have been realized since the lithium-ion battery was first commercialized in 1991, few major breakthroughs have occurred in the past decade. Therefore, future cost reduction will rely on cell manufacturing and broader market acceptance. This article discusses three major aspects for cost reduction: (1) quality control to minimize scrap rate in cell manufacturing; (2) novel electrode processing and engineering to reduce processing cost and increase energy density and throughputs; and (3) material development and optimization for lithium-ion batteries with high-energy density. Insights on increasing energy and power densities of lithium-ion batteries are also addressed.

  1. Start-up fuel and power flattening of sodium-cooled candle core

    International Nuclear Information System (INIS)

    Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

    2013-01-01

    The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing

  2. An Electrochemical Capacitor with Applicable Energy Density of 7.4 Wh/kg at Average Power Density of 3000 W/kg.

    Science.gov (United States)

    Zhai, Teng; Lu, Xihong; Wang, Hanyu; Wang, Gongming; Mathis, Tyler; Liu, Tianyu; Li, Cheng; Tong, Yexiang; Li, Yat

    2015-05-13

    Electrochemical capacitors represent a new class of charge storage devices that can simultaneously achieve high energy density and high power density. Previous reports have been primarily focused on the development of high performance capacitor electrodes. Although these electrodes have achieved excellent specific capacitance based on per unit mass of active materials, the gravimetric energy densities calculated based on the weight of entire capacitor device were fairly small. This is mainly due to the large mass ratio between current collector and active material. We aimed to address this issue by a 2-fold approach of minimizing the mass of current collector and increasing the electrode performance. Here we report an electrochemical capacitor using 3D graphene hollow structure as current collector, vanadium sulfide and manganese oxide as anode and cathode materials, respectively. 3D graphene hollow structure provides a lightweight and highly conductive scaffold for deposition of pseudocapacitive materials. The device achieves an excellent active material ratio of 24%. Significantly, it delivers a remarkable energy density of 7.4 Wh/kg (based on the weight of entire device) at the average power density of 3000 W/kg. This is the highest gravimetric energy density reported for asymmetric electrochemical capacitors at such a high power density.

  3. Feasibility study of ultra-long life fast reactor core concept - 028

    International Nuclear Information System (INIS)

    Kim, T.K.; Taiwo, T.A.

    2010-01-01

    An ultra-long life core concept is proposed targeting capital and operational cost reductions and ultra-high discharge burnup in a fast reactor system. The core concept is achieved by de-rating the power density and adopting annular core geometry to maintain criticality for more than 40 years without refueling. The ultra-long life core has a specific power of ∼10 MW/t and an average driver fuel discharge burnup of ∼300 GWd/t. It is assumed such ultra-high burnup fuel can be developed within an advanced fuel cycle program. Several benefits are expected from the ultra-long life core concept such as capital and operational cost reductions, low proliferation risk, and effectively holding LWR spent fuel without disposal until technologies for a closed nuclear fuel cycle are developed and deployed. As future work, safety analysis, development of the advanced core cooling methods, and comparative cost analysis are expected. (authors)

  4. Role of Density Gradient Driven Trapped Electron Modes in the H-Mode Inner Core with Electron Heating

    Science.gov (United States)

    Ernst, D.

    2015-11-01

    We present new experiments and nonlinear gyrokinetic simulations showing that density gradient driven TEM (DGTEM) turbulence dominates the inner core of H-Mode plasmas during strong electron heating. Thus α-heating may degrade inner core confinement in H-Mode plasmas with moderate density peaking. These DIII-D low torque quiescent H-mode experiments were designed to study DGTEM turbulence. Gyrokinetic simulations using GYRO (and GENE) closely match not only particle, energy, and momentum fluxes, but also density fluctuation spectra, with and without ECH. Adding 3.4 MW ECH doubles Te /Ti from 0.5 to 1.0, which halves the linear TEM critical density gradient, locally flattening the density profile. Density fluctuations from Doppler backscattering (DBS) intensify near ρ = 0.3 during ECH, displaying a band of coherent fluctuations with adjacent toroidal mode numbers. GYRO closely reproduces the DBS spectrum and its change in shape and intensity with ECH, identifying these as coherent TEMs. Prior to ECH, parallel flow shear lowers the effective nonlinear DGTEM critical density gradient 50%, but is negligible during ECH, when transport displays extreme stiffness in the density gradient. GS2 predictions show the DGTEM can be suppressed, to avoid degradation with electron heating, by broadening the current density profile to attain q0 >qmin > 1 . A related experiment in the same regime varied the electron temperature gradient in the outer half-radius (ρ ~ 0 . 65) using ECH, revealing spatially coherent 2D mode structures in the Te fluctuations measured by ECE imaging. Fourier analysis with modulated ECH finds a threshold in Te profile stiffness. Supported by the US DOE under DE-FC02-08ER54966 and DE-FC02-04ER54698.

  5. Power-level regulation and simulation of nonlinear pressurized water reactor core with xenon oscillation using H-infinity loop shaping control

    Directory of Open Access Journals (Sweden)

    Li Gang

    2016-01-01

    Full Text Available This investigation is to solve the power-level control issue of a nonlinear pressurized water reactor core with xenon oscillations. A nonlinear pressurized water reactor core is modeled using the lumped parameter method, and a linear model of the core is then obtained through the small perturbation linearization way. The H∞loop shapingcontrolis utilized to design a robust controller of the linearized core model.The calculated H∞loop shaping controller is applied to the nonlinear core model. The nonlinear core model and the H∞ loop shaping controller build the nonlinear core power-level H∞loop shaping control system.Finally, the nonlinear core power-level H∞loop shaping control system is simulatedconsidering two typical load processes that are a step load maneuver and a ramp load maneuver, and simulation results show that the nonlinear control system is effective.

  6. Development of Core Monitoring System for Nuclear Power Plants (I)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.H.; Kim, Y.B.; Park, M.G; Lee, E.K.; Shin, H.C.; Lee, D.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    1.Object and Necessity of the Study -The main objectives of this study are (1)conversion of APOLLO version BEACON system to HP-UX version core monitoring system, (2)provision of the technical bases to enhance the in-house capability of developing more advanced core monitoring system. 2.Results of the Study - In this study, the revolutionary core monitoring technologies such as; nodal analysis and isotope depletion calculation method, advanced schemes for power distribution control, and treatment of nuclear databank were established. The verification and validation work has been successfully performed by comparing the results with those of the design code and measurement data. The advanced graphic user interface and plant interface method have been implemented to ensure the future upgrade capability. The Unix shell scripts and system dependent software are also improved to support administrative functions of the system. (author). 14 refs., 112 figs., 52 tabs.

  7. Self-powered in-core neutron detector assembly with uniform perturbation characteristics

    International Nuclear Information System (INIS)

    1981-01-01

    An in-core neutron detector assembly consisting of a number of longitudinally extending self-powered detectors is described. The uniform mechanical structures and materials are placed symmetrically at each active detector portion thus ensuring that local perturbation factors are uniform. (U.K.)

  8. Neutronic Core Performance of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Villarino, Eduardo; Hergenreder, Daniel; Matzkin, S

    2000-01-01

    The actual design state of core of CAREM-25 reactor is presented.It is shown that the core design complains with the safety and operation established requirements.It is analyzed the behavior of the reactor safety and control systems (single failure of the fast shut down system, single failure of the shut down system, single failure of the second shut down system, reactivity worth of the adjust and control system in normal operation and hot shut down, reactivity worth of the adjust and control system and the scheme of movement of the control rod during the operation cycle).It is shown the burnup profile of fuel elements with the proposed scheme of refueling and the burnup and power density distribution at different moments of the operation cycle.The power peaking factor of the equilibrium core is 2.56, the minimum DNBR is 1.90 and its average is 2.09 during the operation cycle

  9. NON-EQUILIBRIUM CHEMISTRY OF DYNAMICALLY EVOLVING PRESTELLAR CORES. II. IONIZATION AND MAGNETIC FIELD

    International Nuclear Information System (INIS)

    Tassis, Konstantinos; Willacy, Karen; Yorke, Harold W.; Turner, Neal J.

    2012-01-01

    We study the effect that non-equilibrium chemistry in dynamical models of collapsing molecular cloud cores has on measurements of the magnetic field in these cores, the degree of ionization, and the mean molecular weight of ions. We find that OH and CN, usually used in Zeeman observations of the line-of-sight magnetic field, have an abundance that decreases toward the center of the core much faster than the density increases. As a result, Zeeman observations tend to sample the outer layers of the core and consistently underestimate the core magnetic field. The degree of ionization follows a complicated dependence on the number density at central densities up to 10 5 cm –3 for magnetic models and 10 6 cm –3 in non-magnetic models. At higher central densities, the scaling approaches a power law with a slope of –0.6 and a normalization which depends on the cosmic-ray ionization rate ζ and the temperature T as (ζT) 1/2 . The mean molecular weight of ions is systematically lower than the usually assumed value of 20-30, and, at high densities, approaches a value of 3 due to the asymptotic dominance of the H + 3 ion. This significantly lower value implies that ambipolar diffusion operates faster.

  10. Active core profile and transport modification by application of ion Bernstein wave power in the Princeton Beta Experiment-Modification

    Science.gov (United States)

    LeBlanc, B.; Batha, S.; Bell, R.; Bernabei, S.; Blush, L.; de la Luna, E.; Doerner, R.; Dunlap, J.; England, A.; Garcia, I.; Ignat, D.; Isler, R.; Jones, S.; Kaita, R.; Kaye, S.; Kugel, H.; Levinton, F.; Luckhardt, S.; Mutoh, T.; Okabayashi, M.; Ono, M.; Paoletti, F.; Paul, S.; Petravich, G.; Post-Zwicker, A.; Sauthoff, N.; Schmitz, L.; Sesnic, S.; Takahashi, H.; Talvard, M.; Tighe, W.; Tynan, G.; von Goeler, S.; Woskov, P.; Zolfaghari, A.

    1995-03-01

    Application of Ion Bernstein Wave Heating (IBWH) into the Princeton Beta Experiment-Modification (PBX-M) [Phys. Fluids B 2, 1271 (1990)] tokamak stabilizes sawtooth oscillations and generates peaked density profiles. A transport barrier, spatially correlated with the IBWH power deposition profile, is observed in the core of IBWH-assisted neutral beam injection (NBI) discharges. A precursor to the fully developed barrier is seen in the soft x-ray data during edge localized mode (ELM) activity. Sustained IBWH operation is conducive to a regime where the barrier supports large ∇ne, ∇Te, ∇νφ, and ∇Ti, delimiting the confinement zone. This regime is reminiscent of the H(high) mode, but with a confinement zone moved inward. The core region has better than H-mode confinement while the peripheral region is L(low)-mode-like. The peaked profile enhances NBI core deposition and increases nuclear reactivity. An increase in central Ti results from χi reduction (compared to the H mode) and better beam penetration. Bootstrap current fractions of up to 0.32-0.35 locally and 0.28 overall were obtained when an additional NBI burst is applied to this plasma.

  11. Active core profile and transport modification by application of ion Bernstein wave power in the Princeton Beta Experiment-Modification

    International Nuclear Information System (INIS)

    LeBlanc, B.; Batha, S.; Bell, R.; Bernabei, S.; Blush, L.; de la Luna, E.; Doerner, R.; Dunlap, J.; England, A.; Garcia, I.; Ignat, D.; Isler, R.; Jones, S.; Kaita, R.; Kaye, S.; Kugel, H.; Levinton, F.; Luckhardt, S.; Mutoh, T.; Okabayashi, M.; Ono, M.; Paoletti, F.; Paul, S.; Petravich, G.; Post-Zwicker, A.; Sauthoff, N.; Schmitz, L.; Sesnic, S.; Takahashi, H.; Talvard, M.; Tighe, W.; Tynan, G.; von Goeler, S.; Woskov, P.; Zolfaghari, A.

    1995-01-01

    Application of Ion Bernstein Wave Heating (IBWH) into the Princeton Beta Experiment-Modification (PBX-M) [Phys. Fluids B 2, 1271 (1990)] tokamak stabilizes sawtooth oscillations and generates peaked density profiles. A transport barrier, spatially correlated with the IBWH power deposition profile, is observed in the core of IBWH-assisted neutral beam injection (NBI) discharges. A precursor to the fully developed barrier is seen in the soft x-ray data during edge localized mode (ELM) activity. Sustained IBWH operation is conducive to a regime where the barrier supports large ∇n e , ∇T e , ∇ν φ , and ∇T i , delimiting the confinement zone. This regime is reminiscent of the H(high) mode, but with a confinement zone moved inward. The core region has better than H-mode confinement while the peripheral region is L(low)-mode-like. The peaked profile enhances NBI core deposition and increases nuclear reactivity. An increase in central T i results from χ i reduction (compared to the H mode) and better beam penetration. Bootstrap current fractions of up to 0.32--0.35 locally and 0.28 overall were obtained when an additional NBI burst is applied to this plasma

  12. 47 CFR 25.208 - Power flux density limits.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 2 2010-10-01 2010-10-01 false Power flux density limits. 25.208 Section 25.208 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) COMMON CARRIER SERVICES SATELLITE... emissions from all co-frequency space stations of a single non-geostationary-satellite orbit (NGSO) system...

  13. SOLAR NEUTRINO PHYSICS OSCILLATIONS: SENSITIVITY TO THE ELECTRONIC DENSITY IN THE SUN'S CORE

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Ilidio [Centro Multidisciplinar de Astrofisica, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Turck-Chieze, Sylvaine, E-mail: ilidio.lopes@ist.utl.pt, E-mail: ilopes@uevora.pt, E-mail: sylvaine.turck-chieze@cea.fr [CEA/IRFU/Service d' Astrophysique, CE Saclay, F-91191 Gif sur Yvette (France)

    2013-03-01

    Solar neutrinos coming from different nuclear reactions are now detected with high statistics. Consequently, an accurate spectroscopic analysis of the neutrino fluxes arriving on Earth's detectors becomes available, in the context of neutrino oscillations. In this work, we explore the possibility of using this information to infer the radial profile of the electronic density in the solar core. So, we discuss the constraints on the Sun's density and chemical composition that can be determined from solar neutrino observations. This approach constitutes an independent and alternative diagnostic to the helioseismic investigations already done. The direct inversion method, which we propose to obtain the radial solar electronic density profile, is almost independent of the solar model.

  14. Toroidal HTS transformer with cold magnetic core - analysis with FEM software

    International Nuclear Information System (INIS)

    Grzesik, B; Stepien, M; Jez, R

    2010-01-01

    The aim of this paper is to present a thorough characterization of the toroidal HTS transformer by means of FEM analysis. The analysis was a 2D/3D harmonic electromagnetic and thermal analysis. The toroidal transformer operated in LN2 by being immersed together with the magnetic core in it, for which its power losses were acceptable. Two extreme variants of windings were analysed. The first one called parallel and the second called perpendicular. Three variants of the magnetic core were considered. In the first one the core was put outside of the windings, in the second the core was inside of the windings and in the third variant the core was outside as well as inside of the windings. The windings were made of HTS tape BiSCCO-2223/Ag while the magnetic core was made of the nanocrystalline material Finemet. The two windings, with a 1:1 turn-to-turn ratio, were uniformly distributed along the whole torus circumference. The output power, efficiency and power density are in the results of the analysis. The temperature distribution was also calculated. In summary, the performance of the transformer is better than those currently known.

  15. High Energy Density Sciences with High Power Lasers at SACLA

    Science.gov (United States)

    Kodama, Ryosuke

    2013-10-01

    One of the interesting topics on high energy density sciences with high power lasers is creation of extremely high pressures in material. The pressures of more than 0.1 TPa are the energy density corresponding to the chemical bonding energy, resulting in expectation of dramatic changes in the chemical reactions. At pressures of more than TPa, most of material would be melted on the shock Hugoniot curve. However, if the temperature is less than 1eV or lower than a melting point at pressures of more than TPa, novel solid states of matter must be created through a pressured phase transition. One of the interesting materials must be carbon. At pressures of more than TPa, the diamond structure changes to BC and cubic at more than 3TPa. To create such novel states of matter, several kinds of isentropic-like compression techniques are being developed with high power lasers. To explore the ``Tera-Pascal Science,'' now we have a new tool which is an x-ray free electron laser as well as high power lasers. The XFEL will clear the details of the HED states and also efficiently create hot dense matter. We have started a new project on high energy density sciences using an XFEL (SACLA) in Japan, which is a HERMES (High Energy density Revolution of Matter in Extreme States) project.

  16. A novel optimization method, Gravitational Search Algorithm (GSA), for PWR core optimization

    International Nuclear Information System (INIS)

    Mahmoudi, S.M.; Aghaie, M.; Bahonar, M.; Poursalehi, N.

    2016-01-01

    Highlights: • The Gravitational Search Algorithm (GSA) is introduced. • The advantage of GSA is verified in Shekel’s Foxholes. • Reload optimizing in WWER-1000 and WWER-440 cases are performed. • Maximizing K eff , minimizing PPFs and flattening power density is considered. - Abstract: In-core fuel management optimization (ICFMO) is one of the most challenging concepts of nuclear engineering. In recent decades several meta-heuristic algorithms or computational intelligence methods have been expanded to optimize reactor core loading pattern. This paper presents a new method of using Gravitational Search Algorithm (GSA) for in-core fuel management optimization. The GSA is constructed based on the law of gravity and the notion of mass interactions. It uses the theory of Newtonian physics and searcher agents are the collection of masses. In this work, at the first step, GSA method is compared with other meta-heuristic algorithms on Shekel’s Foxholes problem. In the second step for finding the best core, the GSA algorithm has been performed for three PWR test cases including WWER-1000 and WWER-440 reactors. In these cases, Multi objective optimizations with the following goals are considered, increment of multiplication factor (K eff ), decrement of power peaking factor (PPF) and power density flattening. It is notable that for neutronic calculation, PARCS (Purdue Advanced Reactor Core Simulator) code is used. The results demonstrate that GSA algorithm have promising performance and could be proposed for other optimization problems of nuclear engineering field.

  17. Density fluctuation effects on collective neutrino oscillations in O-Ne-Mg core-collapse supernovae

    International Nuclear Information System (INIS)

    Cherry, John F.; Fuller, George M.; Wu Mengru; Qian Yongzhong; Carlson, J.; Duan Huaiyu

    2011-01-01

    We investigate the effect of matter density fluctuations on supernova collective neutrino flavor oscillations. In particular, we use full multiangle, three-flavor, self-consistent simulations of the evolution of the neutrino flavor field in the envelope of an O-Ne-Mg core-collapse supernova at shock breakout (neutronization neutrino burst) to study the effect of the matter density ''bump'' left by the He-burning shell. We find a seemingly counterintuitive increase in the overall ν e survival probability created by this matter density feature. We discuss this behavior in terms of the interplay between the matter density profile and neutrino collective effects. While our results give new insights into this interplay, they also suggest an immediate consequence for supernova neutrino burst detection: it will be difficult to use a burst signal to extract information on fossil burning shells or other fluctuations of this scale in the matter density profile. Consistent with previous studies, our results also show that the interplay of neutrino self-coupling and matter fluctuation could cause a significant increase in the ν e survival probability at very low energy.

  18. Power density effect on feasibility of water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    Sidik, Permana; Takaki, Naoyuki; Sekimoto, Hiroshi

    2008-01-01

    Breeding is made possible by the high value of neutron regeneration ratio η for 233 U in thermal energy region. The reactor is fueled by 233 U-Th oxide and it has used the light water as moderator. Some characteristics such as spectrum, η value, criticality, breeding performance and number density are evaluated. Several power densities are evaluated in order to analyze its effect to the breeding performance. The η value of fissile 233 U obtains higher value than 2 which may satisfy the breeding capability especially for thermal reactor for all investigated MFR. The increasing enrichment and decreasing conversion ratio are more significant for MFR 233 U enrichment. Number density of 233 Pa decreases significantly with decreasing power density which leads the reactor has better breeding performance because lower capture rate of 233 Pa. (author)

  19. ELECTRON-CAPTURE AND β-DECAY RATES FOR sd-SHELL NUCLEI IN STELLAR ENVIRONMENTS RELEVANT TO HIGH-DENSITY O–NE–MG CORES

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Toshio [Department of Physics and Graduate School of Integrated Basic Sciences, College of Humanities and Sciences, Nihon University Sakurajosui 3-25-40, Setagaya-ku, Tokyo 156-8550 (Japan); Toki, Hiroshi [Research Center for Nuclear Physics (RCNP), Osaka University, Ibaraki, Osaka 567-0047 (Japan); Nomoto, Ken’ichi, E-mail: suzuki@phys.chs.nihon-u.ac.jp [Kavli Institute for the Physics and Mathematics of the Universe (WPI), The University of Tokyo, Kashiwa, Chiba 277-8583 (Japan)

    2016-02-01

    Electron-capture and β-decay rates for nuclear pairs in the sd-shell are evaluated at high densities and high temperatures relevant to the final evolution of electron-degenerate O–Ne–Mg cores of stars with initial masses of 8–10 M{sub ⊙}. Electron capture induces a rapid contraction of the electron-degenerate O–Ne–Mg core. The outcome of rapid contraction depends on the evolutionary changes in the central density and temperature, which are determined by the competing processes of contraction, cooling, and heating. The fate of the stars is determined by these competitions, whether they end up with electron-capture supernovae or Fe core-collapse supernovae. Since the competing processes are induced by electron capture and β-decay, the accurate weak rates are crucially important. The rates are obtained for pairs with A = 20, 23, 24, 25, and 27 by shell-model calculations in the sd-shell with the USDB Hamiltonian. Effects of Coulomb corrections on the rates are evaluated. The rates for pairs with A = 23 and 25 are important for nuclear Urca processes that determine the cooling rate of the O–Ne–Mg core, while those for pairs with A = 20 and 24 are important for the core contraction and heat generation rates in the core. We provide these nuclear rates at stellar environments in tables with fine enough meshes at various densities and temperatures for studies of astrophysical processes sensitive to the rates. In particular, the accurate rate tables are crucially important for the final fates of not only O–Ne–Mg cores but also a wider range of stars, such as C–O cores of lower-mass stars.

  20. ELECTRON-CAPTURE AND β-DECAY RATES FOR sd-SHELL NUCLEI IN STELLAR ENVIRONMENTS RELEVANT TO HIGH-DENSITY O–NE–MG CORES

    International Nuclear Information System (INIS)

    Suzuki, Toshio; Toki, Hiroshi; Nomoto, Ken’ichi

    2016-01-01

    Electron-capture and β-decay rates for nuclear pairs in the sd-shell are evaluated at high densities and high temperatures relevant to the final evolution of electron-degenerate O–Ne–Mg cores of stars with initial masses of 8–10 M ⊙ . Electron capture induces a rapid contraction of the electron-degenerate O–Ne–Mg core. The outcome of rapid contraction depends on the evolutionary changes in the central density and temperature, which are determined by the competing processes of contraction, cooling, and heating. The fate of the stars is determined by these competitions, whether they end up with electron-capture supernovae or Fe core-collapse supernovae. Since the competing processes are induced by electron capture and β-decay, the accurate weak rates are crucially important. The rates are obtained for pairs with A = 20, 23, 24, 25, and 27 by shell-model calculations in the sd-shell with the USDB Hamiltonian. Effects of Coulomb corrections on the rates are evaluated. The rates for pairs with A = 23 and 25 are important for nuclear Urca processes that determine the cooling rate of the O–Ne–Mg core, while those for pairs with A = 20 and 24 are important for the core contraction and heat generation rates in the core. We provide these nuclear rates at stellar environments in tables with fine enough meshes at various densities and temperatures for studies of astrophysical processes sensitive to the rates. In particular, the accurate rate tables are crucially important for the final fates of not only O–Ne–Mg cores but also a wider range of stars, such as C–O cores of lower-mass stars

  1. Graphite core design in UK reactors

    International Nuclear Information System (INIS)

    Davies, M.W.

    1996-01-01

    The cores in the first power producing Magnox reactors in the UK were designed with only a limited amount of information available regarding the anisotropic dimensional change behaviour of Pile Grade graphite. As more information was gained it was necessary to make modifications to the design, some minor, some major. As the cores being built became larger, and with the switch to the Advanced Gas-cooled Reactor (AGR) with its much higher power density, additional problems had to be overcome such as increased dimensional change and radiolytic oxidation by the carbon dioxide coolant. For the AGRs a more isotropic graphite was required, with a lower initial open pore volume and higher strength. Gilsocarbon graphite was developed and was selected for all the AGRs built in the UK. Methane bearing coolants are used to limit radiolytic oxidation. (author). 5 figs

  2. Physicians’ Professionally Responsible Power: A Core Concept of Clinical Ethics

    Science.gov (United States)

    McCullough, Laurence B.

    2016-01-01

    The gathering of power unto themselves by physicians, a process supported by evidence-based practice, clinical guidelines, licensure, organizational culture, and other social factors, makes the ethics of power—the legitimation of physicians’ power—a core concept of clinical ethics. In the absence of legitimation, the physician’s power over patients becomes problematic, even predatory. As has occurred in previous issues of the Journal, the papers in the 2016 clinical ethics issue bear on the professionally responsible deployment of power by physicians. This introduction explores themes of physicians’ power in papers from an international group of authors who address autonomy and trust, the virtues of perinatal hospice, conjoined twins in ethics and law, addiction and autonomy in clinical research on addicting substances, euthanasia of patients with dementia in Belgium, and a pragmatic approach to clinical futility. PMID:26671961

  3. On-line extraction of the variance caused by burn-up in in-core three-dimensional power distribution

    International Nuclear Information System (INIS)

    Wang Yaqi; Luo Zhengpei; Li Fu; Liu Wenfeng

    2001-01-01

    In most of PWRs, the ex-core ion-chambers are the sole real-time sensors to respond to in-core power and its axial offset. However, the calibration coefficient of the ion-chambers depends on the (3D) power distribution and varies with the burn-up. People expect to know the variance in distribution caused by burn-up directly from the signals of ion-chambers. This expectation is not realized as yet, because an ion-chamber almost only responds to its nearest fuel assemblies. The authors then developed a two-step method for burn-up characteristic extraction: the harmonics synthesis method and harmonics' burn-up grouping. Using the extracted burn-up characteristics, the relationship between the readings of the ex-core ion-chambers and the in-core 3D power distribution is set up. Through the simulation on the heating reactor, the method of burn-up characteristic extraction is verified under engineering conditions. It is possible to on-line extract the variance caused by burn-up in 3D power distribution

  4. A DETERMINATION OF THE FLUX DENSITY IN CORE OF DISTRIBUTION TRANSFORMERS, WHAT BUILT WITH THE COMMON USING OF GRAIN AND NON GRAIN ORIENTED MAGNETIC STEELS

    Directory of Open Access Journals (Sweden)

    I.V. Pentegov

    2015-12-01

    Full Text Available Purpose. The development of calculation method to determinate the flux densities in different parts of the magnetic cores of distribution transformers, what built from different types magnetic steel (mixed core. Methodology. The method is based on the scientific positions of Theoretical Electrical Engineering – the theory of the electromagnetic field in nonlinear mediums to determine the distribution of magnetic flux in mixed core of transformer, what are using different types of steel what have the different magnetic properties. Results. The developed method gives possible to make calculation of the flux density and influence of skin effect in different parts of the magnetic cores of distribution transformer, where are used mix of grain oriented (GO and non grain oriented (NGO steels. Was determinate the general basic conditions for the calculation of flux density in the laminations from grain and non grain oriented steels of the magnetic core: the strength of magnetic field for the laminations of particular part of mixed core is the same; the sum of the magnetic fluxes in GO and NGO steels in particular part of mixed core is equal with the designed magnetic flux in this part of mixed core. Discover, the magnetic flux in mixed core of the transformer has specific distribution between magnetic steels. The flux density is higher in laminations from GO steel and smaller in laminations from the NGO steel. That is happened because for magnetic flux is easier pass through laminations from GO steel, what has better magnetic conductance than laminations from NGO steel. Originality. The common using of different types of magnetic steels in cores for distribution transformers gives possibility to make design of transformer with low level of no load losses, high efficiency and with optimal cost. Practical value. The determination of the flux density in different parts of magnetic core with GO and NGO steels gives possibility make accurate calculation of

  5. Nuclear data uncertainties for local power densities in the Martin-Hoogenboom benchmark

    International Nuclear Information System (INIS)

    Van der Marck, S.C.; Rochman, D.A.

    2013-01-01

    The recently developed method of fast Total Monte Carlo to propagate nuclear data uncertainties was applied to the Martin-Hoogenboom benchmark. This Martin- Hoogenboom benchmark prescribes that one calculates local pin powers (of light water cooled reactor) with a statistical uncertainty lower than 1% everywhere. Here we report, for the first time, an estimate of the nuclear data uncertainties for these local pin powers. For each of the more than 6 million local power tallies, the uncertainty due to nuclear data uncertainties was calculated, based on random variation of data for 235 U, 238 U, 239 Pu and H in H 2 O thermal scattering. In the center of the core region, the nuclear data uncertainty is 0.9%. Towards the edges of the core, this uncertainty increases to roughly 3%. The nuclear data uncertainties have been shown to be larger than the statistical uncertainties that the benchmark prescribes

  6. Robust feedback-linearization control for axial power distribution in pressurized water reactors during load-following operation

    OpenAIRE

    M. Zaidabadi nejad; G.R. Ansarifar

    2018-01-01

    Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the impor...

  7. Frontiers in pulse-power-based high energy density plasma physics and its applications

    International Nuclear Information System (INIS)

    Horioka, Kazuhiko

    2008-03-01

    The papers in this volume of report were presented at the Symposium on Frontiers in Pulse-power-based High Energy Density Physics' held by National Institute for Fusion Science. The topics include the present status of high energy density plasma researches, extreme ultraviolet sources, intense radiation sources, high power ion beams, and R and D of related pulse power technologies. The 13 of the presented papers are indexed individually. (J.P.N.)

  8. Performance analysis and comparison of an Atkinson cycle coupled to variable temperature heat reservoirs under maximum power and maximum power density conditions

    International Nuclear Information System (INIS)

    Wang, P.-Y.; Hou, S.-S.

    2005-01-01

    In this paper, performance analysis and comparison based on the maximum power and maximum power density conditions have been conducted for an Atkinson cycle coupled to variable temperature heat reservoirs. The Atkinson cycle is internally reversible but externally irreversible, since there is external irreversibility of heat transfer during the processes of constant volume heat addition and constant pressure heat rejection. This study is based purely on classical thermodynamic analysis methodology. It should be especially emphasized that all the results and conclusions are based on classical thermodynamics. The power density, defined as the ratio of power output to maximum specific volume in the cycle, is taken as the optimization objective because it considers the effects of engine size as related to investment cost. The results show that an engine design based on maximum power density with constant effectiveness of the hot and cold side heat exchangers or constant inlet temperature ratio of the heat reservoirs will have smaller size but higher efficiency, compression ratio, expansion ratio and maximum temperature than one based on maximum power. From the view points of engine size and thermal efficiency, an engine design based on maximum power density is better than one based on maximum power conditions. However, due to the higher compression ratio and maximum temperature in the cycle, an engine design based on maximum power density conditions requires tougher materials for engine construction than one based on maximum power conditions

  9. Self-powered in-core neutron detector assembly with uniform perturbation characteristics

    International Nuclear Information System (INIS)

    Todt, W.H.; Playfoot, K.C.

    1979-01-01

    Disclosed is a self-powered in-core neutron detector assembly in which a plurality of longitudinally extending self-powered detectors have neutron responsive active portions spaced along a longitudinal path. A low neutron absorptive extension extends from the active portions of the spaced active portions of the detectors in symmetrical longitudinal relationship with the spaced active detector portions of each succeeding detector. The detector extension terminates with the detector assembly to provide a uniform perturbation characteristic over the entire assembly length

  10. Space structures, power, and power conditioning; Proceedings of the Meeting, Los Angeles, CA, Jan. 11-13, 1988

    International Nuclear Information System (INIS)

    Askew, R.F.

    1988-01-01

    Various papers on space structures, power, and power conditioning are presented. Among the topics discussed are: heterogeneous gas core reaction for space nuclear power, pulsed gas core reactor for burst power, fundamental considerations of gas core reactor systems, oscillating thermionic conversion for high-density space power, thermoelectromagnetic pumps for space nuclear power systems, lightweight electrochemical converter for space power applications, ballistic acceleration by superheated hydrogen, laser-induced current switching in gaseous discharge, electron-beam-controlled semiconductor switches, laser-controlled semiconductor closing and opening switch. Also addressed are: semiconductor-metal eutectic composites for high-power switching, optical probes for the characterization of surface breakdown, 40 kV/20 kA pseudospark switch for laser applications, insulation direction for high-power space systems, state space simulation of spacecraft power systems, structural vibration of space power station systems, minimum-time control of large space structures, novel fusion reaction for space power and propulsion, repetition rate system evaluations, cryogenic silicon photoconductive switches for high-power lasers, multilevel diamondlike carbon capacitor structure, surface breakdown of prestressed insulators, C-Mo and C-Zr alloys for space power systems, magnetic insulation for the space environment

  11. Xenon-induced axial power oscillations in the 400 MW PBMR

    International Nuclear Information System (INIS)

    Strydom, Gerhard

    2008-01-01

    The redistribution of the spatial xenon concentration in the 400 MW Pebble Bed Modular Reactor (PBMR) core has a non-linear, time-dependent feedback effect on the spatial power density during several types of operational transient events. Due to the inherent weak coupling that exists between the iodine and xenon formation and destruction rates, as well as the complicating effect of spatial variance in the thermal flux field, reactor cores have been analyzed for a number of decades for the occurrence and severity of xenon-induced axial power oscillations. Of specific importance is the degree of oscillation damping exhibited by the core during transients, which involves axial variations in the local power density. In this paper the TINTE reactor dynamics code is used to assess the stability of the current 400 MW PBMR core design with regard to axial xenon oscillations. The focus is mainly on the determination of the inherent xenon and power oscillation damping properties by utilizing a set of hypothetical control rod insertion transients at various power levels. The oscillation damping properties of two 100%-50%-100% load-follow transients, one of which includes the de-stabilizing axial effects of moving control rods, are also discussed in some detail. The study shows that, although first axial mode oscillations do occur in the 400 MW PBMR core, the inherent damping of these oscillations is high, and that none of the investigated load-follow transients resulted in diverging oscillations. It is also shown that the PBMR core exhibits no radial oscillation components for these xenon-induced axial power oscillations

  12. Neutronics simulations on hypothetical power excursion and possible core melt scenarios in CANDU6

    International Nuclear Information System (INIS)

    Kim, Yonghee

    2015-01-01

    LOCA (Loss of coolant accident) is an outstanding safety issue in the CANDU reactor system since the coolant void reactivity is strongly positive. To deal with the LOCA, the CANDU systems are equipped with specially designed quickly-acting secondary shutdown system. Nevertheless, the so-called design-extended conditions are requested to be taken into account in the safety analysis for nuclear reactor systems after the Fukushima accident. As a DEC scenario, the worst accident situation in a CANDU reactor system is a unprotected LOCA, which is supposed to lead to a power excursion and possibly a core melt-down. In this work, the hypothetical unprotected LOCA scenario is simulated in view of the power excursion and fuel temperature changes by using a simplified point-kinetics (PK) model accounting for the fuel temperature change. In the PK model, the core reactivity is assumed to be affected by a large break LOCA and the fuel temperature is simulated to account for the Doppler effect. In addition, unlike the conventional PK simulation, we have also considered the Xe-I model to evaluate the impact of Xe during the LOCA. Also, we tried to simulate the fuel and core melt-down scenario in terms of the reactivity through a series of neutronics calculations for hypothetical core conditions. In case of a power excursion and possible fuel melt-down situation, the reactor system behavior is very uncertain. In this work, we tried to understand the impacts of fuel melt and relocation within the pressure vessel on the core reactivity and failure of pressure and calandria tubes. (author)

  13. SU-F-T-136: Breath Hold Lung Phantom Study in Using CT Density Versus Relative Stopping Power Ratio for Proton Pencil Beam Scanning System

    Energy Technology Data Exchange (ETDEWEB)

    Syh, J; Wu, H; Rosen, L [Willis-Knighton Medical Center, Shreveport, LA (United States)

    2016-06-15

    Purpose: To evaluate mass density effects of CT conversion table and its variation in current treatment planning system of spot scanning proton beam using an IROC proton lung phantom for this study. Methods: A proton lung phantom study was acquired to Imaging and Radiation Oncology Core Houston (IROC) Quality Assurance Center. Inside the lung phantom, GAF Chromic films and couples of thermal luminescent dosimeter (TLD) capsules embedded in specified PTV and adjacent structures to monitor delivered dosage and 3D dose distribution profiles. Various material such as cork (Lung), blue water (heart), Techron HPV (ribs) and organic material of balsa wood and cork as dosimetry inserts within phantom of solid water (soft tissue). Relative stopping power (RLSP) values were provided. Our treatment planning system (TPS) doesn’t require SP instead relative density was converted relative to water. However lung phantom was irradiated by planning with density override and the results were compared with IROC measurements. The second attempt was conducted without density override and compared with IROC’s. Results: The higher passing rate of imaging and measurement results of the lung phantom irradiation met the criteria by IROC without density override. The film at coronal plane was found to be shift due to inclined cylinder insertion. The converted CT density worked as expected to correlate relative stopping power. Conclusion: The proton lung phantom provided by IROC is a useful tool to qualify our commissioned proton pencil beam delivery with TPS within reliable confidence. The relative mass stopping power ratios of materials were converted from the relative physical density relative to water and the results were satisfied.

  14. Spectral Effects on Fast Wave Core Heating and Current Drive

    International Nuclear Information System (INIS)

    Phillips, C.K.; Bell, R.E.; Berry, L.A.; Bonoli, P.T.; Harvey, R.W.; Hosea, J.C.; Jaeger, E.F.; LeBlanc, B.P.; Ryan, P.M.; Taylor, G.; Valeo, E.J.; Wilson, J.R.; Wright, J.C.; Yuh, H. and the NSTX Team

    2009-01-01

    Recent results obtained with high harmonic fast wave (HHFW) heating and current drive (CD) on NSTX strongly support the hypothesis that the onset of perpendicular fast wave propagation right at or very near the launcher is a primary cause for a reduction in core heating efficiency at long wavelengths that is also observed in ICRF heating experiments in numerous tokamaks. A dramatic increase in core heating efficiency was first achieved in NSTX L-mode helium majority plasmas when the onset for perpendicular wave propagation was moved away from the antenna and nearby vessel structures. Efficient core heating in deuterium majority L mode and H mode discharges, in which the edge density is typically higher than in comparable helium majority plasmas, was then accomplished by reducing the edge density in front of the launcher with lithium conditioning and avoiding operational points prone to instabilities. These results indicate that careful tailoring of the edge density profiles in ITER should be considered to limit rf power losses to the antenna and plasma facing materials. Finally, in plasmas with reduced rf power losses in the edge regions, the first direct measurements of high harmonic fast wave current drive were obtained with the motional Stark effect (MSE) diagnostic. The location and radial dependence of HHFW CD measured by MSE are in reasonable agreement with predictions from both full wave and ray tracing simulations

  15. Stopping power of degenerate electron liquid at metallic densities

    International Nuclear Information System (INIS)

    Tanaka, Shigenori; Ichimaru, Setsuo

    1985-01-01

    We calculate the stopping power of the degenerate electron liquid at metallic densities in the dielectric formalism. The strong Coulomb-coupling effects beyond the random-phase approximation are taken into account through the static and dynamic local-field corrections. It is shown that those strong-coupling and dynamic effects act to enhance the stopping power substantially in the low-velocity regime, leading to an improved agreement with experimental data. (author)

  16. Core-power and decay-time limits for disabled automatic-actuation of LOFT ECCS

    International Nuclear Information System (INIS)

    Hanson, G.H.

    1978-01-01

    The Emergency Core Cooling System (ECCS) for the LOFT reactor may need to be disabled for modifications or repairs of hardware or instrumentation or for component testing during periods when the reactor system is hot and pressurized, or it may be desirable to enable the ECCS to be disabled without the necessity of cooling down and depressurizing the reactor. LTR 113-47 has shown that the LOFT ECCS can be safely bypassed or disabled when the total core power does not exceed 25 kW. A modified policy involves disabling the automatic actuation of the LOFT ECCS, but still retaining the manual activation capability. Disabling of the automatic actuation can be safely utilized, without subjecting the fuel cladding to unacceptable temperatures, when the LOFT power decays to 70 kW; this power level permits a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS

  17. Patients with Rheumatoid Arthritis and Chronic Pain Display Enhanced Alpha Power Density at Rest.

    Science.gov (United States)

    Meneses, Francisco M; Queirós, Fernanda C; Montoya, Pedro; Miranda, José G V; Dubois-Mendes, Selena M; Sá, Katia N; Luz-Santos, Cleber; Baptista, Abrahão F

    2016-01-01

    Patients with chronic pain due to neuropathy or musculoskeletal injury frequently exhibit reduced alpha and increased theta power densities. However, little is known about electrical brain activity and chronic pain in patients with rheumatoid arthritis (RA). For this purpose, we evaluated power densities of spontaneous electroencephalogram (EEG) band frequencies (delta, theta, alpha, and beta) in females with persistent pain due to RA. This was a cross-sectional study of 21 participants with RA and 21 healthy controls (mean age = 47.20; SD = 10.40). EEG was recorded at rest over 5 min with participant's eyes closed. Twenty electrodes were placed over five brain regions (frontal, central, parietal, temporal, and occipital). Significant differences were observed in depression and anxiety with higher scores in RA participants than healthy controls (p = 0.002). Participants with RA exhibited increased average absolute alpha power density in all brain regions when compared to controls [F (1.39) = 6.39, p = 0.016], as well as increased average relative alpha power density [F (1.39) = 5.82, p = 0.021] in all regions, except the frontal region, controlling for depression/anxiety. Absolute theta power density also increased in the frontal, central, and parietal regions for participants with RA when compared to controls [F (1, 39) = 4.51, p = 0.040], controlling for depression/anxiety. Differences were not exhibited on beta and delta absolute and relative power densities. The diffuse increased alpha may suggest a possible neurogenic mechanism for chronic pain in individuals with RA.

  18. Assessment of Microbial Fuel Cell Configurations and Power Densities

    KAUST Repository

    Logan, Bruce E.

    2015-07-30

    Different microbial electrochemical technologies are being developed for a many diverse applications, including wastewater treatment, biofuel production, water desalination, remote power sources, and as biosensors. Current and energy densities will always be limited relative to batteries and chemical fuel cells, but these technologies have other advantages based on the self-sustaining nature of the microorganisms that can donate or accept electrons from an electrode, the range of fuels that can be used, and versatility in the chemicals that can be produced. The high cost of membranes will likely limit applications of microbial electrochemical technologies that might require a membrane. For microbial fuel cells, which do not need a membrane, questions remain on whether larger-scale systems can produce power densities similar to those obtained in laboratory-scale systems. It is shown here that configuration and fuel (pure chemicals in laboratory media versus actual wastewaters) remain the key factors in power production, rather than the scale of the application. Systems must be scaled up through careful consideration of electrode spacing and packing per unit volume of reactor.

  19. Assessment of Microbial Fuel Cell Configurations and Power Densities

    KAUST Repository

    Logan, Bruce E.; Wallack, Maxwell J; Kim, Kyoung-Yeol; He, Weihua; Feng, Yujie; Saikaly, Pascal

    2015-01-01

    Different microbial electrochemical technologies are being developed for a many diverse applications, including wastewater treatment, biofuel production, water desalination, remote power sources, and as biosensors. Current and energy densities will always be limited relative to batteries and chemical fuel cells, but these technologies have other advantages based on the self-sustaining nature of the microorganisms that can donate or accept electrons from an electrode, the range of fuels that can be used, and versatility in the chemicals that can be produced. The high cost of membranes will likely limit applications of microbial electrochemical technologies that might require a membrane. For microbial fuel cells, which do not need a membrane, questions remain on whether larger-scale systems can produce power densities similar to those obtained in laboratory-scale systems. It is shown here that configuration and fuel (pure chemicals in laboratory media versus actual wastewaters) remain the key factors in power production, rather than the scale of the application. Systems must be scaled up through careful consideration of electrode spacing and packing per unit volume of reactor.

  20. The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

    Science.gov (United States)

    Susilo, J.; Suparlina, L.; Deswandri; Sunaryo, G. R.

    2018-02-01

    The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

  1. The simultaneous measurements of core and outer core density fluctuations in L-H transition using CO2 laser collective scattering diagnostic in the EAST superconducting tokamak

    International Nuclear Information System (INIS)

    Cao, G.M.; Li, Y.D.; Zhang, X.D.; Sun, P.J.; Hu, L.Q.; Li, J.G.; Wu, G.J.

    2013-01-01

    The H-mode is the projected basic operation scenario for the ITER tokamak. The turbulence de-correlation by the synergistic effect of zonal flow and equilibrium ExB flow shear is believed to be the reason for L-H transition, however, the detailed physical mechanism has not been identified so far. Tangential multi-channel CO 2 laser collective scattering diagnostic system (mainly k r measurement) was first installed to investigate electron density fluctuations on EAST tokamak. The measurements in a spontaneous dithering L-H transition show that in core plasma (0 < r/a < 0.5) the low-frequency fluctuations strengthen greatly before L-H transition; meanwhile in outer core plasma (0.2 < r/a < 1) the low-frequency fluctuations strengthen slightly. Bispectral analysis reveals that the coupling strength between low- and high-frequency fluctuations in both core and outer core plasma strengthens greatly before the transition, but the latter is greater than the former. The results indicate that the low-frequency fluctuations of the core and outer core plasma play active, but different, roles in the spontaneous L-H transition. (author)

  2. Design comparisons of TRU burner cores with similar sodium void worth

    International Nuclear Information System (INIS)

    Sang Ji, Kim; Young Il, Kim; Young Jin, Kim; Nam Zin, Cho

    2001-01-01

    This study summarizes the neutronic performance and fuel cycle behavior of five geometrically-different transuranic (TRU) burner cores with similar low sodium void reactivity. The conceptual cores encompass core geometries for annular, two-region homogeneous, dual pin type, pan-shaped and H-shaped cores. They have been designed with the same assembly specifications and managed to have similar end-of-cycle sodium void reactivities and beginning-of-cycle peak power densities through the changes in the core size and configuration. The requirement of low sodium void reactivity is shown to lead each design concept to characteristic neutronics performance and fuel cycle behavior. The H-/pan-shaped cores allow the core compaction as well as higher rate of TRU burning. (author)

  3. Power density of piezoelectric transformers improved using a contact heat transfer structure.

    Science.gov (United States)

    Shao, Wei Wei; Chen, Li Juan; Pan, Cheng Liang; Liu, Yong Bin; Feng, Zhi Hua

    2012-01-01

    Based on contact heat transfer, a novel method to increase power density of piezoelectric transformers is proposed. A heat transfer structure is realized by directly attaching a dissipater to the piezoelectric transformer plate. By maintaining the vibration mode of the transformer and limiting additional energy losses from the contact interface, an appropriate design can improve power density of the transformer on a large scale, resulting from effective suppression of its working temperature rise. A prototype device was fabricated from a rectangular piezoelectric transformer, a copper heat transfer sheet, a thermal grease insulation pad, and an aluminum heat radiator. The experimental results show the transformer maintains a maximum power density of 135 W/cm(3) and an efficiency of 90.8% with a temperature rise of less than 10 °C after more than 36 h, without notable changes in performance. © 2012 IEEE

  4. Ab initio translationally invariant nonlocal one-body densities from no-core shell-model theory

    Science.gov (United States)

    Burrows, M.; Elster, Ch.; Popa, G.; Launey, K. D.; Nogga, A.; Maris, P.

    2018-02-01

    Background: It is well known that effective nuclear interactions are in general nonlocal. Thus if nuclear densities obtained from ab initio no-core shell-model (NCSM) calculations are to be used in reaction calculations, translationally invariant nonlocal densities must be available. Purpose: Though it is standard to extract translationally invariant one-body local densities from NCSM calculations to calculate local nuclear observables like radii and transition amplitudes, the corresponding nonlocal one-body densities have not been considered so far. A major reason for this is that the procedure for removing the center-of-mass component from NCSM wave functions up to now has only been developed for local densities. Results: A formulation for removing center-of-mass contributions from nonlocal one-body densities obtained from NCSM and symmetry-adapted NCSM (SA-NCSM) calculations is derived, and applied to the ground state densities of 4He, 6Li, 12C, and 16O. The nonlocality is studied as a function of angular momentum components in momentum as well as coordinate space. Conclusions: We find that the nonlocality for the ground state densities of the nuclei under consideration increases as a function of the angular momentum. The relative magnitude of those contributions decreases with increasing angular momentum. In general, the nonlocal structure of the one-body density matrices we studied is given by the shell structure of the nucleus, and cannot be described with simple functional forms.

  5. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Aoyama, Motoo; Koyama, Jun-ichi; Uchikawa, Sadao; Bessho, Yasunori; Nakajima, Akiyoshi; Maruyama, Hiromi; Ozawa, Michihiro; Nakamura, Mitsuya.

    1990-01-01

    The present invention concerns fuel assemblies charged in a BWR type reactor and the reactor core. The fuel assembly comprises fuel rods containing burnable poisons and fuel rods not containing burnable poisons. Both of the highest and the lowest gadolinia concentrations of the fuel rods containing gadolinia as burnable poisons are present in the lower region of the fuel assembly. This can increase the spectral shift effect without increasing the maximum linear power density. (I.N.)

  6. High-density multicore fiber with heterogeneous core arrangement

    DEFF Research Database (Denmark)

    Amma, Y.; Sasaki, Y.; Takenaga, K.

    2015-01-01

    A 30-core fiber with heterogeneous cores that achieved large spatial multiplicity and low crosstalk of less than −40 dB at 100 km was demonstrated. The correlation lengths were estimated to be more than 1 m.......A 30-core fiber with heterogeneous cores that achieved large spatial multiplicity and low crosstalk of less than −40 dB at 100 km was demonstrated. The correlation lengths were estimated to be more than 1 m....

  7. 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

  8. Active learning for noisy oracle via density power divergence.

    Science.gov (United States)

    Sogawa, Yasuhiro; Ueno, Tsuyoshi; Kawahara, Yoshinobu; Washio, Takashi

    2013-10-01

    The accuracy of active learning is critically influenced by the existence of noisy labels given by a noisy oracle. In this paper, we propose a novel pool-based active learning framework through robust measures based on density power divergence. By minimizing density power divergence, such as β-divergence and γ-divergence, one can estimate the model accurately even under the existence of noisy labels within data. Accordingly, we develop query selecting measures for pool-based active learning using these divergences. In addition, we propose an evaluation scheme for these measures based on asymptotic statistical analyses, which enables us to perform active learning by evaluating an estimation error directly. Experiments with benchmark datasets and real-world image datasets show that our active learning scheme performs better than several baseline methods. Copyright © 2013 Elsevier Ltd. All rights reserved.

  9. Workshop on High Power ICH Antenna Designs for High Density Tokamaks

    Science.gov (United States)

    Aamodt, R. E.

    1990-02-01

    A workshop in high power ICH antenna designs for high density tokamaks was held to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of RF auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made.

  10. Methods to enhance blanket power density in low-power fusion devices

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Miller, L.G.; Bohn, T.S.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Wessol, D.E.; Abdou, M.A.

    1982-06-01

    The overall objective of this task is to investigate the extent to which the power density in the FED breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow testing of advanced reactor blanket modules on INTOR at representative conditions. The tentative approach adopted for this task consists of three parts. First, the requirements for augmented heating of the test module are outlined for different applications of interest. Second, methods are identified which have potential for augmenting the heating power in a test module, and this list of methods is narrowed to those which appear to be most useful. Finally, these methods are examined in more detail to determine the practical benefits of employing each

  11. Operation guide device for nuclear power plants

    International Nuclear Information System (INIS)

    Araki, Tsuneyasu

    1982-01-01

    Purpose: To enable to maintain the soundness of nuclear fuels and each of equipments by compensating the effect of the xenon density on the reactor core thermal power resulted upon load following operation of a nuclear reactor. Constitution: The device comprises an instrumentation system for measuring the status of the nuclear reactor, a reactor core performance calculator for calculating the reactor core performance based on the output from the instrumentation system, a xenon density calculator for calculating the xenon density based on the output from the performance calculator, a memory unit for storing the output from the reactor core performance calculator and the xenon density calculator and for transferring the stored memory to a nuclear reactor status forecasting device and an alternative load pattern searching device for searching, in cooperation with the memory unit, an alternative load pattern which is within an operation restrictive condition and most closed to a demanded load pattern when a monitor for the deviation from the flowrate distribution detects the deviation from the operation restrictive conditions. (Yoshino, Y.)

  12. Flexible Aqueous Li-Ion Battery with High Energy and Power Densities.

    Science.gov (United States)

    Yang, Chongyin; Ji, Xiao; Fan, Xiulin; Gao, Tao; Suo, Liumin; Wang, Fei; Sun, Wei; Chen, Ji; Chen, Long; Han, Fudong; Miao, Ling; Xu, Kang; Gerasopoulos, Konstantinos; Wang, Chunsheng

    2017-11-01

    A flexible and wearable aqueous symmetrical lithium-ion battery is developed using a single LiVPO 4 F material as both cathode and anode in a "water-in-salt" gel polymer electrolyte. The symmetric lithium-ion chemistry exhibits high energy and power density and long cycle life, due to the formation of a robust solid electrolyte interphase consisting of Li 2 CO 3 -LiF, which enables fast Li-ion transport. Energy densities of 141 Wh kg -1 , power densities of 20 600 W kg -1 , and output voltage of 2.4 V can be delivered during >4000 cycles, which is far superior to reported aqueous energy storage devices at the same power level. Moreover, the full cell shows unprecedented tolerance to mechanical stress such as bending and cutting, where it not only does not catastrophically fail, as most nonaqueous cells would, but also maintains cell performance and continues to operate in ambient environment, a unique feature apparently derived from the high stability of the "water-in-salt" gel polymer electrolyte. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  13. The collapse of a molecular cloud core to stellar densities using radiation non-ideal magnetohydrodynamics

    Science.gov (United States)

    Wurster, James; Bate, Matthew R.; Price, Daniel J.

    2018-04-01

    We present results from radiation non-ideal magnetohydrodynamics (MHD) calculations that follow the collapse of rotating, magnetized, molecular cloud cores to stellar densities. These are the first such calculations to include all three non-ideal effects: ambipolar diffusion, Ohmic resistivity, and the Hall effect. We employ an ionization model in which cosmic ray ionization dominates at low temperatures and thermal ionization takes over at high temperatures. We explore the effects of varying the cosmic ray ionization rate from ζcr = 10-10 to 10-16 s-1. Models with ionization rates ≳10-12 s-1 produce results that are indistinguishable from ideal MHD. Decreasing the cosmic ray ionization rate extends the lifetime of the first hydrostatic core up to a factor of 2, but the lifetimes are still substantially shorter than those obtained without magnetic fields. Outflows from the first hydrostatic core phase are launched in all models, but the outflows become broader and slower as the ionization rate is reduced. The outflow morphology following stellar core formation is complex and strongly dependent on the cosmic ray ionization rate. Calculations with high ionization rates quickly produce a fast (≈14 km s-1) bipolar outflow that is distinct from the first core outflow, but with the lowest ionization rate, a slower (≈3-4 km s-1) conical outflow develops gradually and seamlessly merges into the first core outflow.

  14. Frontiers of particle beam and high energy density plasma science using pulse power technology

    International Nuclear Information System (INIS)

    Masugata, Katsumi

    2011-04-01

    The papers presented at the symposium on “Frontiers of Particle Beam and High Energy Density Plasma Science using Pulse Power Technology” held in November 20-21, 2009 at National Institute for Fusion Science are collected. The papers reflect the present status and resent progress in the experiment and theoretical works on high power particle beams and high energy density plasmas produced by pulsed power technology. (author)

  15. Investigation on the Effect of Kenaf Core and Stalk Fiber on the Medium Density Fiber Board Properties Made of Poplar Fibers

    Directory of Open Access Journals (Sweden)

    Fahimeh SH.Alizadeh

    2012-01-01

    Full Text Available In order to optimize the use of material non-forest resources, in this study the possibility of using the kenaf stalk fibers mixed with poplar fibers in producing medium density fiber board was considered. Variable factors such as density at two levels (0.55, 0.75 g/cm3 and the percentage incorporation of fiber (%50 poplar fibers, - %50 kenaf core fiber, %50 poplar fiber, -% 50 kenaf stalk fiber and %100 poplar fibers were considered. Steaming time and temperature (175°C, 10min, press time and temperature (5 min, 175°C, Pressing pressure (30 kg/cm3, fiber cake moisture (%12 and urea-formaldehyde resin with Concentration of %50 of the study factors were fixed. Results show that adding kenaf core fibers to the poplar fibers increases modulus of elasticity and water absorption but thickness swelling reduces. Increased density in board made with kenaf core has caused increase in bending strength, modulus of elasticity and internal bond strength and their water absorption and thickness swelling after 2 and 24 hours were competitive with poplar (MDF. On the other hand Populus fiber– kenaf stalk board mechanical and physical properties were competitive with (MDF board made of %100 poplar fibers. Finally we can say that according to the statistical analysis, the best treatment in this study was using kenaf core fibers, in making poplar (MDF with 0.75 g/cm3 density.

  16. Research on 3D power distribution of PWR reactor core based on RBF neural network

    International Nuclear Information System (INIS)

    Xia Hong; Li Bin; Liu Jianxin

    2014-01-01

    Real-time monitor for 3D power distribution is critical to nuclear safety and high efficiency of NPP's operation as well as the control system optimization. A method was proposed to set up a real-time monitor system for 3D power distribution by using of ex-core neutron detecting system and RBF neural network for improving the instantaneity of the monitoring results and reducing the fitting error of the 3D power distribution. A series of experiments were operated on a 300 MW PWR simulation system. The results demonstrate that the new monitor system works very well under condition of certain burnup range during the fuel cycle and reconstructs the real-time 3D distribution of reactor core power. The accuracy of the model is improved effectively with the help of several methods. (authors)

  17. Neutronic design of the RSG-GAS compact core without CIP

    International Nuclear Information System (INIS)

    Susilo, Jati; Kuntoro, Iman

    2002-01-01

    Improvement of the efficiency of reactor operation can be chivvied by some ways, such as, the uranium density of the fuel, loading pattern and configuration of core elements. The paper deals with determination of optimal configuration of the compact core with out CIP. Calculations were carried out by means of SRAC-PIJ module for cross section generation and SRAC-ASMBURN for core calculations. The optimal compact core obtained, showed that no-CIP compact core increase highest reactivity value about 0,84 % Δk/k and longest time operation about 1,19 time in the safety criteria that is power peaking factor less then 1,4 and margin control element worth less then volume in the first design that -2,2% Δk/k

  18. Neutronic design of the RSG-GAS compact core without CIP

    International Nuclear Information System (INIS)

    Jati-Susilo; Iman-Kuntoro

    2003-01-01

    Improvement of the efficiency of reactor operation can be achieved by some ways, such as, the uranium density of the fuel, loading pattern and configuration of core elements. The paper deals with determination of optimal configuration of the compact core with out CIP. Calculations were carried out by means of SRAC-PIJ module for cross section generation and SRAC-ASMBURN for core calculations. The optimal compact core obtained, showed that no-CIP compact core increase highest reactivity value about 1.06 % Δk/k and longest time operation about 1.19 time in the safety criteria that is power peaking factor less then 1.4 and margin control element worth less then value in the first design that -2.2% Δk/k

  19. Optimization design of toroidal core for magnetic energy harvesting near power line by considering saturation effect

    Science.gov (United States)

    Park, Bumjin; Kim, Dongwook; Park, Jaehyoung; Kim, Kibeom; Koo, Jay; Park, HyunHo; Ahn, Seungyoung

    2018-05-01

    Recently, magnetic energy harvesting technologies have been studied actively for self-sustainable operation of applications around power line. However, magnetic energy harvesting around power lines has the problem of magnetic saturation, which can cause power performance degradation of the harvester. In this paper, optimal design of a toroidal core for magnetic energy harvesters has been proposed with consideration of magnetic saturation near power lines. Using Permeability-H curve and Ampere's circuital law, the optimum dimensional parameters needed to generate induced voltage were analyzed via calculation and simulation. To reflect a real environment, we consider the nonlinear characteristic of the magnetic core material and supply current through a 3-phase distribution panel used in the industry. The effectiveness of the proposed design methodology is verified by experiments in a power distribution panel and takes 60.9 V from power line current of 60 A at 60 Hz.

  20. Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

    International Nuclear Information System (INIS)

    Zheng, Meiyin; Tian, Wenxi; Zhang, Dalin; Qiu, Suizheng; Su, Guanghui

    2015-01-01

    Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

  1. The behaviour of WWER nuclear power stations under abnormal operational conditions

    International Nuclear Information System (INIS)

    Ackermann, G.; Dreger, P.; Prasser, H.M.; Reichenbach, D.

    1985-01-01

    The big power WWER type reactors can show the volumetric oscillations of the power density into the reactor core because of xenon poisoning. The ununiform xenon distribution occurs in the case of nonideal mixing of the coolant into the reactor vessel. This effect usually leads to decreasing of the reactor power. The theoretical and experimental investigations with the goal to evaluate the volumetric distribution of the temperature near the coolant input into the reactor core is discussed in this paper as well as the three dimensional model of the reactor core together with the primary coolant circuit model to calculate volumetric xenon oscillations

  2. The ARIES-RS power core - recent development in Li/V designs

    International Nuclear Information System (INIS)

    Sze Dai-Kai; Billone, M.C.; Hua, T.Q.; Tillack, M.; Najmabadi, F.; Wang Xueren; Malang, S.; El-Guebaly, L.A.; Sviatoslavsky, I.N.; Blanchard, J.P.; Crowell, J.A.; Khater, H.Y.; Mogahed, E.A.; Waganer, L.M.; Lee, D.; Cole, D.

    1998-01-01

    The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirement. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design. This paper summarizes the power core design of the ARIES-RS power plant study. (orig.)

  3. A High Power Density Integrated Charger for Electric Vehicles with Active Ripple Compensation

    Directory of Open Access Journals (Sweden)

    Liwen Pan

    2015-01-01

    Full Text Available This paper suggests a high power density on-board integrated charger with active ripple compensation circuit for electric vehicles. To obtain a high power density and high efficiency, silicon carbide devices are reported to meet the requirement of high-switching-frequency operation. An integrated bidirectional converter is proposed to function as AC/DC battery charger and to transfer energy between battery pack and motor drive of the traction system. In addition, the conventional H-bridge circuit suffers from ripple power pulsating at second-order line frequency, and a scheme of active ripple compensation circuit has been explored to solve this second-order ripple problem, in which a pair of power switches shared traction mode, a ripple energy storage capacitor, and an energy transfer inductor. Simulation results in MATLAB/Simulink validated the eligibility of the proposed topology. The integrated charger can work as a 70 kW motor drive circuit or a converter with an active ripple compensation circuit for 3 kW charging the battery. The impact of the proposed topology and control strategy on the integrated charger power losses, efficiency, power density, and thermal performance has also been analysed and simulated.

  4. Nonlinear dielectric thin films for high-power electric storage with energy density comparable with electrochemical supercapacitors.

    Science.gov (United States)

    Yao, Kui; Chen, Shuting; Rahimabady, Mojtaba; Mirshekarloo, Meysam Sharifzadeh; Yu, Shuhui; Tay, Francis Eng Hock; Sritharan, Thirumany; Lu, Li

    2011-09-01

    Although batteries possess high energy storage density, their output power is limited by the slow movement of charge carriers, and thus capacitors are often required to deliver high power output. Dielectric capacitors have high power density with fast discharge rate, but their energy density is typically much lower than electrochemical supercapacitors. Increasing the energy density of dielectric materials is highly desired to extend their applications in many emerging power system applications. In this paper, we review the mechanisms and major characteristics of electric energy storage with electrochemical supercapacitors and dielectric capacitors. Three types of in-house-produced ferroic nonlinear dielectric thin film materials with high energy density are described, including (Pb(0.97)La(0.02))(Zr(0.90)Sn(0.05)Ti(0.05))O(3) (PLZST) antiferroelectric ceramic thin films, Pb(Zn(1/3)Nb(2/3))O(3-)Pb(Mg(1/3)Nb(2/3))O(3-)PbTiO(3) (PZN-PMN-PT) relaxor ferroelectric ceramic thin films, and poly(vinylidene fluoride) (PVDF)-based polymer blend thin films. The results showed that these thin film materials are promising for electric storage with outstandingly high power density and fairly high energy density, comparable with electrochemical supercapacitors.

  5. High Power Spark Delivery System Using Hollow Core Kagome Lattice Fibers

    Directory of Open Access Journals (Sweden)

    Ciprian Dumitrache

    2014-08-01

    Full Text Available This study examines the use of the recently developed hollow core kagome lattice fibers for delivery of high power laser pulses. Compared to other photonic crystal fibers (PCFs, the hollow core kagome fibers have larger core diameter (~50 µm, which allows for higher energy coupling in the fiber while also maintaining high beam quality at the output (M2 = 1.25. We have conducted a study of the maximum deliverable energy versus laser pulse duration using a Nd:YAG laser at 1064 nm. Pulse energies as high as 30 mJ were transmitted for 30 ns pulse durations. This represents, to our knowledge; the highest laser pulse energy delivered using PCFs. Two fiber damage mechanisms were identified as damage at the fiber input and damage within the bulk of the fiber. Finally, we have demonstrated fiber delivered laser ignition on a single-cylinder gasoline direct injection engine.

  6. Workshop on high power ICH antenna designs for high density tokamaks

    International Nuclear Information System (INIS)

    Aamodt, R.E.

    1990-01-01

    A workshop in high power ICH antenna designs for high density tokamaks was held in Boulder, Colorado on January 31 through February 2, 1990. The purposes of the workshop were to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of rf auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made

  7. On the density within the dark-matter core in our galaxy

    Directory of Open Access Journals (Sweden)

    Ninković S.

    2007-01-01

    Full Text Available Assuming that the disc of our Galaxy, the Milky Way, obeys the classical exponential law, that it is maximal and the Sun is rather far from the maximum of its circular velocity, one finds that, most likely, the galactic corona (subsystem containing the dark matter has a nearly constant density within its core which contains the position of the Sun. The approach applied in the present paper is local, i.e. quantities characterizing the solar neighbourhood are treated. The assumptions and the result could explain why the ratio of the moduli of the Oort constants is expected to exceed the value of 1.0 which corresponds to the locally flat rotation curve of the Milky Way.

  8. Determination of the decay power for a U3O8 designed core using the ORIGEN 2.1 code

    International Nuclear Information System (INIS)

    Castro, Jose; Gallardo, Alberto; Madariaga, Marcelo

    2014-01-01

    After the operation of a nuclear research reactor at a higher power (more than 300 kW), a cooling time is required to remove the residual heat from the core due to the heat produced by the energy emitted by fission products, this fact is common in reactors. There is a short time where the heat output falls to 6 % after the reactor shutdown, the importance of knowing this power is because of the accidental events that this power could cause and affect the fuel after a sudden shutdown in the cooling system of the reactor and there is any other refrigeration system, only that one surrounding the reactor core. This report shows the results of the calculation of the U 3 O 8 core residual power a for the RP-10, using the ORIGEN 2.1 calculation code, verifying the safety of the proposed core within the safety limits accepted for the reactor. (authors).

  9. ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

    Science.gov (United States)

    Whyte, Dennis; ADX Team

    2015-11-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

  10. Assessing different parameters estimation methods of Weibull distribution to compute wind power density

    International Nuclear Information System (INIS)

    Mohammadi, Kasra; Alavi, Omid; Mostafaeipour, Ali; Goudarzi, Navid; Jalilvand, Mahdi

    2016-01-01

    Highlights: • Effectiveness of six numerical methods is evaluated to determine wind power density. • More appropriate method for computing the daily wind power density is estimated. • Four windy stations located in the south part of Alberta, Canada namely is investigated. • The more appropriate parameters estimation method was not identical among all examined stations. - Abstract: In this study, the effectiveness of six numerical methods is evaluated to determine the shape (k) and scale (c) parameters of Weibull distribution function for the purpose of calculating the wind power density. The selected methods are graphical method (GP), empirical method of Justus (EMJ), empirical method of Lysen (EML), energy pattern factor method (EPF), maximum likelihood method (ML) and modified maximum likelihood method (MML). The purpose of this study is to identify the more appropriate method for computing the wind power density in four stations distributed in Alberta province of Canada namely Edmonton City Center Awos, Grande Prairie A, Lethbridge A and Waterton Park Gate. To provide a complete analysis, the evaluations are performed on both daily and monthly scales. The results indicate that the precision of computed wind power density values change when different parameters estimation methods are used to determine the k and c parameters. Four methods of EMJ, EML, EPF and ML present very favorable efficiency while the GP method shows weak ability for all stations. However, it is found that the more effective method is not similar among stations owing to the difference in the wind characteristics.

  11. Fuel and Core Design Verification for Extended Power Up-rate in Ringhals Unit 3

    International Nuclear Information System (INIS)

    Gabrielsson, Petter; Stepniewski, Marek; Almberger, Jan

    2006-01-01

    Vattenfall's Westinghouse 3-loop PWR Ringhals 3 at the western coast of Sweden is scheduled for an extended power up-rate from 2783 to 3160 MWt in 2007, in the frame of the so called GREAT-project. The project will realize an up-rating initially planned and analysed back in 1995, but with a number of significant improvements outlined in this paper. For the licensing of the up-rated power level, a complete revision of the safety analyses, radiological analyses and systems verifications in FSAR is being performed by Westinghouse Electrics Belgium. The work is performed in close cooperation with Vattenfall in the areas of core calculations and input data. For more than a decade, Vattenfall has performed all core design and reload safety evaluations (RSE) for Ringhals, independent of fuel vendors and safety analysts. In GREAT all core parameters in the safety analysis checklist (SAC) used for the safety analyses are determined based upon a set of nine reference loading patterns designed by Vattenfall covering a wide range of fuel and core designs and extreme cycle-to-cycle variations. To facilitate the calculation of SAC parameters Westinghouse has provided a Reload Safety Evaluation Procedure report (RSEP) with detailed specifications for the calculation of all core parameters used in the analyses. The procedure has been automatized by Vattenfall in a set of scripts executing 3D core simulator calculations and extracting the key results. The same tools will be used in Vattenfall's future RSE for Ringhals 3. This approach is taken to obtain consistency between core designs and core calculations for the safety analyses and the cycle specific calculations, to minimize the risk for future violations of the safety analyses. (authors)

  12. Heavy density concrete for nuclear radiation shielding and power stations: [Part]2

    International Nuclear Information System (INIS)

    Singha Roy, P.K.

    1987-01-01

    This article is the second part of the paper entitled 'Heavy density concrete for nuclear radiation shielding and power stations'. In this part, some of the important properties of heavy density concrete are discussed. They include density, water retentivity, air content, permeability with special reference to concrete mixes used in India's nuclear power reactors. All these properties are affected to various extents by heating. Indian shield concrete is rarely subjected to temperatures above 60degC during its life, because of thermal shield protection. During placement, the maximum anticipated rise in temperature due to heat of hydration is restricted to around 45degC by chilling, if necessary to reduce shrinkage stresses and cracks. (M.G.B.)

  13. Core fusion accidents in nuclear power reactors. Knowledge review

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

  14. On-power verification of the dynamic response of self-powered in-core detectors

    International Nuclear Information System (INIS)

    Serdula, K.; Beaudet, M.

    1996-01-01

    Self-powered in-core detectors are used for on-line safety and regulation purposes in CANDU reactors. Such applications require use of detectors whose response is primarily prompt to changes in flux. In-service verification of the detectors' response is required to ensure significant degradation in performance has not occurred during long-term operation. Changes in the detector characteristics occur due to nuclear interactions and failures. Present verification requires significant station resources and disrupts power production. Use of the 'noise' in the detector signal is being investigated as an alternative to assess the dynamic response of the detectors during long-term operation. Measurements of reference 'signatures' were obtained from replacement shutdown system detectors. Results show 'noise' measurements are a promising alternative to the current verification method. Identification of changes in the detector response function assist in accurate diagnosis and prognosis of changes in detector signals due to process changes. (author)

  15. Fabrication of 3D Air-core MEMS Inductors for High Frequency Power Electronic Applications

    DEFF Research Database (Denmark)

    Lê Thanh, Hoà; Mizushima, Io; Nour, Yasser

    2018-01-01

    footprints have an inductance from 34.2 to 44.6 nH and a quality factor from 10 to 13 at frequencies ranging from 30 to 72 MHz. The air-core inductors show threefold lower parasitic capacitance and up to a 140% higher-quality factor and a 230% higher-operation frequency than silicon-core inductors. A 33 MHz...... boost converter mounted with an air-core toroidal inductor achieves an efficiency of 68.2%, which is better than converters mounted with a Si-core inductor (64.1%). Our inductors show good thermal cycling stability, and they are mechanically stable after vibration and 2-m-drop tests.......We report a fabrication technology for 3D air-core inductors for small footprint and very-high-frequency power conversions. Our process is scalable and highly generic for fabricating inductors with a wide range of geometries and core shapes. We demonstrate spiral, solenoid, and toroidal inductors...

  16. Implementation of a fast running full core pin power reconstruction method in DYN3D

    International Nuclear Information System (INIS)

    Gomez-Torres, Armando Miguel; Sanchez-Espinoza, Victor Hugo; Kliem, Sören; Gommlich, Andre

    2014-01-01

    Highlights: • New pin power reconstruction (PPR) method for the nodal diffusion code DYN3D. • Flexible PPR method applicable to a single, a group or to all fuel assemblies (square, hex). • Combination of nodal with pin-wise solutions (non-conform geometry). • PPR capabilities shown for REA of a Minicore (REA) PWR whole core. - Abstract: This paper presents a substantial extension of the pin power reconstruction (PPR) method used in the reactor dynamics code DYN3D with the aim to better describe the heterogeneity within the fuel assembly during reactor simulations. The flexibility of the new implemented PPR permits the local spatial refinement of one fuel assembly, of a cluster of fuel assemblies, of a quarter or eight of a core or even of a whole core. The application of PPR in core regions of interest will pave the way for the coupling with sub-channel codes enabling the prediction of local safety parameters. One of the main advantages of considering regions and not only a hot fuel assembly (FA) is the fact that the cross flow within this region can be taken into account by the subchannel code. The implementation of the new PPR method has been tested analysing a rod ejection accident (REA) in a PWR minicore consisting of 3 × 3 FA. Finally, the new capabilities of DNY3D are demonstrated by the analysing a boron dilution transient in a PWR MOX core and the pin power of a VVER-1000 reactor at stationary conditions

  17. Implementation of a fast running full core pin power reconstruction method in DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Torres, Armando Miguel [Instituto Nacional de Investigaciones Nucleares, Department of Nuclear Systems, Carretera Mexico – Toluca s/n, La Marquesa, 52750 Ocoyoacac (Mexico); Sanchez-Espinoza, Victor Hugo, E-mail: victor.sanchez@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-vom-Helmhotz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Kliem, Sören; Gommlich, Andre [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany)

    2014-07-01

    Highlights: • New pin power reconstruction (PPR) method for the nodal diffusion code DYN3D. • Flexible PPR method applicable to a single, a group or to all fuel assemblies (square, hex). • Combination of nodal with pin-wise solutions (non-conform geometry). • PPR capabilities shown for REA of a Minicore (REA) PWR whole core. - Abstract: This paper presents a substantial extension of the pin power reconstruction (PPR) method used in the reactor dynamics code DYN3D with the aim to better describe the heterogeneity within the fuel assembly during reactor simulations. The flexibility of the new implemented PPR permits the local spatial refinement of one fuel assembly, of a cluster of fuel assemblies, of a quarter or eight of a core or even of a whole core. The application of PPR in core regions of interest will pave the way for the coupling with sub-channel codes enabling the prediction of local safety parameters. One of the main advantages of considering regions and not only a hot fuel assembly (FA) is the fact that the cross flow within this region can be taken into account by the subchannel code. The implementation of the new PPR method has been tested analysing a rod ejection accident (REA) in a PWR minicore consisting of 3 × 3 FA. Finally, the new capabilities of DNY3D are demonstrated by the analysing a boron dilution transient in a PWR MOX core and the pin power of a VVER-1000 reactor at stationary conditions.

  18. Stability region for a prompt power variation of a coupled-core system with positive prompt feedback

    International Nuclear Information System (INIS)

    Watanabe, S.; Nishina, K.

    1984-01-01

    A stability analysis using a one-group model is presented for a coupled-core system. Positive prompt feedback of a γp /SUB j/ form is assumed, where p /SUB j/ is the fractional power variation of core j. Prompt power variations over a range of a few milliseconds after a disturbance are analyzed. The analysis combines Lapunov's method, prompt jump approximation, and the eigenfunction expansion of coupling region response flux. The last is treated as a pseudo-delayed neutron precursor. An asymptotic stability region is found for p /SUB j/. For an asymmetric flux variation over a system of two coupled cores, either p /SUB I/ or p /SUB II/ can slightly exceed, by virtue of the coupling effect, the critical value (β/γ-1) of a single-core case. Such a stability region is increased by additional inclusion of the coupling region fundamental mode in the treatment. The coupling region contributes to stability through its delayed response and coupling. An optimum core separation distance for stability is found

  19. A new approach to PWR power control using intelligent techniques

    International Nuclear Information System (INIS)

    Boroushaki, M.; Ghofrani, M.B.; Lucas, C.; Yazdanpanah, M.J.; Sadati, N.

    2004-01-01

    Improved load following capability is one of the main technical performances of advanced PWR(APWR). Controlling the nuclear reactor core during load following operation encounters some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking, while the core is subject to large and sharp variation of local power density during transients. Axial offset (A.O) is the parameter usually used to represent of core power peaking, in form of a practical parameter. This paper, proposes a new intelligent approach to A.o control of PWR nuclear reactors core during load following operation. This method uses a neural network model of the core to predict the dynamic behavior of the core and a fuzzy critic based on the operator knowledge and experience for the purpose of decision-making during load following operations. Simulation results show that this method can use optimum control rod groups maneuver with variable overlapping and may improve the reactor load following capability

  20. Unavailability Analysis of the Reactor Core Protection System using Reliability Block Diagram

    International Nuclear Information System (INIS)

    Shin, Hyun Kook; Kim, Sung Ho; Choi, Woong Suk; Kim, Jae Hack

    2006-01-01

    The reactor core of nuclear power plants needs to be monitored for the early detection of core abnormal conditions to protect plants from a severe accident. The core protection calculator system (CPCS) has been provided to calculate the departure from nucleate boiling ratio (DNBR) and the local power density (LPD) based on measured parameters of reactor and coolant system. The original CPCS for OPR 1000 has been designed and implemented based on the concurrent 3205 computer system whose components are obsolete. The CPCS based on Westinghouse Common-Q system has recently been implemented for the Shin-Kori Nuclear Power Plant, Units 1 and 2(SKN 1 and 2). An R and D project has been launched to develop new core protection system called as RCOPS (Reactor Core Protection System) with the partnership of KOPEC and Doosan Heavy Industries and Construction Co. RCOPS is implemented on the HFC-6000 safety class programmable logic controller (PLC). In this paper, the reliability of RCOPS is analyzed using the reliability block diagram (RBD) method. The calculated results are compared with that of the CPCS for SKN 1 and 2

  1. Radiation power profiles and density limit with a divertor in the W7-AS stellarator

    International Nuclear Information System (INIS)

    Giannone, L.; Burhenn, R.; McCormick, K.; Brakel, R.; Feng, Y.; Grigull, P.; Igitkhanov, Y.

    2002-01-01

    The addition of a divertor into the W7-AS stellarator has allowed access to a high density regime where the radiation profiles reach a steady state. In earlier limiter discharges, the plasma suffered a radiative collapse at high densities. In contrast to limiter experiments, where the impurity confinement time measured by Al laser blow-off increased with increasing line integrated density, in divertor discharges, above a density threshold, the impurity confinement time decreased with increasing line integrated density. The observation that the divertor plasma radiates mainly at the plasma edge rather than the plasma centre is a further indication that changes to the impurity transport coefficients at these high densities are the basis for the achievement of steady state discharges in the divertor configuration of W7-AS. The maximum line integrated density reached with a divertor is compared to that reached with a limiter. The previously derived scaling law for the density limit with a limiter shows that the achieved densities do not exceed those predicted when the higher deposited power is taken into account. In a divertor the radiated power is located at the plasma edge and increasing the density, cooling the plasma edge and radiating sufficient power to cause plasma detachment determines the density limit. (author)

  2. Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

    2009-01-01

    In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

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

    International Nuclear Information System (INIS)

    Hecker, H.C.

    1984-04-01

    This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U 233 -Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred

  4. ALMA OBSERVATIONS OF A HIGH-DENSITY CORE IN TAURUS: DYNAMICAL GAS INTERACTION AT THE POSSIBLE SITE OF A MULTIPLE STAR FORMATION

    Energy Technology Data Exchange (ETDEWEB)

    Tokuda, Kazuki; Onishi, Toshikazu [Department of Physical Science, Graduate School of Science, Osaka Prefecture University, 1-1 Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Saigo, Kazuya; Kawamura, Akiko [National Astronomical Observatory of Japan, Mitaka, Tokyo 181-8588 (Japan); Fukui, Yasuo; Inutsuka, Shu-ichiro; Tachihara, Kengo [Department of Physics, Nagoya University, Chikusa-ku, Nagoya 464-8602 (Japan); Matsumoto, Tomoaki [Faculty of Humanity and Environment, Hosei University, Fujimi, Chiyoda-ku, Tokyo 102-8160 (Japan); Machida, Masahiro N. [Department of Earth and Planetary Sciences, Kyushu University, Fukuoka 812-8581 (Japan); Tomida, Kengo, E-mail: s_k.tokuda@p.s.osakafu-u.ac.jp [Department of Astrophysical Sciences, Princeton University, Princeton, NJ 08544 (United States)

    2014-07-01

    Starless dense cores eventually collapse dynamically, forming protostars inside them, and the physical properties of the cores determine the nature of the forming protostars. We report ALMA observations of dust continuum emission and molecular rotational lines toward MC27 or L1521F, which is considered to be very close to the first protostellar core phase. We found a few starless high-density cores, one of which has a very high density of ∼10{sup 7} cm{sup –3}, within a region of several hundred AU around a very low-luminosity protostar detected by Spitzer. A very compact bipolar outflow with a dynamical timescale of a few hundred years was found toward the protostar. The molecular line observation shows several cores with an arc-like structure, possibly due to the dynamical gas interaction. These complex structures revealed in the present observations suggest that the initial condition of star formation is highly dynamical in nature, which is considered to be a key factor in understanding fundamental issues of star formation such as the formation of multiple stars and the origin of the initial mass function of stars.

  5. Non-power law behavior of the radial profile of phase-space density of halos

    International Nuclear Information System (INIS)

    Popolo, A. Del

    2011-01-01

    We study the pseudo phase-space density, ρ(r)/σ 3 (r), of ΛCDM dark matter halos with and without baryons (baryons+DM, and pure DM), by using the model introduced in Del Popolo (2009), which takes into account the effect of dynamical friction, ordered and random angular momentum, baryons adiabatic contraction and dark matter baryons interplay. We examine the radial dependence of ρ(r)/σ 3 (r) over 9 orders of magnitude in radius for structures on galactic and cluster of galaxies scales. We find that ρ(r)/σ 3 (r) is approximately a power-law only in the range of halo radius resolved by current simulations (down to 0.1% of the virial radius) while it has a non-power law behavior below the quoted scale, with inner profiles changing with mass. The non-power-law behavior is more evident for halos constituted both of dark matter and baryons while halos constituted just of dark matter and with angular momentum chosen to reproduce a Navarro-Frenk-White (NFW) density profile, are characterized by an approximately power-law behavior. The results of the present paper lead to conclude that density profiles of the NFW type are compatible with a power-law behavior of ρ(r)/σ 3 (r), while those flattening to the halo center, like those found in Del Popolo (2009) or the Einasto profile, or the Burkert profile, cannot produce radial profile of the pseudo-phase-space density that are power-laws at all radii. The results argue against universality of the pseudo phase-space density and as a consequence argue against universality of density profiles constituted by dark matter and baryons as also discussed in Del Popolo (2009)

  6. Density matrix embedding in an antisymmetrized geminal power bath

    International Nuclear Information System (INIS)

    Tsuchimochi, Takashi; Welborn, Matthew; Van Voorhis, Troy

    2015-01-01

    Density matrix embedding theory (DMET) has emerged as a powerful tool for performing wave function-in-wave function embedding for strongly correlated systems. In traditional DMET, an accurate calculation is performed on a small impurity embedded in a mean field bath. Here, we extend the original DMET equations to account for correlation in the bath via an antisymmetrized geminal power (AGP) wave function. The resulting formalism has a number of advantages. First, it allows one to properly treat the weak correlation limit of independent pairs, which DMET is unable to do with a mean-field bath. Second, it associates a size extensive correlation energy with a given density matrix (for the models tested), which AGP by itself is incapable of providing. Third, it provides a reasonable description of charge redistribution in strongly correlated but non-periodic systems. Thus, AGP-DMET appears to be a good starting point for describing electron correlation in molecules, which are aperiodic and possess both strong and weak electron correlation

  7. Determining the magnetically nonlinear characteristics of a three phase core-type power transformer

    International Nuclear Information System (INIS)

    Dolinar, Matjaz; Stumberger, Gorazd; Polajzer, Bostjan; Dolinar, Drago

    2006-01-01

    This paper presents nonlinear iron core model of a three-phase, three-limb power transformer which is given by the current-dependant characteristics of flux linkages. The magnetically nonlinear characteristics are determined by controlled magnetic excitation of all three limbs which allows to take into account the variable magnetic-cross couplings between different coils placed on limbs, caused by saturation. The corresponding partial derivatives of measured flux linkage characteristics are used in the transformer circuit model as a magnetically nonlinear iron core model in order to analyze the behaviour of a nonsymmetrically excited transformer. Numerical results using transformer model with the determined iron core model agree very well with the measured results

  8. Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems

    International Nuclear Information System (INIS)

    Chow, S.

    1976-01-01

    A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. In the power generation mode, the plasma and propellant flows are shut off, and the driver elements supply thermal power to the power conversion system, which generates electricity for primary electric propulsion purposes

  9. Core-power and decay-time limits for disabled automatic-actuation of LOFT ECCS

    International Nuclear Information System (INIS)

    Hanson, G.H.

    1978-01-01

    The Emergency Core Cooling System (ECCS) for the LOFT reactor may need to be disabled for modifications or repairs of hardware or instrumentation or for component testing during periods when the reactor system is hot and pressurized, or it may be desirable to enable the ECCS to be disabled without the necessity of cooling down and depressurizing the reactor. A policy involves disabling the automatic-actuation of the LOFT ECCS, but still retaining the manual actuation capability. Disabling of the automatic actuation can be safely utilized, without subjecting the fuel cladding to unacceptable temperatures, when the LOFT power decays to 33 kW; this power level permits a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS. For the operating power of the L2-2 Experiment, the required decay-periods (with operating periods of 40 and 2000 hours) are about 21 and 389 hours, respectively. With operating periods of 40 and 2000 hours at Core-I full power, the required decay-periods are about 42 and 973 hours, respectively. After these decay periods the automatic actuation of the LOFT ECCS can be disabled assuming a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS. The automatic and manual lineup of the ECCS may be waived if decay power is less than 11 kW

  10. New techniques for designing the initial and reload cores with constant long cycle lengths

    International Nuclear Information System (INIS)

    Shi, Jun; Levine, Samuel; Ivanov, Kostadin

    2017-01-01

    Highlights: • New techniques for designing the initial and reload cores with constant long cycle lengths are developed. • Core loading pattern (LP) calculations and comparisons have been made on two different designs. • Results show that significant savings in fuel costs can be accrued if a non-low leakage LP design strategy is enacted. - Abstract: Several utilities have increased the output power of their nuclear power plant to increase their income and profit. Thus, the utility increases the power density of the reactor, which has other consequences. One consequence is to increase the depletion of the fuel assemblies (FAs) and reduce the end-of-cycle (EOC) sum of fissionable nuclides in each FA, ∑_E_O_C. The power density and the ∑_E_O_C remaining in the FAs at EOC must be sufficiently large in many FAs when designing the loading pattern, LP, for the first and reload cycles to maintain constant cycle lengths at minimum fuel cost. Also of importance is the cycle length as well as several other factors. In fact, the most important result of this study is to understand that the ∑_E_O_Cs in the FAs must be such that in the next cycle they can sustain the energy during depletion to prevent too much power shifting to the fresh FAs and, thus, sending the maximum peak pin power, PPP_m_a_x, above its constraint. This paper presents new methods for designing the LPs for the initial and follow on cycles to minimize the fuel costs. Studsvik’s CMS code system provides a 1000 MWe LP design in their sample inputs, which is applied in this study. The first 3 cycles of this core are analyzed to minimize fuel costs, and all three cycles have the same cycle length of ∼650 days. Cycle 1 is designed to allow many used FAs to be loaded into cycles 2 and 3 to reduce their fuel costs. This could not be achieved if cycle 1 was a low leakage LP (Shi et al., 2015). Significant fuel cost savings are achieved when the new designs are applied to the higher leakage LP designs

  11. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

    The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

  12. High-density multicore fibers

    DEFF Research Database (Denmark)

    Takenaga, K.; Matsuo, S.; Saitoh, K.

    2016-01-01

    High-density single-mode multicore fibers were designed and fabricated. A heterogeneous 30-core fiber realized a low crosstalk of −55 dB. A quasi-single-mode homogeneous 31-core fiber attained the highest core count as a single-mode multicore fiber.......High-density single-mode multicore fibers were designed and fabricated. A heterogeneous 30-core fiber realized a low crosstalk of −55 dB. A quasi-single-mode homogeneous 31-core fiber attained the highest core count as a single-mode multicore fiber....

  13. Feasibility of maintaining natural convection mode core cooling in research reactor power upgrades

    International Nuclear Information System (INIS)

    Ha, J.J.; Belhadj, M.; Aldemir, T.; Christensen, R.N.

    1987-01-01

    Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top 16 N activity (PTNA), and 2) nucleate boiling in core channels. The feasibility assessment of a power upgrade while maintaining natural convection mode core cooling requires addressing these operational concerns. Previous studies have shown that: a) The conventional technique for reducing PTNA by plume dispersion may not be effective in a large power upgrade of research reactors with small pools. b) Currently used correlations to predict onset of nucleate boiling (ONB) in thin, rectangular core channels are not valid for low-velocity, upward flows such as encountered in natural convection cooling. The PTNA depends on the velocity distribution in the reactor pool. COMMIX-1A code is used to determine the three-dimensional velocity fields in The Ohio State University Research Reactor (OSURR) pool as a function of varying design conditions, following a power upgrade to 500 kW with LEU fuel. It is shown that a sufficiently deep stagnant water layer can be created below the pool top by properly choosing the disperser flow rate. The ONB heat flux is experimentally determined for channel gaps and upward flow velocities in the range 2mm-4mm and 3-16 cm/sec., respectively. Two alternatives to plume dispersion for reducing PTNA and a new correlation to determine the ONB heat flux in thin, rectangular channels under low-velocity, upward flow conditions are proposed. (Author)

  14. Analysis of reactivity accidents of the RSG-GAS core with silicide fuel

    International Nuclear Information System (INIS)

    Tukiran

    2002-01-01

    The fuels of RSG-GAS reactor is changed from uranium oxide to uranium silicide. For time being, the fuel of RSG-GAS core are mixed up between oxide and silicide fuels with 250 gr of loading and 2.96 g U/cm 3 of density, respectively. While, silicide fuel with 300 gr of loading is still under research. The advantages of silicide fuels are can be used in high density, so that, it can be stayed longer in the core at higher burn-up, therefore, the length of cycle is longer. The silicide fuel in RSG-GAS core is used in step-wise by using mixed up core. Firstly, it is used silicide fuel with 250 gr of loading and then, silicide fuel with 300 gr of loading (3.55 g U/cm 3 of density). In every step-wise of fuel loading must be analysed its safety margin. In this occasion, it is analysed the reactivity accident of RSG-GAS core with 300 gr of silicide fuel loading. The calculation was done by using POKDYN code which available at P2TRR. The calculation was done by reactivity insertion at start up and power rangers. From all cases which were have been done, the results of analysis showed that there is no anomaly and safety margin break at RSG-GAS core with 300 gr silicide fuel loading

  15. Power distribution monitoring system in the boiling water cooled reactor core

    International Nuclear Information System (INIS)

    Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskij, I.I.

    1987-01-01

    Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

  16. Method of controlling power distribution in FBR type reactors

    International Nuclear Information System (INIS)

    Sawada, Shusaku; Kaneto, Kunikazu.

    1982-01-01

    Purpose: To attain the power distribution flattening with ease by obtaining a radial power distribution substantially in a constant configuration not depending on the burn-up cycle. Method: As the fuel burning proceeds, the radial power distribution is effected by the accumulation of fission products in the inner blancket fuel assemblies which varies the effect thereof as the neutron absorbing substances. Taking notice of the above fact, the power distribution is controlled in a heterogeneous FBR type reactor by varying the core residence period of the inner blancket assemblies in accordance with the charging density of the inner blancket assemblies in the reactor core. (Kawakami, Y.)

  17. Anisotropic nanolaminated CoNiFe cores integrated into microinductors for high-frequency dc–dc power conversion

    International Nuclear Information System (INIS)

    Kim, Jooncheol; Kim, Minsoo; Herrault, Florian; Kim, Jung-Kwun; Allen, Mark G

    2015-01-01

    This paper presents a rectangular, anisotropic nanolaminated CoNiFe core that possesses a magnetically hard axis in the long geometric axis direction. Previously, we have developed nanolaminated cores comprising tens to hundreds of layers of 300–1000 nm thick metallic alloys (i.e. Ni 80 Fe 20 or Co 44 Ni 37 Fe 19 ) based on sequential electrodeposition, demonstrating suppressed eddy-current losses at MHz frequencies. In this work, magnetic anisotropy was induced to the nanolaminated CoNiFe cores by applying an external magnetic field (50–100 mT) during CoNiFe film electrodeposition. The fabricated cores comprised tens to hundreds of layers of 500–1000 nm thick CoNiFe laminations that have the hard-axis magnetic property. Packaged in a 22-turn solenoid test inductor, the anisotropic core showed 10% increased effective permeability and 25% reduced core power losses at MHz operation frequency, compared to an isotropic core of the identical geometry. Operating the anisotropic nanolaminated CoNiFe core in a step-down dc–dc converter (15 V input to 5 V output) demonstrated 81% converter efficiency at a switching frequency of 1.1 MHz and output power of 6.5 W. A solenoid microinductor with microfabricated windings integrated with the anisotropic nanolaminated CoNiFe core was fabricated, demonstrating a constant inductance of 600 nH up to 10 MHz and peak quality factor exceeding 20 at 4 MHz. The performance of the microinductor with the anisotropic nanolaminated CoNiFe core is compared with other previously reported microinductors. (fast track communication)

  18. Self-Powered Neutron and Gamma Detectors for In-Core Measurements

    International Nuclear Information System (INIS)

    Strindehag, O.

    1971-11-01

    The performance of various types of self-powered neutron and gamma detectors intended for control and power distribution measurements in water cooled reactors is discussed. The self-powered detectors are compared with other types of in-core detectors and attention is paid to such properties as neutron and gamma sensitivity, high-temperature performance, burn-up rate and time of response. Also treated are the advantages and disadvantages of using gamma detector data for power distribution calculations instead of data from neutron detectors. With regard to neutron-sensitive detectors, results from several long-term experiments with vanadium and cobalt detectors are presented. The results include reliability and stability data for these two detector types and the Co build-up in cobalt detectors. Experimental results which reveal the fast response of cobalt detectors are presented, and the use of cobalt detectors in reactor safety systems is discussed. Experience of the design and installation of complete flux probes, electronic units and data processing systems for power reactors is reported. The investigation of gamma-sensitive detectors includes detectors with emitters of lead, zirconium, magnesium and Inconel. Measured gamma sensitivities from calibrations both in a reactor and in a gamma cell are given, and the signal levels of self-powered neutron and gamma detectors when applied to power reactors are compared

  19. Self-Powered Neutron and Gamma Detectors for In-Core Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Strindehag, O

    1971-11-15

    The performance of various types of self-powered neutron and gamma detectors intended for control and power distribution measurements in water cooled reactors is discussed. The self-powered detectors are compared with other types of in-core detectors and attention is paid to such properties as neutron and gamma sensitivity, high-temperature performance, burn-up rate and time of response. Also treated are the advantages and disadvantages of using gamma detector data for power distribution calculations instead of data from neutron detectors. With regard to neutron-sensitive detectors, results from several long-term experiments with vanadium and cobalt detectors are presented. The results include reliability and stability data for these two detector types and the Co build-up in cobalt detectors. Experimental results which reveal the fast response of cobalt detectors are presented, and the use of cobalt detectors in reactor safety systems is discussed. Experience of the design and installation of complete flux probes, electronic units and data processing systems for power reactors is reported. The investigation of gamma-sensitive detectors includes detectors with emitters of lead, zirconium, magnesium and Inconel. Measured gamma sensitivities from calibrations both in a reactor and in a gamma cell are given, and the signal levels of self-powered neutron and gamma detectors when applied to power reactors are compared

  20. Profiles of radiation power density in WEGA stellarator

    International Nuclear Information System (INIS)

    Zhang, D.; Otte, M.; Giannone, L.

    2005-01-01

    On the WEGA stellarator, a 12 channel bolometer camera has been used to measure the radiation power losses of the plasma, which is heated by ECR at 2.45 GHz with a maximum power of 26 kW. The typical electron temperatures achieved are around 10 eV. The bolometer is of the Au resistor type and is positioned on the mid-plane, viewing the plasma from the low-field side with a spatial resolution of about 6 cm. The viewing angle is opened to poloidally (±47 o ) and covers the whole cross-section. Angular profiles of radiation power density (emissivity) can be achieved using the measured fluxes to the channels, which are given by the integrals along the sight lines. Using Abel inversion with maximum entropy regularisation, radial profiles of emissivity could be obtained. It is found that the angular profile of emissivity depends on the magnetic configuration, the working gas (Ar, He) and the heating scenario. Peaked and hollow emissivity profiles have been obtained by using different types of heating antenna. By changing the magnetic configuration, strong edge radiation has been observed. The largest emissivity values are obtained in the upper SOL range of Ar-discharges. This edge radiation can be reduced by shifting the flux surfaces inwards or by changing their shape at the antenna. The reconstruction of the radial profile of the emissivity was carried out in the case of a peaked angular profile with minimum edge radiation. The total radiation power was estimated by linear extrapolation of the integrated radiation power in the viewing region to the torus volume. It is typically less than 30% of the ECRH input power, but depending on the ECRH input power, again the magnetic configuration, the working gas as well as the absolute field strength on the magnetic axis. Maximum radiation losses have been obtained around 0.6·B0, where B 0 =87.5 mT is the resonant field strength of the ECRH. No evidence for impurities was obtained from spectroscopic measurements, and thus the

  1. Neutronic Analysis of the RSG-GAS Compact Core without CIP Silicide 3.55 g U/cc and 4.8 g U/cc

    International Nuclear Information System (INIS)

    Jati S; Lily S; Tukiran S

    2004-01-01

    Fuel conversion from U 3 O 8 -Al to U 3 Si 2 -Al 2.96 g U/cc density in the RSG-GAS core had done successfully step by step since 36 th core until 44 th core. So that, since the 45 th core until now (48 th core) had been using full of silicide 2.96 g U/cc. Even though utilization program of silicide fuel with high density (3.55 g U/cc and 4.8 g U/cc) and optimize operation of RSG-GAS core under research. Optimalitation of core with increasing operation cycle have been analyzing about compact core. The mean of compact core is the RSG-GAS core with decrease number of IP or CIP position irradiation. In this research, the neutronic calculation to cover RSG-GAS core and RSG-GAS core without CIP that are using U 3 Si 2 -Al 2.96 g U/cc, 3.55 g U/cc and 4.8 g U/cc had done. Two core calculation done at 15 MW power using SRAC-ASMBURN code. The calculation result show that fuel conversion from 2.96 g U/cc density to 3.55 g U/cc and 4.8 g U/cc will increasing cycle length for both RSG-GAS core and RSG-GAS compact core without CIP. However, increasing of excess reactivity exceeded from nominal value of first design that 9.2%. Change of power peaking factor is not show significant value and still less than 1.4. Core fuelled with U 3 Si 2 -Al 4.8 g U/cc density have maximum discharge burn-up which exceeded from licensing value (70%). RSG-GAS compact core without CIP fuelled U 3 Si 2 -Al 2.96 g U/cc have longer cycle operation then RSG-GAS core and fulfil limitation neutronic parameter at the first design value. (author)

  2. Galactic densities, substructure and the initial power spectrum

    International Nuclear Information System (INIS)

    Bullock, J.S.; Zentner, A.R.

    2003-01-01

    Although the currently favored cold dark matter plus cosmological constant model for structure formation assumes an n = 1 scale-invariant initial power spectrum, most inflation models produce at least mild deviations from n = 1. Because the lever arm from the CMB normalization to galaxy scales is long, even a small 'tilt' can have important implications for galactic observations. Here we calculate the COBS-normalized power spectra for several well-motivated models of inflation and compute implications for the substructure content and central densities of galaxy halos. Using an analytic model, normalized against N-body simulations, we show that while halos in the standard (n = 1) model are overdense by a factor of ∼ 6 compared to observations, several of our example inflation+LCDM models predict halo densities well within the range of observations, which prefer models with n ∼ 0.85. We go on to use a semi-analytic model (also normalized against N-body simulations) to follow the merger histories of galaxy-sized halos and track the orbital decay, disruption, and evolution of the merging substructure. Models with n ∼ 0.85 predict a factor of ∼ 3 fewer subhalos at a fixed circular velocity than the standard n 1 case. Although this level of reduction does not resolve the 'dwarf satellite problem', it does imply that the level of feedback required to match the observed number of dwarfs is sensitive to the initial power spectrum. Finally, the fraction of galaxy-halo mass that is bound up in substructure is consistent with limits imposed by multiply imaged quasars for all models considered: f sat > 0.01 even for an effective tilt of n ∼ 0.8. We conclude that, at their current level, lensing constraints of this kind do not provide an interesting probe of the primordial power spectrum

  3. Validation of reactor core protection system

    International Nuclear Information System (INIS)

    Lee, Sang-Hoon; Bae, Jong-Sik; Baeg, Seung-Yeob; Cho, Chang-Ho; Kim, Chang-Ho; Kim, Sung-Ho; Kim, Hang-Bae; In, Wang-Kee; Park, Young-Ho

    2008-01-01

    Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

  4. THE CASE AGAINST WARM OR SELF-INTERACTING DARK MATTER AS EXPLANATIONS FOR CORES IN LOW SURFACE BRIGHTNESS GALAXIES

    International Nuclear Information System (INIS)

    Kuzio de Naray, Rachel; Martinez, Gregory D.; Bullock, James S.; Kaplinghat, Manoj

    2010-01-01

    Warm dark matter (WDM) and self-interacting dark matter (SIDM) are often motivated by the inferred cores in the dark matter halos of low surface brightness (LSB) galaxies. We test thermal WDM, non-thermal WDM, and SIDM using high-resolution rotation curves of nine LSB galaxies. We fit these dark matter models to the data and determine the halo core radii and central densities. While the minimum core size in WDM models is predicted to decrease with halo mass, we find that the inferred core radii increase with halo mass and also cannot be explained with a single value of the primordial phase-space density. Moreover, if the core size is set by WDM particle properties, then even the smallest cores we infer would require primordial phase-space density values that are orders of magnitude smaller than lower limits obtained from the Lyα forest power spectra. We also find that the dark matter halo core densities vary by a factor of about 30 from system to system while showing no systematic trend with the maximum rotation velocity of the galaxy. This strongly argues against the core size being directly set by large self-interactions (scattering or annihilation) of dark matter. We therefore conclude that the inferred cores do not provide motivation to prefer WDM or SIDM over other dark matter models.

  5. Graphene-based in-plane micro-supercapacitors with high power and energy densities.

    Science.gov (United States)

    Wu, Zhong-Shuai; Parvez, Khaled; Feng, Xinliang; Müllen, Klaus

    2013-01-01

    Micro-supercapacitors are important on-chip micro-power sources for miniaturized electronic devices. Although the performance of micro-supercapacitors has been significantly advanced by fabricating nanostructured materials, developing thin-film manufacture technologies and device architectures, their power or energy densities remain far from those of electrolytic capacitors or lithium thin-film batteries. Here we demonstrate graphene-based in-plane interdigital micro-supercapacitors on arbitrary substrates. The resulting micro-supercapacitors deliver an area capacitance of 80.7 μF cm⁻² and a stack capacitance of 17.9 F cm⁻³. Further, they show a power density of 495 W cm⁻³ that is higher than electrolytic capacitors, and an energy density of 2.5 mWh cm⁻³ that is comparable to lithium thin-film batteries, in association with superior cycling stability. Such microdevices allow for operations at ultrahigh rate up to 1,000 V s⁻¹, three orders of magnitude higher than that of conventional supercapacitors. Micro-supercapacitors with an in-plane geometry have great promise for numerous miniaturized or flexible electronic applications.

  6. Graphene-based in-plane micro-supercapacitors with high power and energy densities

    Science.gov (United States)

    Wu, Zhong–Shuai; Parvez, Khaled; Feng, Xinliang; Müllen, Klaus

    2013-01-01

    Micro-supercapacitors are important on-chip micro-power sources for miniaturized electronic devices. Although the performance of micro-supercapacitors has been significantly advanced by fabricating nanostructured materials, developing thin-film manufacture technologies and device architectures, their power or energy densities remain far from those of electrolytic capacitors or lithium thin-film batteries. Here we demonstrate graphene-based in-plane interdigital micro-supercapacitors on arbitrary substrates. The resulting micro-supercapacitors deliver an area capacitance of 80.7 μF cm−2 and a stack capacitance of 17.9 F cm−3. Further, they show a power density of 495 W cm−3 that is higher than electrolytic capacitors, and an energy density of 2.5 mWh cm−3 that is comparable to lithium thin-film batteries, in association with superior cycling stability. Such microdevices allow for operations at ultrahigh rate up to 1,000 V s−1, three orders of magnitude higher than that of conventional supercapacitors. Micro-supercapacitors with an in-plane geometry have great promise for numerous miniaturized or flexible electronic applications. PMID:24042088

  7. JOYO MK-II core characteristics database. Update to JFS-3-J3.2R

    International Nuclear Information System (INIS)

    Ohkawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi

    2003-04-01

    The 'JOYO' MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition in 2001. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. However, after the database was published, it was recently found that there were errors in the process of making the group constant set JFS-3-J3.2, and it was revised at JFS-3-J3.2R. Then, the group constant set was updated at JFS-3-J3.2R in this database. The MK-II core management data nad core characteristics data were recorded on CD-ROM for user convenience. The structure of the database is the same as in the first edition. The 'Configuration Data' include the core arrangement and refueling record for each operational cycle. The 'Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The 'Output Data' contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The 'Core Characteristics Data' include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. The effect of updating the group constant set, the calculation results of excess reactivity decreased by about 0.15Δk/kk', and the effects to other core characteristics were negligible. (author)

  8. Development of in-core measurements in the reactor KS-150

    International Nuclear Information System (INIS)

    Rana, S.B.

    1977-01-01

    Mapping of the neutron flux density distribution and of the neutron fluence distribution in the KS-150 reactor core was carried out using an in-core measuring system. The system allows the in-service monitoring of important operating properties of the reactor core and fuel elements and consists of a mapping fuel element assembly with built-in SPN detectors, of transmission paths and a computer facility. The measurement of the neutron flux, neutron fluence and temperature fields in the reactor core was carried out during the power start-up of the reactor using self-powered DPZ-1 detectors. The obtained data are given and the axial distribution of neutron flux is graphically represented for different values of burnup at the same configuration of regulating rods, as is the axial distribution of neutron fluence for different configurations of the regulating rods during operation, and the in-service neutron fluence distribution. The maximal fuel temperature of 500.2 degC was found at a distance of 291.2 cm from the upper boundary of the reactor core, at a neutron flux of 1.46x10 14 n/cm 2 s. In comparison with other methods, this method proved easy and quick, the results reliable, reactivity perturbance negligible and the fuel element cost increase a negligible 4%. Neutron flux mapping using in-core self-powered detectors will be performed on a wider scale. (J.P./J.O.)

  9. Hydrogen consumption and power density in a co-flow planar SOFC

    Energy Technology Data Exchange (ETDEWEB)

    Ben Moussa, Hocine; Zitouni, Bariza [Laboratoire d' etude des systemes energetiques industriels (LESEI), Universite de Batna, Batna (Algeria); Oulmi, Kafia [Laboratoire de chimie et de chimie de l' environnement, Universite de Batna, Batna (Algeria); Mahmah, Bouziane; Belhamel, Maiouf [CDER, BP. 62 Route de l' Observatoire. Bouzareah. Alger (Algeria); Mandin, Philippe [Centre de Developpement des Energies Renouvelables (CDER), LECA, UMR 7575 CNRS-ENSCP Paris 6 (France)

    2009-06-15

    In the present work, power density and hydrogen consumption in a co-flow planar solid oxide fuel cell (SOFC) are studied according to the inlet functional parameters; such as the operational temperature, the operational pressure, the flow rates and the mass fractions of the species. Furthermore, the effect of the cell size is investigated. The results of a zero and a one-dimensional numerical electro-dynamic model predict the remaining quantity of the fed hydrogen at the output of the anode flow channel. The remaining hydrogen quantities and the SOFC's power density obtained are discussed as a function of the inlet functional parameters, the geometrical configuration of the cell and several operating cell voltages values. (author)

  10. Magnetic field power density spectra during 'scatter-free' solar particle events

    Science.gov (United States)

    Tan, L. C.; Mason, G. M.

    1993-01-01

    We have examined interplanetary magnetic field power spectral density during four previously identified 3He-rich flare periods when the about 1 MeV nucleon-1 particles exhibited nearly scatter-free transport from the sun to 1 AU. Since the scattering mean free path A was large, it might be expected that interplanetary turbulence was low, yet the spectral density value was low only for one of the four periods. For the other three, however, the spectral index q of the power density spectrum was near 2.0, a value at which quasi-linear theories predict an increase in the scattering mean free path. Comparing the lambda values from the energetic particles with that computed from a recent quasi-linear theory which includes helicity and the propagation direction of waves, we find lambda(QLT)/lambda(SEP) = 0.08 +/- 0.03 for the four events. Thus, the theory fits the q-dependence of lambda; however, as found for previous quasi-linear theories, the absolute value is low.

  11. A High Power Density Integrated Charger for Electric Vehicles with Active Ripple Compensation

    OpenAIRE

    Pan, Liwen; Zhang, Chengning

    2015-01-01

    This paper suggests a high power density on-board integrated charger with active ripple compensation circuit for electric vehicles. To obtain a high power density and high efficiency, silicon carbide devices are reported to meet the requirement of high-switching-frequency operation. An integrated bidirectional converter is proposed to function as AC/DC battery charger and to transfer energy between battery pack and motor drive of the traction system. In addition, the conventional H-bridge cir...

  12. Nuclear characteristic simulation device for reactor core

    International Nuclear Information System (INIS)

    Arakawa, Akio; Kobayashi, Yuji.

    1994-01-01

    In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

  13. Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-05-01

    It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

  14. Development of core design and analyses technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  15. Development of core design and analyses technology for integral reactor

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Dong Hyuk; Shim, Hyung Jin; Kim, Chang Hyo [Seoul National University, Seoul (Korea, Republic of)

    2014-06-15

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

  17. Automatic determination of pressurized water reactor core loading patterns which maximize end-of-cycle reactivity within power peaking and burnup constraints

    International Nuclear Information System (INIS)

    Hobson, G.H.

    1985-01-01

    An automated procedure for determining the optimal core loading pattern for a pressurized water reactor which maximizes end-of-cycle k/sub eff/ while satisfying constraints on power peaking and discharge burnup has been developed. The optimization algorithm combines a two energy group, two-dimensional coarse-mesh finite difference diffusion theory neutronics model to simulate core conditions, a perturbation theory approach to determine reactivity, flux, power and burnup changes as a function of assembly shuffling, and Monte Carlo integer programming to select the optimal loading pattern solution. The core examined was a typical Cycle 2 reload with no burnable poisons. Results indicate that the core loading pattern that maximizes end-of-cycle k/sub eff/ results in a 5.4% decrease in fuel cycle costs compared with the core loading pattern that minimizes the maximum relative radial power peak

  18. Development of measuring system with self-powered neutron detectors for the LR-0 reactor

    International Nuclear Information System (INIS)

    Erben, O.; Horinek, K.; Szasz, Z.

    1989-01-01

    A measuring channel with self-powered detectors was developed for measuring neutron fluxs density in the reactor core. The measuring channel consists of a measuring probe with standard self-powered detectors of Soviet make, a signal pathway, a current/voltage converter and a measuring and recording unit. Neutron flux density in the LR-0 reactor core reaches a maximum of 10 13 m -2 s -1 . Experiments using the channel were carried out both in steady-state operation and after emergency shutdown of the reactor, this from power levels of 2,096 W and 1,830 W. The results of the experiments are tabulated and briefly analyzed. (Z.M.). 4 figs., 3 tabs., 5 refs

  19. Investigating heavy water zero power reactors with a new core configuration based on experiment and calculation results

    Energy Technology Data Exchange (ETDEWEB)

    Nasrazadani, Zahra; Salimi, Raana; Askari, Afrooz; Khorsandi, Jamshid; Mirvakili, Mohammad; Mashayekh, Mohammad [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Esfahan (Iran, Islamic Republic of)

    2017-02-15

    The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor (Keff) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of D2O, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

  20. The verification of PWR-fuel code for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Surian Pinem; Tagor M Sembiring; Tukiran

    2015-01-01

    In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

  1. Formation of massive, dense cores by cloud-cloud collisions

    Science.gov (United States)

    Takahira, Ken; Shima, Kazuhiro; Habe, Asao; Tasker, Elizabeth J.

    2018-05-01

    We performed sub-parsec (˜ 0.014 pc) scale simulations of cloud-cloud collisions of two idealized turbulent molecular clouds (MCs) with different masses in the range of (0.76-2.67) × 104 M_{⊙} and with collision speeds of 5-30 km s-1. Those parameters are larger than in Takahira, Tasker, and Habe (2014, ApJ, 792, 63), in which study the colliding system showed a partial gaseous arc morphology that supports the NANTEN observations of objects indicated to be colliding MCs using numerical simulations. Gas clumps with density greater than 10-20 g cm-3 were identified as pre-stellar cores and tracked through the simulation to investigate the effects of the mass of colliding clouds and the collision speeds on the resulting core population. Our results demonstrate that the smaller cloud property is more important for the results of cloud-cloud collisions. The mass function of formed cores can be approximated by a power-law relation with an index γ = -1.6 in slower cloud-cloud collisions (v ˜ 5 km s-1), and is in good agreement with observation of MCs. A faster relative speed increases the number of cores formed in the early stage of collisions and shortens the gas accretion phase of cores in the shocked region, leading to the suppression of core growth. The bending point appears in the high-mass part of the core mass function and the bending point mass decreases with increase in collision speed for the same combination of colliding clouds. The higher-mass part of the core mass function than the bending point mass can be approximated by a power law with γ = -2-3 that is similar to the power index of the massive part of the observed stellar initial mass function. We discuss implications of our results for the massive-star formation in our Galaxy.

  2. Comparison of energy efficiency and power density in pressure retarded osmosis and reverse electrodialysis.

    Science.gov (United States)

    Yip, Ngai Yin; Elimelech, Menachem

    2014-09-16

    Pressure retarded osmosis (PRO) and reverse electrodialysis (RED) are emerging membrane-based technologies that can convert chemical energy in salinity gradients to useful work. The two processes have intrinsically different working principles: controlled mixing in PRO is achieved by water permeation across salt-rejecting membranes, whereas RED is driven by ion flux across charged membranes. This study compares the energy efficiency and power density performance of PRO and RED with simulated technologically available membranes for natural, anthropogenic, and engineered salinity gradients (seawater-river water, desalination brine-wastewater, and synthetic hypersaline solutions, respectively). The analysis shows that PRO can achieve both greater efficiencies (54-56%) and higher power densities (2.4-38 W/m(2)) than RED (18-38% and 0.77-1.2 W/m(2)). The superior efficiency is attributed to the ability of PRO membranes to more effectively utilize the salinity difference to drive water permeation and better suppress the detrimental leakage of salts. On the other hand, the low conductivity of currently available ion exchange membranes impedes RED ion flux and, thus, constrains the power density. Both technologies exhibit a trade-off between efficiency and power density: employing more permeable but less selective membranes can enhance the power density, but undesired entropy production due to uncontrolled mixing increases and some efficiency is sacrificed. When the concentration difference is increased (i.e., natural → anthropogenic → engineered salinity gradients), PRO osmotic pressure difference rises proportionally but not so for RED Nernst potential, which has logarithmic dependence on the solution concentration. Because of this inherently different characteristic, RED is unable to take advantage of larger salinity gradients, whereas PRO power density is considerably enhanced. Additionally, high solution concentrations suppress the Donnan exclusion effect of the

  3. Integrated core-SOL simulations of L-mode plasma in ITER and Indian demo

    International Nuclear Information System (INIS)

    Wisitsorasak, Apiwat; Onjun, Thawatchai; Kanjanaput, Wittawat

    2015-01-01

    Core-SOL simulations are carried out using 1.5D BALDUR integrated predictive modeling code to investigate tokamak plasma in ITER and Indian DEMO reactors operating in low confinement mode (L-Mode). In each simulation, the plasma current, temperature, and density profiles in both core and SOL region are evolved self-consistency. The SOL is simulated by integrating the fluid equations, including sources, along the field lines. The solutions in SOL subsequently provide as the boundary conditions of core plasma region on low-confinement mode. The core plasma transport model is described using a combination of anomalous transport by Multi-Mode-Model version 2001 (MMM2001) and neoclassical transport calculated by NCLASS module together with the toroidal velocity based on the torque due to Neoclassical Toroidal Viscosity (NTV). In addition, a sensitivity analysis is explored by varying plasma parameters, such as plasma density and auxiliary heating power. Furthermore, the ignition tests are conducted to observed plasma response in each design after shutting down an auxiliary heating. (author)

  4. Influence of core NA on thermal-induced mode instabilities in high power fiber amplifiers

    International Nuclear Information System (INIS)

    Tao, Rumao; Ma, Pengfei; Wang, Xiaolin; Zhou, Pu; Liu, Zejin

    2015-01-01

    We report on the influence of core NA on thermal-induced mode instabilities (MI) in high power fiber amplifiers. The influence of core NA and the V-parameter on MI has been investigated numerically. It shows that core NA has a larger influence on MI for fibers with a smaller core-cladding-ratio, and the influence of core NA on the threshold is more obvious when the amplifiers are pumped at 915 nm. The dependence of the threshold on the V-parameter revealed that the threshold increases linearly as the V-parameter decreases when the V-parameter is larger than 3.5, and the threshold shows an exponential increase as the V-parameter decreases when the V-parameter is less than 3.5. We also discussed the effect of linewidth on MI, which indicates that the influence of linewidth can be neglected for a linewidth smaller than 1 nm when the fiber core NA is smaller than 0.07 and the fiber length is shorter than 20 m. Fiber amplifiers with different core NA were experimentally analyzed, which agreed with the theoretical predictions. (letter)

  5. Thermohydraulic behaviour and heat transfer in the molten core

    International Nuclear Information System (INIS)

    Reineke, H.H.

    1977-01-01

    Increasing the application of nuclear reactors to produce electrical power extremely unprobable accidents should be investigated too. In the Federal Republic of Germany, a research program is performed for some years engaged in accidents at light water reactors in which the melting of the reactor core is presumed. A part of this program is to investigate the thermohydraulic and the heat transfer behavior in an accumulation of molten core material. The knowledge of these events is necessary to analyse the accident exactly. Further on the results of this work are of great importance to build a catcher for the molten core material. As a result of the decay heat the molten material is heated up and the density differences induce a free convection motion. In this work the thermohydraulic behavior and the distribution of the escaping heat fluxes for several accumulations of molten core material were determined. The numerical methods for solving the system of partial differential equation were used to develop computer codes, able to compute the average and local heat fluxes at the walls enclosing the molten core material and the inside increase of the temperature. The numerical computations were confirmed and verified by experimental investigations. In these investigations the molten core material was always assumed as a homogeneous fluid. In this case, the results could be reproduced by simple power laws

  6. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M J; Oyinloye, J O; Chambers, I [Electrowatt Consulting Engineers and Scientists, Warrington, Cheshire (United Kingdom); Scott, C K [Atlantic Nuclear Services, Fredericton, NB (Canada); Omar, A M [Atomic Energy Control Board, Ottawa, ON (Canada)

    1991-12-31

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs.

  7. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    International Nuclear Information System (INIS)

    Lewis, M.J.; Oyinloye, J.O.; Chambers, I.; Scott, C.K.; Omar, A.M.

    1990-01-01

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs

  8. Design study of nuclear power systems for deep space explorers. (2) Electricity supply capabilities of solid cores

    International Nuclear Information System (INIS)

    Yamaji, Akifumi; Takizuka, Takakazu; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime

    2009-01-01

    This study has been carried out in series with the other study, 'Criticality of Low Enriched Uranium Fueled Core' to explore the possibilities of a solid reactor electricity generation system for supplying propulsion power of a deep space explorer. The design ranges of two different systems are determined with respect to the electric power, the radiator mass, and the operating temperatures of the heat-pipes and thermoelectric converters. The two systems are the core surface cooling with heat-pipe system (CSHP), and the core direct cooling with heat-pipe system (CDHP). The evaluated electric powers widely cover the 1 to 100 kW range, which had long been claimed to be the range that lacked the power sources in space. Therefore, the concepts shown by this study may lead to a breakthrough of the human activities in space. The working temperature ranges of the main components, namely the heat-pipes and thermoelectric converters, are wide and covers down to relatively low temperatures. This is desirable from the viewpoints of broadening the choices, reducing the development needs, and improving the reliabilities of the devices. Hence, it is advantageous for an early establishment of the concept. (author)

  9. Enhanced energy density and thermostability in polyimide nanocomposites containing core-shell structured BaTiO3@SiO2 nanofibers

    Science.gov (United States)

    Wang, Junchuan; Long, Yunchen; Sun, Ying; Zhang, Xueqin; Yang, Hong; Lin, Baoping

    2017-12-01

    High energy density polymer nanocomposites with high-temperature resistance are quite desirable for film capacitors and many other power electronics. In this study, polyimide-based (PI) nanocomposite films containing the core-shell structured barium titanate@silicon dioxide (BT@SiO2) nanofibers have been successfully synthesized by the solution casting method. In the BT@SiO2/PI nanocomposite films, the dielectric permittivity as well as the breakdown strength increase significantly. The SiO2 shell layers with moderate dielectric permittivity could effectively mitigate the local field concentration induced by the large mismatch between the dielectric permittivity of BT and PI, which contributes to the enhancement of the breakdown strength of the PI nanocomposite films. As a result, the PI nanocomposite film filled with 3 vol% BT@SiO2 nanofibers exhibits a maximal energy density of 2.31 J cm-3 under the field of 346 kV/mm, which is 62% over the pristine PI (1.42 J cm-3 at 308 kV/mm) and about 200% greater than the best commercial polymer, i.e. biaxially oriented polypropylenes (BOPP) (≈1.2 J cm-3). The thermogravimetric analysis results indicate that the BT@SiO2/PI nanocomposite films have good thermal stability below 500 °C.

  10. Optimized design of a high-power-density PM-assisted synchronous reluctance machine with ferrite magnets for electric vehicles

    Directory of Open Access Journals (Sweden)

    Liu Xiping

    2017-06-01

    Full Text Available This paper proposes a permanent magnet (PM-assisted synchronous reluctance machine (PMASynRM using ferrite magnets with the same power density as rareearth PM synchronous motors employed in Toyota Prius 2010. A suitable rotor structure for high torque density and high power density is discussed with respect to the demagnetization of ferrite magnets, mechanical strength and torque ripple. Some electromagnetic characteristics including torque, output power, loss and efficiency are calculated by 2-D finite element analysis (FEA. The analysis results show that a high power density and high efficiency of PMASynRM are obtained by using ferrite magnets.

  11. Design and Measurement of Planar Toroidal Transformers for Very High Frequency Power Applications

    DEFF Research Database (Denmark)

    Knott, Arnold; Pejtersen, Jens

    2012-01-01

    The quest for higher power density has led to research of very high frequency (30-300 MHz) power converters. Magnetic components based on ferrite cores have limited application within this frequency range due to increased core loss. Air-core magnetics is a viable alternative as they do not exhibit...... core loss. The drawback of most air-core magnetics is that the magnetic field is not contained within a closed shape, and it is thus prone to cause electro magnetic interference. A toroidal air-core inductor configuration can be used to contain the magnetic field. This work presents a novel air......-core toroidal transformer configuration for use in very high frequency power conversion applications. Two prototype transformers (10:10 and 12:12) have been implemented using conventional four layer printed circuit board technology. The transformers have been characterized by two port Z-parameters, which have...

  12. Sensitivity Analysis of Core Damage by Loss of Auxiliary Feed Water during the Extended Loss of All AC Power

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Woo Jae; Chung, Soon Il; Hwang, Su Hyun; Lee, Kyung Jin; Lee, Byung Chul [FNC Tech., Yongin (Korea, Republic of); Yun, Duk Joo; Lee, Seung Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the reactor core damage time for OPR1000 type Nuclear Power Plant (NPP) was analyzed to develop a strategy to handle ELAP and to apply to the EOP. The reactor core damage time in the ELAP condition was calculated according to the time of Auxiliary Feedwater (AFW) loss. Fukushima accident was caused by long hours of Station Black Out (SBO) caused by natural disaster beyond Design Based Accident (DBA) criteria. It led to the reactor core damage. After the accident, the regulatory authorities of each country (Japan, US, EU, IAEA, and etc.) recommended developing the necessary systems and strategies in order to cover up the Extended Loss of All AC Power (ELAP) such as one occurred in the Fukushima accident. And the need of procedure or guideline to cope with ELAP has been raised through the stress test for Wolsong Nuclear Power Plant unit 1. Current Emergency Operating Procedures (EOP) used in domestic nuclear power plant are seemed to be insufficient to cope with ELAP. Therefore, it has been required to be improved. As the result, the time of AFW loss in the ELAP condition influences greatly on core damage time.

  13. Optimization of power and energy densities in supercapacitors

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, David B. [Sandia National Laboratories, PO Box 969 MS 9291, Livermore, CA 94551 (United States)

    2010-06-01

    Supercapacitors use nanoporous electrodes to store large amounts of charge on their high surface areas, and use the ions in electrolytes to carry charge into the pores. Their high power density makes them a potentially useful complement to batteries. However, ion transport through long, narrow channels still limits power and efficiency in these devices. Proper design can mitigate this. Current collector geometry must also be considered once this is done. Here, De Levie's model for porous electrodes is applied to quantitatively predict device performance and to propose optimal device designs for given specifications. Effects unique to nanoscale pores are considered, including that pores may not have enough salt to fully charge. Supercapacitors are of value for electric vehicles, portable electronics, and power conditioning in electrical grids with distributed renewable sources, and that value will increase as new device fabrication methods are developed and proper design accommodates those improvements. Example design outlines for vehicle applications are proposed and compared. (author)

  14. Calculation of core axial power shapes using alternating conditional expectation algorithm

    International Nuclear Information System (INIS)

    Lee, Eun Ki; Kim, Yong Hee; Cha, Kune Ho; Park, Moon Kyu

    1998-01-01

    We have introduced the alternating conditional expectation (ACE) algorithm in the method of reconstructing 20 node axial power shapes from five level detector powers. The ACE algorithm was used to find the optimal relationships between each plane power and normalized five detector powers. The obtained all optimal transformations had simple forms to be represented with polynomials. The reference axial power shapes and simulated detector powers were drawn out of the 3-dimensional results of Reactor Operation and Control Simulation (ROCS) code for various core states. By the ACE algorithm, we obtained the optimal relationship between dependent variable plane power, y, and independent variable detector powers, {Di, i=1,...,5 without any preprocessing, where a total of ≅3490 data sets per each cycle of YongGwang Nuclear (YGN) Power Plant units 3 and 4 are used. To test the validity and accuracy of the new method, about 21,200 cases of reconstructed axial power shapes are compared to original ROCS axial power shapes, and they are also contrasted with those obtained by Fourier fitting method (FFM). The average error of root mean square (rms), axial peak (DFZ), and axial shape index (DASI) of our new method for total 21204 data cases are 0.81%, 0.51% and 0.00204, while FFM 2.29%, 2.37% and 0.00264, respectively. The evaluation results for the data sets not used in the ACE transformations also show that the accuracy of new method is much better than that of FFM

  15. High-energy-density physics researches based on pulse power technology

    International Nuclear Information System (INIS)

    Horioka, Kazuhiko; Nakajima, Mitsuo; Kawamura, Tohru; Sasaki, Toru; Kondo, Kotaro; Yano, Yuuri

    2006-01-01

    Plasmas driven by pulse power device are of interest, concerning the researches on high-energy-density (HED) physics. Dense plasmas are produced using pulse power driven exploding discharges in water. Experimental results show that the wire plasma is tamped and stabilized by the surrounding water and it evolves through a strongly coupled plasma state. A shock-wave-heated, high temperature plasma is produced in a compact pulse power device. Experimental results show that strong shock waves can be produced in the device. In particular, at low initial pressure condition, the shock Mach number reaches 250 and this indicates that the shock heated region is dominated by radiation processes. (author)

  16. Toward power scaling in an acetylene mid-infrared hollow-core optical fiber gas laser: effects of pressure, fiber length, and pump power

    Science.gov (United States)

    Weerasinghe, H. W. Kushan; Dadashzadeh, Neda; Thirugnanasambandam, Manasadevi P.; Debord, Benoît.; Chafer, Matthieu; Gérôme, Frédéric; Benabid, Fetah; Corwin, Kristan L.; Washburn, Brian R.

    2018-02-01

    The effect of gas pressure, fiber length, and optical pump power on an acetylene mid-infrared hollow-core optical fiber gas laser (HOFGLAS) is experimentally determined in order to scale the laser to higher powers. The absorbed optical power and threshold power are measured for different pressures providing an optimum pressure for a given fiber length. We observe a linear dependence of both absorbed pump energy and lasing threshold for the acetylene HOFGLAS, while maintaining a good mode quality with an M-squared of 1.15. The threshold and mode behavior are encouraging for scaling to higher pressures and pump powers.

  17. Introduction of virtual detectors for core monitoring system of korean standard nuclear power plant

    International Nuclear Information System (INIS)

    Eun, Ki Lee.; Yong, Hee Kim.; Jybe, Ho Cha.; Moon, Ghu Park.

    2000-01-01

    A novel algorithm known as the virtual detector method (VDM) is introduced to reconstruct the axial power shape (APS) for the on-line core monitoring system of the Korean Standard Nuclear Power Plant (KSNP). A pure statistical method (SM) is also introduced and the results are compared with the currently implemented five-mode Fourier fitting method (FFM). VDM adopts nine virtual detector informations coupled with a regression model based on the Alternating Conditional Expectation (ACE) algorithm. VDM uses Fourier fitting with the information of nine virtual detectors expanded from the currently implemented FFM, which uses five-level detector information. By introducing virtual detectors, we can increase the number of axial detectors, and thus expect the computational errors of APS to be reduced. The two methods (SM and VDM) are applied to in-core mapping data from six cycles of Yong Gwang nuclear power plant Units 3 and 4. For ∼ 3500 cases of APSs extracted from a cycle of operation which is simulated by a three-dimensional nodal code, the accuracy of the three methods (SM, VDM, FFM) is compared. The average root mean square (RMS) error and average of axial peaking error of SM and VDM resulted in reduction of more than 50 % and 70 %, respectively, relative to FFM. VDM and SM also show more realistic axial profiles and predict more accurate axial peaking than FFM. These improvements can contribute to a larger thermal margin. SM shows the most accurate results for all cases. VDM can almost obtain the same results as SM, and using far fewer computation steps. VDM can be a useful tool for precisely reconstructing axial power shapes in a core monitoring system. (authors)

  18. ORR core re-configuration measurements to increase the fast neutron flux in the Magnetic Fusion Energy (MFE) experiments

    International Nuclear Information System (INIS)

    Hobbs, R.W.; Stinnett, R.M.; Sims, T.M.

    1985-06-01

    A study has been made of the relative increases obtainable in the fast neutron flux in the Magnetic Fusion Energy (MFE) experiment positions by reconfiguring the current ORR core. The study was made at the request of the MFE program to examine the percentage increase possible in the current displacement per atom (dpa) rate (assumed proportional to the fast flux). The principle methods investigated to increase the fast flux consisted of reducing the current core size (number of fuel elements) to increase the core average power density and arrangement of the fuel elements in the reduced-size core to tilt the core power distribution towards the MFE positions. The study concluded that fast fluxes in the E-3 core position could be increased by approximately 15 to 20% over current values and in E-5 by approximately 45 to 55%

  19. High Power Density Motors

    Science.gov (United States)

    Kascak, Daniel J.

    2004-01-01

    With the growing concerns of global warming, the need for pollution-free vehicles is ever increasing. Pollution-free flight is one of NASA's goals for the 21" Century. , One method of approaching that goal is hydrogen-fueled aircraft that use fuel cells or turbo- generators to develop electric power that can drive electric motors that turn the aircraft's propulsive fans or propellers. Hydrogen fuel would likely be carried as a liquid, stored in tanks at its boiling point of 20.5 K (-422.5 F). Conventional electric motors, however, are far too heavy (for a given horsepower) to use on aircraft. Fortunately the liquid hydrogen fuel can provide essentially free refrigeration that can be used to cool the windings of motors before the hydrogen is used for fuel. Either High Temperature Superconductors (HTS) or high purity metals such as copper or aluminum may be used in the motor windings. Superconductors have essentially zero electrical resistance to steady current. The electrical resistance of high purity aluminum or copper near liquid hydrogen temperature can be l/lOO* or less of the room temperature resistance. These conductors could provide higher motor efficiency than normal room-temperature motors achieve. But much more importantly, these conductors can carry ten to a hundred times more current than copper conductors do in normal motors operating at room temperature. This is a consequence of the low electrical resistance and of good heat transfer coefficients in boiling LH2. Thus the conductors can produce higher magnetic field strengths and consequently higher motor torque and power. Designs, analysis and actual cryogenic motor tests show that such cryogenic motors could produce three or more times as much power per unit weight as turbine engines can, whereas conventional motors produce only 1/5 as much power per weight as turbine engines. This summer work has been done with Litz wire to maximize the current density. The current is limited by the amount of heat it

  20. Nominal power density analysis of thermoelectric pins with non-constant cross sections

    International Nuclear Information System (INIS)

    Shi, Yaoguang; Mei, Deqing; Yao, Zhehe; Wang, Yancheng; Liu, Haiyan; Chen, Zichen

    2015-01-01

    Highlights: • Nominal power density of TEGs with non-constant cross sections pins is analyzed. • An analytical model of nominal power density (NPD) is developed. • Influences of shape parameter on NPD for different geometric pins are investigated. • Effects of dimensionless efficiency and the temperature ratio on NPD are examined. - Abstract: The investigation of the geometric structure of TEG (thermoelectric generator) pins is essential, as their geometry determines the performance of devices. In this study, nominal power density (NPD) is used to find a better geometric structure of thermoelectric pins of TEGs, since a comparison of maximum dimensionless efficiencies for different geometric pins cannot be used to identify the optimum geometry. The influence of shape parameter on NPD for TEG pins in linear, quadratic and exponential cross-sectional functions is studied. The NPD decreases when the shape parameter increases for different geometric pins, while the maximum values of NPD are the same. Then, the effects of dimensionless efficiency and the temperature ratio on the NPD are analyzed. The NPD decreases with the increase in dimensionless efficiency and temperature ratio. Pins with linear variation in cross section have the highest NPD among the three geometries of pins evaluated

  1. Investigation of density limit processes in DIII-D

    International Nuclear Information System (INIS)

    Maingi, R.; Mahdavi, M.A.; Petrie, T.W.

    1999-02-01

    A series of experiments has been conducted in DIII-D to investigate density-limiting processes. The authors have studied divertor detachment and MARFEs on closed field lines and find semi-quantitative agreement with theoretical calculations of onset conditions. They have shown that the critical density for MARFE onset at low edge temperature scales as I p /a 2 , i.e. similar to Greenwald scaling. They have also shown that the scaling of the critical separatrix density with heating power at partial detachment onset agrees with Borass' model. Both of these processes yield high edge density limits for reactors such as ITER. By using divertor pumping and pellet fueling they have avoided these and other processes and accessed densities > 1.5x Greenwald limit scaling with H-mode confinement, demonstrating that the Greenwald limit is not a fundamental limit on the core density

  2. Investigation of density limit processes in DIII-D

    International Nuclear Information System (INIS)

    Maingi, R.; Baylor, L.R.; Jernigan, T.

    2001-01-01

    A series of experiments has been conducted in DIII-D to investigate density-limiting processes. We have studied divertor detachment and MARFEs on closed field lines and find semi-quantitative agreement with theoretical calculations of onset conditions. We have shown that the critical density for MARFE onset at low edge temperature scales as I p /a 2 , i.e. similar to Greenwald scaling. We have also shown that the scaling of the critical separatrix density with heating power at partial detachment onset agrees with Borass' model. Both of these processes yield high edge density limits for reactors such as ITER. By using divertor pumping and pellet fueling we have avoided these and other processes and accessed densities >1.5x Greenwald limit scaling with H-mode confinement, demonstrating that the Greenwald limit is not a fundamental limit on the core density. (author)

  3. Investigation of density limit processes in DIII-D

    International Nuclear Information System (INIS)

    Maingi, R.; Mahdavi, M.A.; Petrie, T.W.

    1999-01-01

    A series of experiments has been conducted in DIII-D to investigate density-limiting processes. We have studied divertor detachment and MARFEs on closed field lines and find semi-quantitative agreement with theoretical calculations of onset conditions. We have shown that the critical density for MARFE onset at low edge temperature scales as I p /a 2 , i.e. similar to Greenwald scaling. We have also shown that the scaling of the critical separatrix density with heating power at partial detachment onset agrees with Borass' model. Both of these processes yield high edge density limits for reactors such as ITER. By using divertor pumping and pellet fueling we have avoided these and other processes and accessed densities > 1.5x Greenwald limit scaling with H-mode confinement, demonstrating that the Greenwald limit is not a fundamental limit on the core density. (author)

  4. Biowaste-Derived Hierarchical Porous Carbon Nanosheets for Ultrahigh Power Density Supercapacitors.

    Science.gov (United States)

    Yu, Dengfeng; Chen, Chong; Zhao, Gongyuan; Sun, Lei; Du, Baosheng; Zhang, Hong; Li, Zhuo; Sun, Ye; Besenbacher, Flemming; Yu, Miao

    2018-03-05

    Low-cost activated carbons with high capacitive properties remain desirable for supercapacitor applications. Herein, a three-dimensional scaffolding framework of porous carbon nanosheets (PCNSs) has been produced from a typical biowaste, namely, ground cherry calyces, the specific composition and natural structures of which have contributed to the PCNSs having a very large specific surface area of 1612 m 2  g -1 , a hierarchical pore size distribution, a turbostratic carbon structure with a high degree graphitization, and about 10 % oxygen and nitrogen heteroatoms. A high specific capacitance of 350 F g -1 at 0.1 A g -1 has been achieved in a two-electrode system with 6 m KOH; this value is among the highest specific capacitance of biomass-derived carbon materials. More inspiringly, a high energy density of 22.8 Wh kg -1 at a power density of 198.8 W kg -1 can be obtained with 1 m aqueous solution of Li 2 SO 4 , and an ultrahigh energy density of 81.4 Wh kg -1 at a power density of 446.3 W kg -1 is realized with 1-ethyl-3-methylimidazolium tetrafluoroborate electrolyte. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  5. Improving the calculated core stability by the core nuclear design optimization

    International Nuclear Information System (INIS)

    Partanen, P.

    1995-01-01

    Three different equilibrium core loadings for TVO II reactor have been generated in order to improve the core stability properties at uprated power level. The reactor thermal power is assumed to be uprated from 2160 MW th to 2500 MW th , which moves the operating point after a rapid pump rundown where the core stability has been calculated from 1340 MW th and 3200 kg/s to 1675 MW th and 4000 kg/s. The core has been refuelled with ABB Atom Svea-100 -fuel, which has 3,64% w/o U-235 average enrichment in the highly enriched zone. PHOENIX lattice code has been used to provide the homogenized nuclear constants. POLCA4 static core simulator has been used for core loadings and cycle simulations and RAMONA-3B program for simulating the dynamic response to the disturbance for which the stability behaviour has been evaluated. The core decay ratio has been successfully reduced from 0,83 to 0,55 mainly by reducing the power peaking factors. (orig.) (7 figs., 1 tab.)

  6. Self-powered detectors with thulium emitter

    International Nuclear Information System (INIS)

    Haller, P.; Klar, E.

    1978-01-01

    In addition to fission chambers, prompt-indicating self-powered (SPN) detectors are used for measuring the neutron flux density in the core of power reactors. Although current SPN detectors with a cobalt emitter give satisfactora results, detectors with other emitter materials have been analyzed and tested. The author describes the properties and decay pattern of the nuclide thulium and presents the results of measurements made while testing thulium detectors. (orig.) [de

  7. Partial thorium loading in the initial core of Kakrapar atomic power reactor

    International Nuclear Information System (INIS)

    Balakrishnan, M.R.

    1993-01-01

    The first unit of Kakrapar nuclear power station has gone critical with some thorium oxide fuel bundles loaded in its core. The thorium helps to flatten the power by reducing neutron flux in the centre of the reactor. However, the placing of the thorium had to be planned with care, because if the neutron flux at a point where a safety rod is located is depressed, the reactivity worth of the safety rod gets reduced. Using a dynamic programing approach, the Reactor Engineering Division of Bhabha Atomic Research Centre worked out a satisfactory configuration for loading the thorium bundles

  8. Comparison of Battery-Powered and Manual Bone Biopsy Systems for Core Needle Biopsy of Sclerotic Bone Lesions.

    Science.gov (United States)

    Cohen, Micah G; McMahon, Colm J; Kung, Justin W; Wu, Jim S

    2016-05-01

    The purpose of this study was to compare manual and battery-powered bone biopsy systems for diagnostic yield and procedural factors during core needle biopsy of sclerotic bone lesions. A total of 155 consecutive CT-guided core needle biopsies of sclerotic bone lesions were performed at one institution from January 2006 to November 2014. Before March 2012, lesions were biopsied with manual bone drill systems. After March 2012, most biopsies were performed with a battery-powered system and either noncoaxial or coaxial biopsy needles. Diagnostic yield, crush artifact, CT procedure time, procedure radiation dose, conscious sedation dose, and complications were compared between the manual and battery-powered core needle biopsy systems by Fisher exact test and t test. One-way ANOVA was used for subgroup analysis of the two battery-powered systems for procedure time and radiation dose. The diagnostic yield for all sclerotic lesions was 60.0% (93/155) and was significantly higher with the battery-powered system (73.0% [27/37]) than with the manual systems (55.9% [66/118]) (p = 0.047). There was no significant difference between the two systems in terms of crush artifact, procedure time, radiation dose, conscious sedation administered, or complications. In subgroup analysis, the coaxial battery-powered biopsies had shorter procedure times (p = 0.01) and lower radiation doses (p = 0.002) than the coaxial manual systems, but the noncoaxial battery-powered biopsies had longer average procedure times and higher radiation doses than the coaxial manual systems. In biopsy of sclerotic bone lesions, use of a battery-powered bone drill system improves diagnostic yield over use of a manual system.

  9. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y H; Cha, K H; Lee, S H [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  10. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  11. Power peak in vicinity of WWER-440 control assembly

    International Nuclear Information System (INIS)

    Mikus, J.

    2002-01-01

    This paper presents information concerning the WWER-440 local power peaking problem induced by a control assembly and corresponding investigation possibilities on the light-water zero-power reactor LR-O at the Nuclear Research Institute Rez plc. Brief description of the disposable CA model, experimental arrangement and conditions on the LR-O reactor, preparation of the relevant measurements in the WWER-440 type cores with CA model, as well as some preliminary results of the fission density distribution obtained in a core without boron and with fuel assemblies having profiled enrichment are mentioned too (Author)

  12. Comprehensive performance analyses and optimization of the irreversible thermodynamic cycle engines (TCE) under maximum power (MP) and maximum power density (MPD) conditions

    International Nuclear Information System (INIS)

    Gonca, Guven; Sahin, Bahri; Ust, Yasin; Parlak, Adnan

    2015-01-01

    This paper presents comprehensive performance analyses and comparisons for air-standard irreversible thermodynamic cycle engines (TCE) based on the power output, power density, thermal efficiency, maximum dimensionless power output (MP), maximum dimensionless power density (MPD) and maximum thermal efficiency (MEF) criteria. Internal irreversibility of the cycles occurred during the irreversible-adiabatic processes is considered by using isentropic efficiencies of compression and expansion processes. The performances of the cycles are obtained by using engine design parameters such as isentropic temperature ratio of the compression process, pressure ratio, stroke ratio, cut-off ratio, Miller cycle ratio, exhaust temperature ratio, cycle temperature ratio and cycle pressure ratio. The effects of engine design parameters on the maximum and optimal performances are investigated. - Highlights: • Performance analyses are conducted for irreversible thermodynamic cycle engines. • Comprehensive computations are performed. • Maximum and optimum performances of the engines are shown. • The effects of design parameters on performance and power density are examined. • The results obtained may be guidelines to the engine designers

  13. REJUVENATING THE MATTER POWER SPECTRUM: RESTORING INFORMATION WITH A LOGARITHMIC DENSITY MAPPING

    International Nuclear Information System (INIS)

    Neyrinck, Mark C.; Szalay, Alexander S.; Szapudi, Istvan

    2009-01-01

    We find that nonlinearities in the dark matter power spectrum are dramatically smaller if the density field first undergoes a logarithmic mapping. In the Millennium simulation, this procedure gives a power spectrum with a shape hardly departing from the linear power spectrum for k ∼ -1 at all redshifts. Also, this procedure unveils pristine Fisher information on a range of scales reaching a factor of 2-3 smaller than in the standard power spectrum, yielding 10 times more cumulative signal to noise at z = 0.

  14. Neutronics analysis on mini test fuel in the RSG-GAS core

    International Nuclear Information System (INIS)

    Tukiran S; Tagor M Sembiring

    2016-01-01

    Research on UMo fuel for research reactor has been developed. The fuel of research reactor is uranium molybdenum low enrichment with high density. For supporting the development of fuel fabrication, an neutronic analysis of mini fuel plates in the RSG-GAS core was performed. The aim of analysis is to determine the numbers of fuel cycles in the core to know the maximum fuel burn-up. The mini fuel plates of U_7Mo-Al and U_6Zr-Al with densities of 7.0 gU/cc and 5.2 gU/cc, respectively, will be irradiated in the RSG-GAS core. The size of both fuels, namely 630 x 70.75 x 1.30 mm were inserted to the 3 plates of dummy fuel. Before the fuel will be irradiated in the core, a calculation for safety analysis from neutronics and thermal-hydraulics aspects were required. However, in this paper, it will be discussed safety analysis of the U_7Mo-Al and U_6Zr-Al mini fuels from neutronic point of view. The calculation was done using WIMSD-5B and Batan-3DIFF codes. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. If it is compared, the power density of U_7Mo-Al mini fuel is bigger than U_6Zr-Al fuel. (author)

  15. Thulium heat source for high-endurance and high-energy density power systems

    International Nuclear Information System (INIS)

    Walter, C.E.; Kammeraad, J.E.; Van Konynenburg, R.; VanSant, J.H.

    1991-05-01

    We are studying the performance characteristics of radioisotope heat source designs for high-endurance and high-energy-density power systems that use thulium-170. Heat sources in the power range of 5--50 kW th coupled with a power conversion efficiency of ∼30%, can easily satisfy current missions for autonomous underwater vehicles. New naval missions will be possible because thulium isotope power systems have a factor of one-to-two hundred higher endurance and energy density than chemical and electrochemical systems. Thulium-170 also has several other attractive features, including the fact that it decays to stable ytterbium-170 with a half-life of four months. For terrestrial applications, refueling on that time scale should be acceptable in view of the advantage of its benign decay. The heat source designs we are studying account for the requirements of isotope production, shielding, and integration with power conversion components. These requirements are driven by environmental and safety considerations. Thulium is present in the form of thin refractory thulia disks that allow power conversion at high peak temperature. We give estimates of power system state points, performance, mass, and volume characteristics. Monte Carlo radiation analysis provides a detailed assessment of shield requirements and heat transfer under normal and distressed conditions is also considered. 11 refs., 7 figs., 4 tabs

  16. Evaluation of in-core measurements with self-powered detectors

    International Nuclear Information System (INIS)

    Spoden, E.

    1977-01-01

    A calculational method is presented for transformation of n-β currents into relative thermal neutron fluxes taking into account an epithermal correction. The difference between epithermal corrections with cross section ratios is demonstrated based on calculations with NR approximation and multiplied by the ratio of mean epithermal and thermal first collision probabilities as well as on calculations using the model of infinite dilution. The effects of this difference on axial power density distribution, relative power of the fuel assemblies, and the inhomogeneity coefficients are discussed in the light of a characteristic measurement with rhodium emitter in a WWER-type reactor. The effect of errors in the nuclear data on the mean error of relative thermal neutron flux is evaluated. (author)

  17. Noise variation by compressive stress on the model core of power transformers

    Energy Technology Data Exchange (ETDEWEB)

    Mizokami, Masato, E-mail: mizokami.g76.masato@jp.nssmc.com; Kurosaki, Yousuke

    2015-05-01

    The reduction of audible noise generated by cores for power transformers has been required due to environmental concern. It is known that compressive stress in the rolling direction of electrical steel affects magnetostriction and it can result in an increase in noise level. In this research, the effect of compressive stress to noise was investigated on a 3-phase 3-limb model core. Compressive stress was applied in the rolling direction of the limbs from the outside of the core. It increased the sound pressure levels and the slope of the rise was about 2 dBA/MPa. Magnetostriction on single sheet samples was also measured under compressive stress and the harmonic components of the magnetostriction were compared with those of noise. It revealed that the variation in magnetostriction with compressive stress did not entirely correspond to that in noise. In one of the experiments, localized bending happened on one limb during compressing the core. While deformation of the core had not been intended, the noise was measured. The deformation increased the noise by more than 10 dBA and it occurred on most of the harmonic components. - Highlights: • Audible noise was measured on a model core to which compressive stress was applied. • The stress in the rolling direction of the steel causes a rise in noise level. • The slope of the rise in sound pressure level up to 2.5 MPa is about 2 dBA/MPa. • Variation in magnetostriction by stress does not entirely agree with that in noise. • Bend arisen in the core causes an extreme increase in noise.

  18. Noise variation by compressive stress on the model core of power transformers

    International Nuclear Information System (INIS)

    Mizokami, Masato; Kurosaki, Yousuke

    2015-01-01

    The reduction of audible noise generated by cores for power transformers has been required due to environmental concern. It is known that compressive stress in the rolling direction of electrical steel affects magnetostriction and it can result in an increase in noise level. In this research, the effect of compressive stress to noise was investigated on a 3-phase 3-limb model core. Compressive stress was applied in the rolling direction of the limbs from the outside of the core. It increased the sound pressure levels and the slope of the rise was about 2 dBA/MPa. Magnetostriction on single sheet samples was also measured under compressive stress and the harmonic components of the magnetostriction were compared with those of noise. It revealed that the variation in magnetostriction with compressive stress did not entirely correspond to that in noise. In one of the experiments, localized bending happened on one limb during compressing the core. While deformation of the core had not been intended, the noise was measured. The deformation increased the noise by more than 10 dBA and it occurred on most of the harmonic components. - Highlights: • Audible noise was measured on a model core to which compressive stress was applied. • The stress in the rolling direction of the steel causes a rise in noise level. • The slope of the rise in sound pressure level up to 2.5 MPa is about 2 dBA/MPa. • Variation in magnetostriction by stress does not entirely agree with that in noise. • Bend arisen in the core causes an extreme increase in noise

  19. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  20. High power density reactors based on direct cooled particle beds

    International Nuclear Information System (INIS)

    Powell, J.R.; Horn, F.L.

    1985-01-01

    Reactors based on direct cooled HTGR type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out long the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBR's) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed. 12 figs

  1. MA-core loaded untuned RF compression cavity for HIRFL-CSR

    International Nuclear Information System (INIS)

    Mei Lirong; Xu Zhe; Yuan Youjin; Jin Peng; Bian Zhibin; Zhao Hongwei; Xia Jiawen

    2012-01-01

    To meet the requirements of high energy density physics and plasma physics research at HIRFL-CSR the goal of achieving a higher accelerating gap voltage was proposed. Therefore, a magnetic alloy (MA)-core loaded radio frequency (RF) cavity that can provide a higher accelerating gap voltage compared to standard ferrite loaded cavities has been studied at IMP. In order to select the proper magnetic alloy material to load the RF compression cavity, measurements of four different kinds of sample MA-cores have been carried out. By testing the small cores, the core composition was selected to obtain the desired performance. According to the theoretical calculation and simulation, which show reasonable consistency for the MA-core loaded cavity, the desired performance can be achieved. Finally about 1000 kW power will be needed to meet the requirements of 50 kV accelerating gap voltage by calculation.

  2. Durability of Low Platinum Fuel Cells Operating at High Power Density

    Energy Technology Data Exchange (ETDEWEB)

    Polevaya, Olga [Nuvera Fuel Cells Inc.; Blanchet, Scott [Nuvera Fuel Cells Inc.; Ahluwalia, Rajesh [Argonne National Lab; Borup, Rod [Los-Alamos National Lab; Mukundan, Rangachary [Los-Alamos National Lab

    2014-03-19

    Understanding and improving the durability of cost-competitive fuel cell stacks is imperative to successful deployment of the technology. Stacks will need to operate well beyond today’s state-of-the-art rated power density with very low platinum loading in order to achieve the cost targets set forth by DOE ($15/kW) and ultimately be competitive with incumbent technologies. An accelerated cost-reduction path presented by Nuvera focused on substantially increasing power density to address non-PGM material costs as well as platinum. The study developed a practical understanding of the degradation mechanisms impacting durability of fuel cells with low platinum loading (≤0.2mg/cm2) operating at high power density (≥1.0W/cm2) and worked out approaches for improving the durability of low-loaded, high-power stack designs. Of specific interest is the impact of combining low platinum loading with high power density operation, as this offers the best chance of achieving long-term cost targets. A design-of-experiments approach was utilized to reveal and quantify the sensitivity of durability-critical material properties to high current density at two levels of platinum loading (the more conventional 0.45 mgPt.cm–1 and the much lower 0.2 mgPt.cm–2) across several cell architectures. We studied the relevance of selected component accelerated stress tests (AST) to fuel cell operation in power producing mode. New stress tests (NST) were designed to investigate the sensitivity to the addition of electrical current on the ASTs, along with combined humidity and load cycles and, eventually, relate to the combined city/highway drive cycle. Changes in the cathode electrochemical surface area (ECSA) and average oxygen partial pressure on the catalyst layer with aging under AST and NST protocols were compared based on the number of completed cycles. Studies showed elevated sensitivity of Pt growth to the potential limits and the initial particle size distribution. The ECSA loss

  3. A combined compensation method for the output voltage of an insulated core transformer power supply

    Energy Technology Data Exchange (ETDEWEB)

    Yang, L.; Yang, J., E-mail: jyang@mail.hust.edu.cn; Liu, K. F.; Qin, B.; Chen, D. Z. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2014-06-15

    An insulated core transformer (ICT) power supply is an ideal high-voltage generator for irradiation accelerators with energy lower than 3 MeV. However, there is a significant problem that the structure of the segmented cores leads to an increase in the leakage flux and voltage differences between rectifier disks. A high level of consistency in the output of the disks helps to achieve a compact structure by improving the utilization of both the rectifier components and the insulation distances, and consequently increase the output voltage of the power supply. The output voltages of the disks which are far away from the primary coils need to be improved to reduce their inhomogeneity. In this study, by investigating and comparing the existing compensation methods, a new combined compensation method is proposed, which increases the turns on the secondary coils and employs parallel capacitors to improve the consistency of the disks, while covering the entire operating range of the power supply. This method turns out to be both feasible and effective during the development of an ICT power supply. The non-uniformity of the output voltages of the disks is less than 3.5% from no-load to full-load, and the power supply reaches an output specification of 350 kV/60 mA.

  4. The Physical Mechanism of Core-Wide and Local Instabilities at the Forsmark-1 BWR

    International Nuclear Information System (INIS)

    Analytis, G. Th.

    1998-10-01

    During the last 15 years, the problem of BWR instabilities has attracted the attention of a number of researchers. From the theoretical point of view, one would be interested in physically understanding the mechanisms responsible for the in- and out-of-phase core wide power oscillations observed at certain operating points of the power-flow map in different BWRs. From the practical point of view, one must try to avoid these 'incidents' since either locally, or globally, the power may substantially exceed the prescribed levels. In this work, we shall use RAMONA3-12 and analyse a rather unusual instability incident at Forsmark-1 in which in addition to the core-wide fundamental spatial mode oscillation, there were local large amplitude power oscillations at different radial positions in the core. We were able to reproduce these unusual experimental findings by assuming that there are large amplitude Density Wave Oscillations (DWOs) in different bundles, induced by the fact that these bundles were not seated properly into the lower fuel support plate. (author)

  5. Fuel assembly outlet temperature profile influence on core by-pass flow and power distribution determination in WWER -440 reactors

    International Nuclear Information System (INIS)

    Petenyi, V.; Klucarova, K.; Remis, J.

    2003-01-01

    The in core instrumentation of the WWER-440 reactors consists of the thermocouple system and the system of self powered detectors (SPD). The thermocouple systems are positioned about 50 cm above the fuel bundle upper flow-mixing grid. The usual assumption is that, the coolant is well mixed in the Tc location, i.e. the temperature is constant through the flow cross-section area. The present evaluations by using the FLUENT 5.5.14 code reveal that, this assumption is not fulfilled. There exists a temperature profile that depends on fuel assembly geometry and on inner power profile of the fuel assembly. The paper presents the estimation of this effect and its influence on the core power distribution and the core by-pass flow determination. Comparison with measurements in Mochovce NPP will also be a part of this presentation (Authors)

  6. Dual-core Itanium Processor

    CERN Multimedia

    2006-01-01

    Intel’s first dual-core Itanium processor, code-named "Montecito" is a major release of Intel's Itanium 2 Processor Family, which implements the Intel Itanium architecture on a dual-core processor with two cores per die (integrated circuit). Itanium 2 is much more powerful than its predecessor. It has lower power consumption and thermal dissipation.

  7. Low density, microcellular, dopable, agar/gelatin foams for pulsed power experiments

    Energy Technology Data Exchange (ETDEWEB)

    McNamara, W.F. [Orion International Technologies, Inc., Albuquerque, NM (United States); Aubert, J.H. [Sandia National Lab., Albuquerque, NM (United States)

    1997-04-01

    Low-density, microcellular foams prepared from the natural polymers agar and gelatin have been developed for pulsed-power physics experiments. Numerous experiments were supported with foams having densities at or below 10 mg/cm{sup 3}. For some of the experiments, the agar/gelatin foam was uniformly doped with metallic elements using soluble salts. Depending on the method of preparation, cell sizes were typically below 10 microns and for one process were below 1.0 micron.

  8. Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-07-01

    Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (Dpower is 300 MWth. First calculation, we calculate survey parameter for UN-PuN fuel with fissile contain from Plutonium waste LWR for each geometry. The minimum power density is around 72 Watt/cc, and maximum power density 114 Watt/cc. After we calculate with various geometry core, when we use the balance geometry, the k-eff value flattest and more stable than the others.

  9. Flexible two-ply yarn supercapacitors based on carbon nanotube/stainless steel core spun yarns decorated with Co3O4 nanoparticles and MnOx composites

    International Nuclear Information System (INIS)

    Su, Fenghua; Lyu, Xiaoming; Liu, Cansen; Miao, Menghe

    2016-01-01

    Highlights: • Carbon nanotube/stainless steel core-sheath yarns as electrode for two-ply supercapacitor. • The yarns were decorated with Co 3 O 4 nanoparticles and MnO x composites. • Yarn electrodes deposited on Co 3 O 4 and MnO x have excellent specific capacitance. • The two-ply yarn supercapacitor based on MnO x has high energy density and power density. • The yarn supercapacitors are highly flexible and strong for wearable electronics. - Abstract: High performance two-ply yarn supercapacitors are fabricated by electrodeposition of transition metal oxide pseudocapacitive materials on carbon nanotube/stainless steel (CNT/SS) core spun yarns. The SS core incorporated in the carbon nanotube yarn has dramatically improved the conductivity of the substrate and the efficiency of electrodeposition of metal oxides on the electrodes. The SS core acts as current collector in the final two-ply yarn supercapacitors with significantly improved specific capacitance and up-scaled length of supercapacitors. Manganese oxide and cobalt oxide electrodeposited on as-spun CNT/SS yarn are compared for their electrochemical performance in two-ply yarn supercapacitors. Both supercapacitors displayed improvement in capacitance, energy and power densities, especially MnO x /CNT/SS yarn supercapacitor. The solid-state supercapacitor based on MnO x /CNT/SS composite yarn shows excellent electrochemical properties with a specific capacitance of 217.61 F/cm 3 at 0.02 V/s and an energy density of 4.84 mWh/cm 3 at a power density of 435.22 mW/cm 3 .

  10. Status of the Munich compact core reactor project

    International Nuclear Information System (INIS)

    Boening, K.; Glaeser, W.; Meier, J.; Rau, G.; Roehrmoser, A.; Steichele, E.

    1991-01-01

    A new national, high performance research reactor is being planned in Munich which shall provide an unperturbed thermal neutron flux of about 8.10 14 cm -2 s -1 at only 20 MW power. This object can be achieved with the concept of a small, light water cooled 'compact core' situated in the center of a large heavy water moderator tank. In the first part of this paper an overview is given on the status of the necessary fuel enrichment is discussed. In becomes clear that the potential of the compact core concept can only be developed without compromises when using highly enriched uranium in combination with the new high density fuel. (orig.)

  11. EXCURS: a computing programme for analysis of core transient behaviour in a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Saito, Shinzo

    1977-09-01

    In the code EXCURS developed for core transient behaviour calculation of a sodium-cooled fast reactor, a one-channel model is used to represent thermal behaviour of the reactor core. Calculations are made for three different channels; i.e. average, hot and hottest. In the average channel the power density and coolant velocity are equal to the mean values of the whole core. In the hot channel, a maximum power density of the core and a specific coolant velocity are introduced. In the hottest channel, engineering hot channel factors are considered to the hot channel. A one-point neutron kinetics equation with six delayed neutron groups is used to calculate the time-dependent power behaviour. Externally introduced reactivity effect and control rod movement in the case of a scram are taken into account. In the feedback effects evaluated on the basis of the average channel temperatures are considered Doppler effect, fuel axial expansion, cladding expansion, coolant expansion and structure expansion. The decay heat after reactor scram is also considered. Heat balance is taken in each cross section, neglecting the axial heat transfer except for the coolant region. Temperature dependence of the physical properties of materials is considered by second-order polynomials approximation, and also the fuel melting process. Each channel can be divided into a maximum of 20 regions in both radially and axially. The reactor core transient behaviour due to reactivity insertion or loss-of-coolant flow can be studied by EXCURS. The calculated results are plotted optionally by connected code EXPLOT. (auth.)

  12. Conceptual study of axial offset fluctuations upon stepwise power changes in a thorium–plutonium core to improve load-following conditions

    International Nuclear Information System (INIS)

    Lau, Cheuk Wah; Dykin, Victor; Nylén, Henrik; Björk, Klara Insulander; Sandberg, Urban

    2014-01-01

    Highlights: • Thorium–plutonium mixed oxide to improve nuclear reactors load-following capability. • SIMULATE-3 was the main calculation tool. • The Ringhals-3 PWR unit in Sweden was used as a reference. • Lower xenon poisoning and shorter reactor dead time. - Abstract: The increased share of renewable energy, such as wind and solar power, will increase the demand for load-following power sources, and nuclear reactors could be one option. However, during rapid load-following events, traditional UOX cores could be restricted by the volatile oscillation of the power distribution. Therefore, a conceptual study on stability properties of Th-MOX PWR concerning axial offset power excursion during load-following events are investigated and discussed. The study is performed in SIMULATE-3 for a realistic PWR core (Ringhals-3) at the end of cycle, where the largest amplitude of the axial offset oscillations is expected. It is shown that the Th-MOX core possesses much better stability characteristics and shorter reactor dead time compared with a traditional UOX core, and the main reasons are the lower sensitivity to perturbations in the neutron spectrum, lower xenon poisoning and lower thermal neutron flux

  13. LTCC magnetic components for high density power converter

    Science.gov (United States)

    Lebourgeois, Richard; Labouré, Eric; Lembeye, Yves; Ferrieux, Jean-Paul

    2018-04-01

    This paper deals with multilayer magnetic components for power electronics application and specifically for high frequency switching. New formulations based on nickel-zinc-copper spinel ferrites were developed for high power and high frequency applications. These ferrites can be sintered at low temperature (around 900°C) which makes them compatible with the LTCC (Low Temperature Co-fired Ceramics) technology. Metallic parts of silver or gold can be fully integrated inside the ferrite while guaranteeing the integrity of both the ferrite and the metal. To make inductors or transformers with the required properties, it is mandatory to have nonmagnetic parts between the turns of the winding. Then it is essential to find a dielectric material, which can be co-sintered both with the ferrite and the metal. We will present the solution we found to this problem and we will describe the results we obtained for a multilayer co-sintered transformer. We will see that these new components have good performance compared with the state of the art and are very promising for developing high density switching mode power supplies.

  14. WWER core pattern enhancement using adaptive improved harmony search

    International Nuclear Information System (INIS)

    Nazari, T.; Aghaie, M.; Zolfaghari, A.; Minuchehr, A.; Norouzi, A.

    2013-01-01

    Highlights: ► The classical and improved harmony search algorithms are introduced. ► The advantage of IHS is demonstrated in Shekel's Foxholes. ► The CHS and IHS are compared with other Heuristic algorithms. ► The adaptive improved harmony search is applied for two cases. ► Two cases of WWER core are optimized in BOC FA pattern. - Abstract: The efficient operation and fuel management of PWRs are of utmost importance. Core performance analysis constitutes an essential phase in core fuel management optimization. Finding an optimum core arrangement for loading of fuel assemblies, FAs, in a nuclear core is a complex problem. In this paper, application of classical harmony search (HS) and adaptive improved harmony search (IHS) in loading pattern (LP) design, for pressurized water reactors, is described. In this analysis, finding the best core pattern, which attains maximum multiplication factor, k eff , by considering maximum allowable power picking factors (PPF) is the main objective. Therefore a HS based, LP optimization code is prepared and CITATION code which is a neutronic calculation code, applied to obtain effective multiplication factor, neutron fluxes and power density in desired cores. Using adaptive improved harmony search and neutronic code, generated LP optimization code, could be applicable for PWRs core with many numbers of FAs. In this work, at first step, HS and IHS efficiencies are compared with some other heuristic algorithms in Shekel's Foxholes problem and capability of the adaptive improved harmony search is demonstrated. Results show, efficient application of IHS. At second step, two WWER cases are studied and then IHS proffered improved core patterns with regard to mentioned objective functions.

  15. Joint European contribution to phases 1 and 2 of the BN600 hybrid reactor benchmark core analysis

    International Nuclear Information System (INIS)

    Rimpault, Gerald; Newton, Tim; Smith, Peter

    2000-01-01

    This paper describes the ERANOS code developed within the European cooperation on fast reactors. Reference scheme and ERANOS code validation are included. The method for BN-600 reactor core analysis and the results of phases 1 and two are presented. They include effective multiplication factors, fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution; beta and neutron life time; reaction rate distribution

  16. Safety-related investigations on power distribution in MOX fuel elements in LWR cores

    International Nuclear Information System (INIS)

    Kramer, E.; Langenbuch, S.

    1991-01-01

    For the concept of thermal recycling various fuel assembly designs have been developped during the last years. An overview is given describing the present status of MOX-fuel assembly design for PWR and BWR. The local power distribution within the MOX-fuel assembly and influences between neighbouring MOX- and Uranium fuel assemblies have been analyzed by own calculations. These investigations are limited to specific aspects of the spatial power distribution, which are related to the use of MOX-fuel assemblies within the reactor core of LWR. (orig.) [de

  17. Achieving High-Energy-High-Power Density in a Flexible Quasi-Solid-State Sodium Ion Capacitor.

    Science.gov (United States)

    Li, Hongsen; Peng, Lele; Zhu, Yue; Zhang, Xiaogang; Yu, Guihua

    2016-09-14

    Simultaneous integration of high-energy output with high-power delivery is a major challenge for electrochemical energy storage systems, limiting dual fine attributes on a device. We introduce a quasi-solid-state sodium ion capacitor (NIC) based on a battery type urchin-like Na2Ti3O7 anode and a capacitor type peanut shell derived carbon cathode, using a sodium ion conducting gel polymer as electrolyte, achieving high-energy-high-power characteristics in solid state. Energy densities can reach 111.2 Wh kg(-1) at power density of 800 W kg(-1), and 33.2 Wh kg(-1) at power density of 11200 W kg(-1), which are among the best reported state-of-the-art NICs. The designed device also exhibits long-term cycling stability over 3000 cycles with capacity retention ∼86%. Furthermore, we demonstrate the assembly of a highly flexible quasi-solid-state NIC and it shows no obvious capacity loss under different bending conditions.

  18. Core design experience of WWER-440 reactors when they working on increased power level

    International Nuclear Information System (INIS)

    Adeev, V.; Panov, A.; Melenchuk, I.

    2015-01-01

    The Kola NPP continues commercial operation of 2nd generation fuel (FA-2) and trial operation of 3rd generation fuel (FA-3), which has a number of design features providing the best operational characteristics. This report gives the results of VVER-440 core operation with FA-2 and FA-3 with enrichment increased up to 4.87%, and at the power level uprated to 107% of nominal power level. Brief analysis of obtained data is carried out. Peculiarities and techniques of developing loading patterns with new types of nuclear fuel for operation at the uprated power level are reviewed. (authors)

  19. Evaluation of local power distribution with fine-mesh core model for the HTTR

    International Nuclear Information System (INIS)

    Murata, Isao; Yamashita, Kiyonobu; Maruyama, So; Shindo, Ryuichi; Fujimoto, Nozomu; Sudo, Yukio; Nakata, Tetsuo.

    1991-01-01

    An evaluation method of the local power distribution was developed considering the radial and axial heterogeneity caused by fuel rods, BP rods and block end graphite for the High Temperature Engineering Test Reactor (HTTR) in Japan Atomic Energy Research Institute (JAERI). The evaluation method was verified through the analyses of critical assembly experiments. A good agreement was obtained between calculations and measurements and the difference was less than 3 % on the power distribution. This method was applied to the core design for the HTTR to evaluate the maximum fuel temperature. From these results, it was confirmed that this evaluation method has an enough accuracy and is able to predict the detailed power distribution of the HTTR. (author)

  20. Harmonic Domain Modelling of Transformer Core Nonlinearities Using the DIgSILENT PowerFactory Software

    DEFF Research Database (Denmark)

    Bak, Claus Leth; Bak-Jensen, Birgitte; Wiechowski, Wojciech

    2008-01-01

    This paper demonstrates the results of implementation and verification of an already existing algorithm that allows for calculating saturation characteristics of singlephase power transformers. The algorithm was described for the first time in 1993. Now this algorithm has been implemented using...... the DIgSILENT Programming Language (DPL) as an external script in the harmonic domain calculations of a power system analysis tool PowerFactory [10]. The algorithm is verified by harmonic measurements on a single-phase power transformer. A theoretical analysis of the core nonlinearities phenomena...... in single and three-phase transformers is also presented. This analysis leads to the conclusion that the method can be applied for modelling nonlinearities of three-phase autotransformers....

  1. Development of In-Core Protection System

    International Nuclear Information System (INIS)

    Cho, J. H; Kim, C. H.; Kim, J. H.; Jeong, S. H.; Sohn, S. D.; BaeK, S. M.; YOON, J. H.

    2016-01-01

    In-core Protection System (ICOPS) is an on-line digital computer system which continuously calculates Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD) based on plant parameters to make trip decisions based on the computations. The function of the system is the same as that of Core Protection Calculator System (CPCS) and Reactor Core Protection System (RCOPS) which are applied to Optimized Power Reactor 1000 (OPR1000) and Advanced Power Reactor 1400 (APR1400). The ICOPS has been developed to overcome the algorithm related obstacles in overseas project. To achieve this goal, several algorithms were newly developed and hardware and software design was updated. The functional design requirements document was developed by KEPCO-NF and the component design was conducted by Doosan. System design and software implementation were performed by KEPCO-E and C, and software Verification and Validation (V and V) was performed by KEPCO-E and C and Sure Softtech. The ICOPS has been developed to overcome the algorithm related obstacles in overseas project. The function of I/O simulator was improved even though the hardware platform is the same as that of RCOPS for Shin-Hanul 1 and 2. SCADE was applied to the implementation of ICOPS software, and the V and V system for ICOPS which satisfies international standards was developed. Although several further detailed design works remain, the function of ICOPS has been confirmed. The ICOPS will be applied to APR+ project, and the further works will be performed in following project

  2. Development of In-Core Protection System

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J. H; Kim, C. H.; Kim, J. H.; Jeong, S. H.; Sohn, S. D.; BaeK, S. M.; YOON, J. H. [KEPCO Engineering and Construction Co., Deajeon (Korea, Republic of)

    2016-10-15

    In-core Protection System (ICOPS) is an on-line digital computer system which continuously calculates Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD) based on plant parameters to make trip decisions based on the computations. The function of the system is the same as that of Core Protection Calculator System (CPCS) and Reactor Core Protection System (RCOPS) which are applied to Optimized Power Reactor 1000 (OPR1000) and Advanced Power Reactor 1400 (APR1400). The ICOPS has been developed to overcome the algorithm related obstacles in overseas project. To achieve this goal, several algorithms were newly developed and hardware and software design was updated. The functional design requirements document was developed by KEPCO-NF and the component design was conducted by Doosan. System design and software implementation were performed by KEPCO-E and C, and software Verification and Validation (V and V) was performed by KEPCO-E and C and Sure Softtech. The ICOPS has been developed to overcome the algorithm related obstacles in overseas project. The function of I/O simulator was improved even though the hardware platform is the same as that of RCOPS for Shin-Hanul 1 and 2. SCADE was applied to the implementation of ICOPS software, and the V and V system for ICOPS which satisfies international standards was developed. Although several further detailed design works remain, the function of ICOPS has been confirmed. The ICOPS will be applied to APR+ project, and the further works will be performed in following project.

  3. Prompt Neutron Decay Constant Determination Of Silicide Transition Core Using Noise Method

    International Nuclear Information System (INIS)

    Jujuratisbela, Uju; Yulianto, Yusi Eko; Cahyana

    2001-01-01

    Chairman of BATAN had decided to replace the Oxide fuel element type of RSG-GAS into silicide element type step by step. The replacement will create core transitions. Kinetic characteristic of the transition cores have to be monitored in order to know the deviation of core behavior. For that reason, the kinetic parameters have to be measured. Prompt neutron decay constant (alpha) is one of the kinetic parameters that has to be monitored continuously in the transition cores. In order not to disturb the normal operation of reactor, alpha parameter should be measured by using noise analysis method. The voltage of neutron flux at power of 15 MW is connected to preamplifier and filter then to the Dynamic Signal Analyzer Version-2 and then the auto power spectral density (APSD) was determined by using Fast Fourier transform. From the APSD curve of each channel of JKT03, the cut off frequency of each channel can be determined by using linear regression technique such that the prompt neutron decay constant can be estimated

  4. Automatic determination of pressurized water reactor core loading patterns that maximize beginning-of-cycle reactivity within power-peaking and burnup constraints

    International Nuclear Information System (INIS)

    Hobson, G.H.; Turinsky, P.J.

    1986-01-01

    Computational capability has been developed to automatically determine a good estimate of the core loading pattern, which minimizes fuel cycle costs for a pressurized water reactor (PWR). Equating fuel cycle cost minimization with core reactivity maximization, the objective is to determine the loading pattern that maximizes core reactivity while satisfying power peaking, discharge burnup, and other constraints. The method utilizes a two-dimensional, coarse-mesh, finite difference scheme to evaluate core reactivity and fluxes for an initial reference loading pattern. First-order perturbation theory is applied to determine the effects of assembly shuffling on reactivity, power distribution, end-of-cycle burnup. Monte Carlo integer programming is then used to determine a near-optimal loading pattern within a range of loading patterns near the reference pattern. The process then repeats with the new loading pattern as the reference loading pattern and terminates when no better loading pattern can be determined. The process was applied with both reactivity maximization and radial power-peaking minimization as objectives. Results on a typical large PWR indicate that the cost of obtaining an 8% improvement in radial power-peaking margin is ≅2% in fuel cycle costs, for the reload core loaded without burnable poisons that was studied

  5. Power density forecasting device for nuclear power plant

    International Nuclear Information System (INIS)

    Fukuzaki, Takaharu; Kiguchi, Takashi.

    1978-01-01

    Purpose: To attain effective reactor operation in a bwr type reactor by forecasting the power density of the reactor after adjustment and comparing the same with the present status of the reactor by the on-line calculation in a short time. Constitution: The present status for the reactor is estimated in a present status decision section based on a measurement signal from the reactor and it is stored in an operation result collection section. The reactor status after the forecasting is estimated in a forecasting section based on a setting signal from a forecasting condition setting section and it is compared with the result value from the operation results collection section. If the forecast value does not coincide with the result value in the above comparison, the setting value in the forecast condition setting section is changed in the control section. The above procedures are repeated so as to minimize the difference between the forecast value and the result value to thereby exactly forecast the reactor status and operate the reactor effectively. (Moriyama, K.)

  6. Efficient, High Power Density Hydrocarbon-Fueled Solid Oxide Stack System, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Precision Combustion, Inc. (PCI) proposes to develop and demonstrate an innovative high power density design for direct internal reforming of regolith off-gases...

  7. Efficient, high power density hydrocarbon-fueled solid oxide stack system, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Precision Combustion, Inc. (PCI) proposes to develop and demonstrate an innovative high power density design for direct internal reforming of regolith off-gases...

  8. Improving the AGR fuel testing power density profile versus irradiation-time in the advanced test reactor

    International Nuclear Information System (INIS)

    Chang, Gray S.; Lillo, Misti A.; Maki, John T.; Petti, David A.

    2009-01-01

    The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235 U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235 U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250degC throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235 U

  9. High Power ICRH scenarios in Tore-Supra: a potential route towards improved core confinement at high density?

    Energy Technology Data Exchange (ETDEWEB)

    Rimini, F.G.; Fenzi-Bonizec, C.; Hoang, G.T.; Bourdelle, C.; Clairet, F.; Elbeze, D.; Gil, C.; Giruzzi, G.; Imbeaux, F.; Joffrin, E.; Lotte, P.; Maget, P.; Schunke, P.; Tsitrone, E. [Association Euratom-CEA, CEA/DSM/DRFC, Centre de Cadarache, 13 - Saint Paul lez Durance (France); Hennequin, P.; Honore, C. [CNRS-Ecole Polytechnique, LPTP, 91 - Palaiseau (France)

    2005-07-01

    As shown by many tokamak experiments, including Tore-Supra, an operational scenario based on tailoring the current profile via a combination of electron heating and fast plasma current ramp is particularly attractive for producing conditions favourable to internal transport barriers (ITBs) development. As we have reported in this paper, however, this is not a foolproof method for producing ITBs. In our case, reversed or flat shear conditions were indeed created, but only transiently and in a very narrow region in the plasma centre, {rho} < 0.3, and the KINEZERO code micro-stability analysis does not indicate that any significant stabilization of ITG (ion temperature gradient) or ETG (electron temperature gradient) modes is taking place. We have used the Cronos transport code to predict how we could redesign the experiment to maximize the probability to obtain more favourable conditions for triggering and sustaining high performance ITBs. Our modelling is done assuming that recycling can be moderated and edge density can be kept lower than in the actual experiments. In this case, with a substantial increase in LHCD power, from 2.5 MW to 8 MW, we could approach conditions of flat current profile over a wide region, extending to half radius, with a significant fraction of the total current, {approx} 70%, being non-inductively driven. The predictive Cronos transport model, which includes local reduction of transport according to magnetic shear, yields a long-lasting wide electron ITB located close to mid-radius.

  10. Micro-vibration response of a stochastically excited sandwich beam with a magnetorheological elastomer core and mass

    International Nuclear Information System (INIS)

    Ying, Z G; Ni, Y Q

    2009-01-01

    Magnetorheological (MR) elastomers are used to construct a smart sandwich beam for micro-vibration control. The micro-vibration response of a clamped–free sandwich beam with an MR elastomer core and a supplemental mass under stochastic support micro-motion excitation is studied. The dynamic behavior of MR elastomer as a smart viscoelastic material is described by a complex modulus which is controllable by external magnetic field. The sixth-order partial differential equation of motion of the sandwich beam is derived from the dynamic equilibrium, constitutive and geometric relations. A frequency-domain solution method for the stochastic micro-vibration response of the sandwich beam is developed by using the frequency-response function, power spectral density function and spatial eigensolution. The root-mean-square velocity response in terms of the one-third octave frequency band is calculated, and then the response reduction capacity through optimizing the complex modulus of the core is analyzed. Numerical results illustrate the influences of the MR elastomer core parameters on the root-mean-square velocity response and the high response reduction capacity of the sandwich beam. The developed analysis method is applicable to sandwich beams with arbitrary cores described by complex shear moduli under arbitrary stochastic excitations described by power spectral density functions

  11. Design and test of-80 kV snubber core assemblies for MFTF sustaining-neutral-beam power supplies

    International Nuclear Information System (INIS)

    Bishop, S.R.; Mayhall, D.J.; Wilson, J.H.; De Vore, K.R.; Ross, R.I.; Sears, R.G.

    1981-01-01

    Core snubbers, located near the neutral beam source ends of the Mirror Fusion Test Facility (MFTF) Sustaining Neutral Beam Power Supply System (SNBPSS) source cables, protect the neutral beam source extractor grid wires from overheating and sputtering during internal sparkdowns. The snubbers work by producing an induced counter-emf which limits the fault current and by absorbing the capacitive energy stored on the 80 kV source cables and power supplies. A computer program STACAL was used in snubber magnetic design to choose appropriate tape wound cores to provide 400 Ω resistance and 25 J energy absorption. The cores are mounted horizontally in a dielectric structure. The central source cable bundle passes through the snubber and terminates on three copper buses. Multilam receptacles on the buses connect to the source module jumper cables. Corona rings and shields limit electric field stresses to allow close clearances between snubbers

  12. Ultrathin Coaxial Fiber Supercapacitors Achieving High Energy and Power Densities.

    Science.gov (United States)

    Shen, Caiwei; Xie, Yingxi; Sanghadasa, Mohan; Tang, Yong; Lu, Longsheng; Lin, Liwei

    2017-11-15

    Fiber-based supercapacitors have attracted significant interests because of their potential applications in wearable electronics. Although much progress has been made in recent years, the energy and power densities, mechanical strength, and flexibility of such devices are still in need of improvement for practical applications. Here, we demonstrate an ultrathin microcoaxial fiber supercapacitor (μCFSC) with high energy and power densities (2.7 mW h/cm 3 and 13 W/cm 3 ), as well as excellent mechanical properties. The prototype with the smallest reported overall diameter (∼13 μm) is fabricated by successive coating of functional layers onto a single micro-carbon-fiber via a scalable process. Combining the simulation results via the electrochemical model, we attribute the high performance to the well-controlled thin coatings that make full use of the electrode materials and minimize the ion transport path between electrodes. Moreover, the μCFSC features high bending flexibility and large tensile strength (more than 1 GPa), which make it promising as a building block for various flexible energy storage applications.

  13. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    International Nuclear Information System (INIS)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2012-01-01

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ∼ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  14. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    Science.gov (United States)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2012-11-01

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ~ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  15. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    Energy Technology Data Exchange (ETDEWEB)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki [Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan) and Advanced Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Numazu College of Technology, Ooka 3600, Numazu, Shizuoka 410-8501 (Japan); Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan)

    2012-11-12

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to {approx} 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  16. A nodal method of calculating power distributions for LWR-type reactors with square fuel lattices

    International Nuclear Information System (INIS)

    Hoeglund, Randolph.

    1980-06-01

    A nodal model is developed for calculating the power distribution in the core of a light water reactor with a square fuel lattice. The reactor core is divided into a number of more or less cubic nodes and a nodal coupling equation, which gives the thermal power density in one node as a function of the power densities in the neighbour nodes, is derived from the neutron diffusion equations for two energy groups. The three-dimensional power distribution can be computed iteratively using this coupling equation, for example following the point Jacobi, the Gauss-Seidel or the point successive overrelaxation scheme. The method has been included as the neutronic model in a reactor core simulation computer code BOREAS, where it is combined with a thermal-hydraulic model in order to make a simultaneous computation of the interdependent power and void distributions in a boiling water reactor possible. Also described in this report are a method for temporary one-dimensional iteration developed in order to accelerate the iterative solution of the problem and the Haling principle which is widely used in the planning of reloading operations for BWR reactors. (author)

  17. A geographical model of radio-frequency power density around mobile phone masts

    International Nuclear Information System (INIS)

    Briggs, David; Beale, Linda; Bennett, James; Toledano, Mireille B.; Hoogh, Kees de

    2012-01-01

    Public concern about possible health effects of EMF radiation from mobile phone masts has led to an increase of epidemiological studies and health risk assessments which, in turn, require adequate methods of exposure estimation. Difficulties in exposure modelling are exacerbated both by the complexity of the propagation processes, and the need to obtain estimates for large study populations in order to provide sufficient statistical power to detect or exclude the small relative risks that might exist. Use of geographical information system (GIS) techniques offers the means to make such computations efficiently. This paper describes the development and field validation of a GIS-based exposure model (Geomorf). The model uses a modified Gaussian formulation to represent spatial variations in power densities around mobile phone masts, on the basis of power output, antenna height, tilt and the surrounding propagation environment. Obstruction by topography is allowed for, through use of a visibility function. Model calibration was done using field data from 151 measurement sites (1510 antenna-specific measurements) around a group of masts in a rural location, and 50 measurement sites (658 antenna-specific measurements) in an urban area. Different parameter settings were found to be necessary in urban and rural areas to obtain optimum results. The calibrated models were then validated against independent sets of data gathered from measurement surveys in rural and urban areas, and model performance was compared with that of two commonly used path-loss models (the COST-231 adaptations of the Hata and Walfisch–Ikegami models). Model performance was found to vary somewhat between the rural and urban areas, and at different measurement levels (antenna-specific power density, total power density), but overall gave good estimates (R 2 = 0.641 and 0.615, RMSE = 10.7 and 6.7 dB m at the antenna and site-level respectively). Performance was considerably better than that of both

  18. On the evolution of the density probability density function in strongly self-gravitating systems

    International Nuclear Information System (INIS)

    Girichidis, Philipp; Konstandin, Lukas; Klessen, Ralf S.; Whitworth, Anthony P.

    2014-01-01

    The time evolution of the probability density function (PDF) of the mass density is formulated and solved for systems in free-fall using a simple approximate function for the collapse of a sphere. We demonstrate that a pressure-free collapse results in a power-law tail on the high-density side of the PDF. The slope quickly asymptotes to the functional form P V (ρ)∝ρ –1.54 for the (volume-weighted) PDF and P M (ρ)∝ρ –0.54 for the corresponding mass-weighted distribution. From the simple approximation of the PDF we derive analytic descriptions for mass accretion, finding that dynamically quiet systems with narrow density PDFs lead to retarded star formation and low star formation rates (SFRs). Conversely, strong turbulent motions that broaden the PDF accelerate the collapse causing a bursting mode of star formation. Finally, we compare our theoretical work with observations. The measured SFRs are consistent with our model during the early phases of the collapse. Comparison of observed column density PDFs with those derived from our model suggests that observed star-forming cores are roughly in free-fall.

  19. Power-Production Diagnostic Tools for Low-Density Wind Farms with Applications to Wake Steering

    Science.gov (United States)

    Takle, E. S.; Herzmann, D.; Rajewski, D. A.; Lundquist, J. K.; Rhodes, M. E.

    2016-12-01

    Hansen (2011) provided guidelines for wind farm wake analysis with applications to "high density" wind farms (where average distance between turbines is less than ten times rotor diameter). For "low-density" (average distance greater than fifteen times rotor diameter) wind farms, or sections of wind farms we demonstrate simpler sorting and visualization tools that reveal wake interactions and opportunities for wind farm power prediction and wake steering. SCADA data from a segment of a large mid-continent wind farm, together with surface flux measurements and lidar data are subjected to analysis and visualization of wake interactions. A time-history animated visualization of a plan view of power level of individual turbines provides a quick analysis of wake interaction dynamics. Yaw-based sectoral histograms of enhancement/decline of wind speed and power from wind farm reference levels reveals angular width of wake interactions and identifies the turbine(s) responsible for the power reduction. Concurrent surface flux measurements within the wind farm allowed us to evaluate stability influence on wake loss. A one-season climatology is used to identify high-priority candidates for wake steering based on estimated power recovery. Typical clearing prices on the day-ahead market are used to estimate the added value of wake steering. Current research is exploring options for identifying candidate locations for wind farm "build-in" in existing low-density wind farms.

  20. Accelerator driven sub-critical core

    Science.gov (United States)

    McIntyre, Peter M; Sattarov, Akhdiyor

    2015-03-17

    Systems and methods for operating an accelerator driven sub-critical core. In one embodiment, a fission power generator includes a sub-critical core and a plurality of proton beam generators. Each of the proton beam generators is configured to concurrently provide a proton beam into a different area of the sub-critical core. Each proton beam scatters neutrons within the sub-critical core. The plurality of proton beam generators provides aggregate power to the sub-critical core, via the proton beams, to scatter neutrons sufficient to initiate fission in the sub-critical core.