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Sample records for flux core spheromak

  1. Studies on spheromak plasma production by external-flux-core method, (2)

    International Nuclear Information System (INIS)

    Arata, Masanori; Katsurai, Makoto

    1984-01-01

    The spheromak technique, one of magnetic plasma containment techniques, has such arrangement of magnetic fields that the toroidal magnetic field is produced by the poloidal current flowing in plasma, and the poloidal magnetic field is produced by the toroidal current in plasma and the current in external coils. The authors proposed external flux core method as the technique of plasma formation by this spheromak method, in which the toroidal magnetic field is injected by the discharge using electrodes, whereas the poloidal magnetic field is injected by induction discharge without electrode. Its fundamental action was analyzed by computer simulation and confirmed by experiment. In this study, the behavior of the spheromak plasma produced was investigated in detail and summarized. The contents were the measurement of the spheromak configuration produced and the estimation of plasma parameters. The experimental setup, the principle of action, and the experimental results of magnetic field distribution obtained by a magnetic probe, ion current measured by an electrostatic probe, electron temperature by spectroscopic measurement and the behavior of spheromak plasma observed with an image converter camera are reported. (Kako, I.)

  2. Tilt and shift mode stability in a spheromak with a flux core

    Energy Technology Data Exchange (ETDEWEB)

    Finn, J.M.; Jardin, S.C.

    1984-07-01

    The stability of spheromak equilibria with a flux core, or reversal coil, is studied by means of an ideal MHD code. Results depend critically upon whether the flux hole region (the current free area just inside the separatrix) is treated as perfectly conducting plasma or as vacuum. This indicates that the tilt and shift modes persist as resistive instabilities if they are stable in ideal MHD. Specifically, for nonoptimally shaped equilibria, the flux core must nearly touch the current channel if the flux hole is vacuum, whereas the core may be slightly outside the separatrix if the flux hole has conducting plasma. A larger margin exists for optimally shaped equilibria.

  3. Magnetic flux conversion and relaxation toward a minimum-energy state in S-1 spheromak plasmas

    International Nuclear Information System (INIS)

    Janos, A.

    1985-09-01

    S-1 Spheromak currents and magnetic fluxes have been measured with Rogowski coils and flux loops external to the plasma. Toroidal plasma currents up to 350 kA and spheromak configuration lifetimes over 1.0 msec have been achieved at moderate power levels. The plasma formation in the S-1 Spheromak device is based on an inductive transfer of poloidal and toroidal magnetic flux from a toroidal ''flux core'' to the plasma. Formation is programmed to guide the configuration into a force-free, minimum-energy Taylor state. Properly detailed programming of the formation process is found not to be essential since plasmas adjust themselves during formation to a final equilibrium near the Taylor state. After formation, if the plasma evolves away from the stable state, then distinct relaxation oscillation events occur which restore the configuration to that stable state. The relaxation process involves reconnection of magnetic field lines, and conversion of poloidal to toroidal magnetic flux (and vice versa) has been observed and documented. The scaling of toroidal plasma current and toroidal magnetic flux in the plasma with externally applied currents is consistent with the establishment of a Taylor state after formation. In addition, the magnetic helicity is proportional to that injected from the flux core, independent of how that helicity is generated

  4. Stability of force-free spheromak plasma in spheroidal flux conserver

    International Nuclear Information System (INIS)

    Kaneko, Shobu; Tsutsui, Hiroaki

    1988-01-01

    The Woltjer-Taylor method is applied to spheromak plasmas in spheroidal flux conservers. As models of the flux conserver, both oblate and prolate spheroidal vessels with a center conductor are used. The plasma is not assumed to be nearly spherical, and the Rayleigh-Ritz method and the finite element method are used to evaluate the eigenvalues. The oblate spheromak is shown to be stable irrespective of the shape of the flux conserver. Though the prolate spheromak is unstable if there is no center conductor, it can be stable if the center conductor is installed. (author)

  5. Magnetic helicity balance in the Sustained Spheromak Plasma Experiment

    International Nuclear Information System (INIS)

    Stallard, B.W.; Hooper, E.B.; Woodruff, S.; Bulmer, R.H.; Hill, D.N.; McLean, H.S.; Wood, R.D.

    2003-01-01

    The magnetic helicity balance between the helicity input injected by a magnetized coaxial gun, the rate-of-change in plasma helicity content, and helicity dissipation in electrode sheaths and Ohmic losses have been examined in the Sustained Spheromak Plasma Experiment (SSPX) [E. B. Hooper, L. D. Pearlstein, and R. H. Bulmer, Nucl. Fusion 39, 863 (1999)]. Helicity is treated as a flux function in the mean-field approximation, allowing separation of helicity drive and losses between closed and open field volumes. For nearly sustained spheromak plasmas with low fluctuations, helicity balance analysis implies a decreasing transport of helicity from the gun input into the spheromak core at higher spheromak electron temperature. Long pulse discharges with continuously increasing helicity and larger fluctuations show higher helicity coupling from the edge to the spheromak core. The magnitude of the sheath voltage drop, inferred from cathode heating and a current threshold dependence of the gun voltage, shows that sheath losses are important and reduce the helicity injection efficiency in SSPX

  6. Steady-state operation of spheromaks by inductive techniques

    International Nuclear Information System (INIS)

    Janos, A.

    1984-04-01

    A method to maintain a steady-state spheromak configuration inductively using the S-1 Spheromak device is described. The S-1 Spheromak formation apparatus can be utilized to inject magnetic helicity continuously (C.W., not pulsed or D.C.) into the spheromak configuration after equilibrium is achieved in the linked mode of operation. Oscillation of both poloidal- and toroidal-field currents in the flux core (psi-phi Pumping), with proper phasing, injects a net time-averaged helicity into the plasma. Steady-state maintenance relies on flux conversion, which has been earlier identified. Relevant experimental data from the operation of S-1 are described. Helicity flow has been measured and the proposed injection scheme simulated. In a reasonable time practical voltages and frequencies can inject an amount of helicity comparable to that in the initial plasma. Plasma currents can be maintained or increased. This pumping technique is similar to F-THETA Pumping of a Reversed-Field-Pinch but is applied to this inverse-pinch formation

  7. Sustained spheromak physics experiment

    International Nuclear Information System (INIS)

    Hooper, E.B.; Bulmer, R.H.; Cohen, B.I.

    2001-01-01

    The Sustained Spheromak Physics Experiment, SSPX, will study spheromak physics with particular attention to energy confinement and magnetic fluctuations in a spheromak sustained by electrostatic helicity injection. In order to operate in a low collisionality mode, requiring T e >100 eV, vacuum techniques developed for tokamaks will be applied, and a divertor will be used for the first time in a spheromak. The discharge will operate for pulse lengths of several milliseconds, long compared to the time to establish a steady-state equilibrium but short compared to the L/R time of the flux conserver. The spheromak and helicity injector ('gun') are closely coupled, as shown by an ideal MHD model with force-free injector and edge plasmas. The current from the gun passes along the symmetry axis of the spheromak, and the resulting toroidal magnetic field causes the safety factor, q, to diverge on the separatrix. The q-profile depends on the ratio of the injector current to spheromak current and on the magnetic flux coupling the injector to the spheromak. New diagnostics include magnetic field measurements by a reflectometer operating in combined O- and X-modes and by a transient internal probe (TIP). (author)

  8. Sustained spheromak physics experiment

    International Nuclear Information System (INIS)

    Hooper, E.B.; Bulmer, R.H.; Cohen, B.I.

    1999-01-01

    The Sustained Spheromak Physics Experiment, SSPX, will study spheromak physics with particular attention to energy confinement and magnetic fluctuations in a spheromak sustained by electrostatic helicity injection. In order to operate in a low collisionality mode, requiring T e > 100 eV, vacuum techniques developed for tokamaks will be applied, and a divertor will be used for the first time in a spheromak. The discharge will operate for pulse lengths of several milliseconds, long compared to the time to establish a steady-state equilibrium but short compared to the L/R time of the flux conserver. The spheromak and helicity injector ('gun') are closely coupled, as shown by an ideal MHD model with force-free injector and edge plasmas. The current from the gun passes along the symmetry axis of the spheromak, and the resulting toroidal magnetic field causes the safety factor, q, to diverge on the separatrix. The q-profile depends on the ratio of the injector current to spheromak current and on the magnetic flux coupling the injector to the spheromak. New diagnostics include magnetic field measurements by a reflectometer operating in combined O- and X-modes and by a transient internal probe (TIP). (author)

  9. Review of spheromak research

    International Nuclear Information System (INIS)

    Jarboe, T.R.

    1994-01-01

    Spheromak research from 1979 to the present is reviewed including over 160 references. Emphasis is on understanding and interpretation of results. In addition to summarizing results some new interpretations are presented. An introduction and brief history is followed by a discussion of generalized helicity and its time derivative. Formation and sustainment are discussed including five different methods, flux core, θ-pinch z-pinch, coaxial source, conical θ-pinch, and kinked z-pinch. All methods are helicity injections. Steady-state methods and rules for designing spheromak experiments are covered, followed by equilibrium and stability. Methods of stabilizing the tilt and shift modes are discussed as well as their impact on the reactor designs. Current-driven and pressure-driven instabilities as well as relaxation in general are covered. Energy confinement is discussed in terms of helicity decay time and βs limits. The confinement in high and low open-flux geometries are compared and the reactor implications discussed. (author)

  10. Magnetic Reconnection Results on the Swarthmore Spheromak Experiment

    Science.gov (United States)

    Kornack, T. W.; Sollins, P. K.; Brown, M. R.

    1997-11-01

    Linear and 2D arrays of magnetic probes are used to study magnetic reconnection in the Swarthmore Spheromak Experiment (SSX). Opposing coaxial plasma guns form two identical spheromaks into adjacent 0.5 m diameter copper flux conservers. The flux conservers have symmetrical openings that allow the spheromaks to merge in a controlled manner. The stable equilibrium of the spheromaks provides a reservoir of magnetic flux for reconnection experiments. Currently, the magnetic configuration of the spheromaks allows the study of counter-helicity reconnection. Preliminary analysis will be presented and may include 2D B field movies of the reconnection region, measurement of the reconnection rate and comparison to the Sweet-Parker and standard Petschek models.

  11. Design of a spheromak compressor driven by high explosives

    International Nuclear Information System (INIS)

    Henins, I.; Fernandez, J.C.; Jarboe, T.R.; Marsh, S.P.; Marklin, G.J.; Mayo, R.M.; Wysocki, F.J.

    1990-01-01

    High energy density spheromaks can be used to accelerate a thin section of the flux conserver wall to high velocities. The energy density of a spheromak, formed by conventional helicity injection into a flux conserver, can be increased by reducing the flux conserver volume after the spheromak is formed. A method of accomplishing this is by imploding one wall of the flux conserver with high explosives. The authors have embarked on a program to demonstrate that a spheromak can be used as an energy transfer medium, and that a velocity gain over high-explosive driven plate velocities can be achieved. To do this, a plasma gun helicity source that will inject a spheromak with suitable initial energy density and lifetime is needed. Also, an implodable flux conserver that remains intact and clean during the implosion must be developed. The flux conserver problem is probably the more challenging one, because very little experimental work has been done in the past on explosively driven metal plates into a high vacuum, with sizes and travel distances appropriate for their application. There are two necessary practical requirements for an explosive compression of a flux conserver. The first is that the imploding wall does not rupture. The second is that gasses or other debri are not ejected which could penetrate and poison the spheromak plasma, and thus reduce the spheromak lifetime below what is necessary to carry out the spheromak compression and the subsequent acceleration of the flyer plate. The authors have designed and fabricated a plasma gun to be used for injecting the initial spheromak plasma into the collapsible flux conserver

  12. Experimental identification of the kink instability as a poloidal flux amplification mechanism for coaxial gun spheromak formation

    International Nuclear Information System (INIS)

    Hsu, S.C.; Bellan, P.M.

    2003-01-01

    The magnetohydrodynamic kink instability is observed and identified experimentally as a poloidal flux amplification mechanism for coaxial gun spheromak formation. Plasmas in this experiment fall into three distinct regimes which depend on the peak gun current to magnetic flux ratio, with (I) low values resulting in a straight plasma column with helical magnetic field, (II) intermediate values leading to kinking of the column axis, and (III) high values leading immediately to a detached plasma. Onset of column kinking agrees quantitatively with the Kruskal-Shafranov limit, and the kink acts as a dynamo which converts toroidal to poloidal flux. Regime II clearly leads to both poloidal flux amplification and the development of a spheromak configuration

  13. Experimental identification of the kink instability as a poloidal flux amplification mechanism for coaxial gun spheromak formation.

    Science.gov (United States)

    Hsu, S C; Bellan, P M

    2003-05-30

    The magnetohydrodynamic kink instability is observed and identified experimentally as a poloidal flux amplification mechanism for coaxial gun spheromak formation. Plasmas in this experiment fall into three distinct regimes which depend on the peak gun current to magnetic flux ratio, with (I) low values resulting in a straight plasma column with helical magnetic field, (II) intermediate values leading to kinking of the column axis, and (III) high values leading immediately to a detached plasma. Onset of column kinking agrees quantitatively with the Kruskal-Shafranov limit, and the kink acts as a dynamo which converts toroidal to poloidal flux. Regime II clearly leads to both poloidal flux amplification and the development of a spheromak configuration.

  14. Magnetohydrodynamic stability of spheromak plasma in spheroidal flux conserver

    International Nuclear Information System (INIS)

    Kaneko, Shobu; Kamitani, Atsushi.

    1985-11-01

    The MHD equilibrium configurations of spheromak plasmas in a spheroidal flux conserver are determined by use of a pressure distribution whose derivative dp/dψ vanishes on the magnetic axis, and by use of an optimized distribution. Here p is the pressure and ψ is the flux function. These equilibria are shown to be stable for symmetric modes. The stability for localized modes is investigated by the Mercier criterion. The values of the maximum beta ratio β max are evaluated for both pressure distributions and are shown to become about two times larger by optimization. If the condition, q axis max are found to be less than 30 %. The oblate spheroidal flux conserver is shown to be better than the toroidal conserver with a rectangular cross section from the standpoint of stability. (author)

  15. Spheromak Merging Experiments on SSX

    Science.gov (United States)

    Brown, M. R.; Kornack, T. W.; Sollins, P. K.; Luh, W. J.

    1997-11-01

    Spheromak merging experiments are underway at the Swarthmore Spheromak Experiment (SSX) at Swarthmore College. The spheromaks are formed by identical magnetized plasma guns and equilibrium is established in close fitting 0.5 m diameter copper flux conservers. Partial merging is achieved through openings in the back wall. We observe the formation of a reconnection boundary layer at the interface of the two spheromaks using a linear probe array. The characteristic scale of the flux reversal is about 1 cm (consistent with the diffusion scale δ_diff, the ion Larmor radius ρi and the ion inertial length c/ω_pi). Movies of the formation and evolution of the layer will be presented. Correlations between reconnection events and pulses of soft x-rays and energetic particles will be presented if available. Plans for 2D and 3D imaging of the layer will also be discussed.

  16. Steady-state spheromak

    International Nuclear Information System (INIS)

    Jarboe, T.R.

    1982-01-01

    A major effort is being made in the national program to make the operation of axisymmetric, toroidal confinement systems steady state by the application of expensive rf current drive. Described here is a method by which such a confinement system, the spheromak, can be refluxed indefinitely through the application of dc power. As a step towards dc sustainment we have operated the present CTX source in the slow source mode with a longer power application time (approx. 0.1 ms) and successfully generated long-lived spheromaks. If the erosion of the electrodes can be controlled as well as it is with MPD arcs then dc operation should be very clean. If only a small fraction (approx. 10% for an experiment) of the poloidal flux of the spheromak connects to the source then the dc sustainment can be very efficient. The amount of connecting flux that is necessary for sustainment needs to be determined experimentally

  17. Sustained Spheromak Physics Experiment, SSPX

    International Nuclear Information System (INIS)

    Hooper, E.B.

    1997-01-01

    The Sustained Spheromak Physics Experiment is proposed for experimental studies of spheromak confinement issues in a controlled way: in steady state relative to the confinement timescale and at low collisionality. Experiments in a flux - conserver will provide data on transport in the presence of resistive modes in shear-stabilized systems and establish operating regimes which pave the way for true steady-state experiments with the equilibrium field supplied by external coils. The proposal is based on analysis of past experiments, including the achievement of T e = 400 eV in a decaying spheromak in CTX. Electrostatic helicity injection from a coaxial ''''gun'''' into a shaped flux conserver will form and sustain the plasma for several milliseconds. The flux conserver minimizes fluxline intersection with the walls and provides MHD stability. Improvements from previous experiments include modem wall conditioning (especially boronization), a divertor for density and impurity control, and a bias magnetic flux for configurational flexibility. The bias flux will provide innovative experimental opportunities, including testing helicity drive on the large-radius plasma boundary. Diagnostics include Thomson scattering for T e measurements and ultra-short pulse reflectrometry to measure density and magnetic field profiles and turbulence. We expect to operate at T e of several hundred eV, allowing improved understanding of energy and current transport due to resistive MHD turbulence during sustained operation. This will provide an exciting advance in spheromak physics and a firm basis for future experiments in the fusion regime

  18. Sustained spheromak experiments in CTX

    International Nuclear Information System (INIS)

    Jarboe, T.R.; Barnes, C.W.; Henins, I.; Hoida, H.W.; Linford, R.K.; Sherwood, A.R.

    1983-01-01

    So far, spheromaks can be sustained as long as the source is injecting helicity. When the injection stops the configuration decays. Spheromks have been sustained for more than 1 ms with total lifetimes of more than 2 ms. The physical properties of the sustained spheromak are under investigation in this paper. Preliminary data indicate that (B) approx. = 2 kG, n approx. = 2 x 10 14 -cm -3 and T /sub e/ approx. = 20-30 eV. An helicity decay rate is determined from the ratio of an estimate of the helicity content of the spheromak and the rate of helicity flow from the source. In the coaxial source geometry a constant value of poloidal flux /PHI/ /sub p/ is placed inside the center electrode. By applying a voltage V between the two electrodes toroidal flux is injected (/PHI/ /sub t/ =V) which links the poloidal flux. The rate of helicity injection is then 2V/PHI/ /sub p/ . The helicity content of the spheromak is estimated by measuring the fields at one point and using the model described above to calculate the profiles. The result is that /TAU/ /sub Hel/ approx. = 200 us. This value is about the same as the /TAU/ /sub B/ 2 of a decaying spheromak with similar parameters. These results indicate that helicity injection is possible and that a large fraction (30-100%) of the injected helicity is absorbed

  19. Tearing-mode stability of a forming Spheromak plasma

    International Nuclear Information System (INIS)

    Heidbrink, W.W.; Jardin, S.C.; Chance, M.S.

    1981-10-01

    The results of numerical calculations of Δ' for a class of equilibria typical of those encountered during the early formation stage of the S1 Spheromak are presented. The equilibrium plasma is assumed to be cylindrically symmetric and pressureless. It encloses a current carrying perfect conductor (flux core) and is surrounded by a vacuum with zero longitudinal field. Stability boundaries in the space formed by the equilibrium parameters are mapped. The plasma is tearing mode stable provided B/sub z//B/sub theta/ at the flux core is below a certain critical value which depends on the equilibrium parameters. For typical equilibria, this critical value is 0.65

  20. Scaling studies of spheromak formation and equilibrium

    International Nuclear Information System (INIS)

    Geddes, C.G.; Kornack, T.W.; Brown, M.R.

    1998-01-01

    Formation and equilibrium studies have been performed on the Swarthmore Spheromak Experiment (SSX). Spheromaks are formed with a magnetized coaxial plasma gun and equilibrium is established in both small (d small =0.16 m) and large (d large =3d small =0.50 m) copper flux conservers. Using magnetic probe arrays it has been verified that spheromak formation is governed solely by gun physics (in particular the ratio of gun current to flux, μ 0 I gun /Φ gun ) and is independent of the flux conserver dimensions. It has also been verified that equilibrium is well described by the force free condition ∇xB=λB (λ=constant), particularly early in decay. Departures from the force-free state are due to current profile effects described by a quadratic function λ=λ(ψ). Force-free SSX spheromaks will be merged to study magnetic reconnection in simple magnetofluid structures. copyright 1998 American Institute of Physics

  1. Increased particle confinement with the use of external dc bias field in the CTX spheromak

    International Nuclear Information System (INIS)

    Barnes, C.W.; Hoida, H.W.; Henins, I.; Fernandez, J.C.; Jarboe, T.R.; Marklin, G.J.

    1985-01-01

    Spheromaks are formed in a mesh flux conserver in the presence of an external dc bias field. The spheromaks remain stable to tilt instabilities with ratios of bias-to-spheromak flux of up to 47 +- 7%. Normally applied bias flux puts the spheromak separatrix inside the metal mesh and improves the particle confinement

  2. Spheromak tilting and its stability control

    International Nuclear Information System (INIS)

    Hayashi, T.; Sato, T.

    1983-01-01

    Spheromak tilting instability was studied. A numerical technique to create a rather arbitrarily-shaped spheromak like the one with a flux hole was investigated. The dynamics governing the tilting instability, namely, the influence of the magnetic index, the toroidal current (q-profile) and the resistivity upon the tilting growth rate, and the roles of magnetc reconnection upon the nonlinear development were studied. The best way to control the tilting instability was invented. The stabilizing effects of the vertical wall, the isolated conducting cylindrical belt, and the horizontal wall were studied. Central pole stabilization was also investigated. The influence of the wall condition, namely, whether the wall acted as a flux conserver in the spheromak creation stage or not is discussed. The present study has shown that the three- dimensional simulation is indeed useful and practical in not only studying the underlying physics but also finding a stabilization technique of spheromaks. (Kato, T.)

  3. Theoretical issues in Spheromak research

    International Nuclear Information System (INIS)

    Cohen, R. H.; Hooper, E.B.; LoDestro, L.L.; Mattor, N.; Pearlstein, L.D.; Ryutov, D.D.

    1997-01-01

    This report summarizes the state of theoretical knowledge of several physics issues important to the spheromak. It was prepared as part of the preparation for the Sustained Spheromak Physics Experiment (SSPX), which addresses these goals: energy confinement and the physics which determines it; the physics of transition from a short-pulsed experiment, in which the equilibrium and stability are determined by a conducting wall (''''flux conserver'''') to one in which the equilibrium is supported by external coils. Physics is examined in this report in four important areas. The status of present theoretical understanding is reviewed, physics which needs to be addressed more fully is identified, and tools which are available or require more development are described. Specifically, the topics include: MHD equilibrium and design, review of MHD stability, spheromak dynamo, and edge plasma in spheromaks

  4. Stability of spheroidal spheromak plasma by use of force-free approximation

    International Nuclear Information System (INIS)

    Kaneko, Shobu; Tsutsui, Hiroaki.

    1987-09-01

    The Woltjer-Taylor method is applied to spheromak plasmas in spheroidal flux conservers. As models of the flux conserver, both oblate and prolate spheroidal vessels with a center conductor are used. The plasma is not assumed to be nearly spherical, and the Rayleigh-Ritz method and the finite element method are used to evaluate the eigenvalues. The oblate spheromak is shown to be stable irrespective of the shape of the flux conserver. Though the prolate spheromak is unstable if there is no center conductor, it can be stable if the center conductor is installed. (author)

  5. The Spheromak path to fusion energy

    International Nuclear Information System (INIS)

    Hooper, E.B.; Barnes, C.W.; Bellan, P.M.

    1998-01-01

    The spheromak is a simple and robust magnetofluid configuration with several attractive reactor attributes including compact geometry, no material center post, high engineering β, and sustained steady state operation through helicity injection. Spheromak physics was extensively studied in the US program and abroad (especially Japan) in the 1980' s with work continuing into the 1990s in Japan and the UK. Scientific results included demonstration of self-organization at constant helicity, control of the tilt and shift modes by shaped flux conservers, elucidation of the role of magnetic reconnection in the magnetic dynamo, and sustainment of a spheromak by helicity injection. Several groups attained electron temperatures above 100 eV in decaying plasmas, with CTX reaching 400 eV. This experiment had high magnetic field (>l T on the edge and ∼ 3 T near the symmetry axis) and good confinement. More recently, analysis of CTX found the energy confinement in the plasma core to be consistent with Rechester-Rosenbluth transport in a fluctuating magnetic field, potentially scaling to good confinement at higher electron temperatures. The SPHEX group developed an understanding of the dynamo in sustained spheromaks but in a relatively cold device. These and other physics results provide a foundation for a new ''concept exploration'' experiment to study the physics of a hot, sustained spheromak. If successful, this work leads to a next generation, proof-of-principle program. The new SSPX experiment will address the physics of a large-scale sustained spheromak in a national laboratory (LLNL) setting. The key issue in near term spheromak research will be to explore the possibly deleterious effects of sustainment on confinement. Other important issues include exploring the β scaling of confinement, scaling with Lundquist number S, and determining the need for active current-profile control. Collaborators from universities and other national laboratories are contributing

  6. Configuration of gun-generated spheromak in effectively closed metal flux conserver

    International Nuclear Information System (INIS)

    Kato, Yushi; Nishikawa, Masahiro; Honda, Yoshihide; Satomi, Norio; Watanabe, Kenji

    1988-01-01

    In the CTCC-II spheromak experiment, the gun-generated plasma is confined in a spheroidal aluminum flux conserver (FC) with a choking coil. This coil produces the additional magnetic field to close perfectly all magnetic surfaces into the FC, i.e. the entrance hole for plasma injection is enable to be closed by magnetic field. Hence the plasma is confined in the effectively closed metal FC. In this experiment the average life time is 1.2 msec, and electron density and temperature are n e = 2 x 10 13 /cc, T e = 30 eV, respectively. The configuration with a flux hole region in which the toroidal magnetic field vanishes around the geometrical axis has been observed in the FC. The radius of the flux hole depends on the condition how the external choking field is applied. The flux hole enhances the magnetic shear near the plasma surfaces and, therefore, has a stabilizing effect even without inserting the central conducting pole. (author)

  7. Investigations into the relationship between spheromak, solar, and astrophysical plasmas

    International Nuclear Information System (INIS)

    Bellan, P.M.; Hsu, S.C.; Hansen, J.F.; Tokman, M.; Pracko, S.E.; Romero-Talamas, C.A.

    2003-01-01

    Spheromaks offer the potential for a simple, low cost fusion reactor and involve physics similar to certain solar and astrophysical phenomena. A program to improve understanding of spheromaks by exploiting this relationship is underway using (i) a planar spheromak gun and (ii) a solar prominence simulator. These devices differ in symmetry but both involve spheromak technology whereby high-voltage is applied across electrodes linking a bias magnetic flux created by external coils. The planar spheromak gun consists of a co-planar disk and annulus linked by a poloidal bias field. Application of high voltage across the gap between disk and annulus drives a current along the bias field. If the current to flux ratio exceeds the inverse of the characteristic linear dimension, a spheromak is ejected. A distinct kink forms just below the ejection threshold. The solar simulation gun consists of two adjacent electromagnets which generate a 'horse-shoe' arched bias field. A current is driven along this arched field by a capacitor bank. The current channel first undergoes pinching, then writhes, and finally bulges outwards due to the hoop force. (author)

  8. Re-examination of spheromak experiments and opportunities

    International Nuclear Information System (INIS)

    Hooper, B.E.; Hammer, J.H.; Barnes, C.W.; Fernandez, J.C.; Wysocki, F.J.

    1996-01-01

    The results of spheromak experiments are reexamined in light of the hypothesis that the core energy confinement is considerably better than the global confinement and that it extrapolates favorably with magnetic Reynolds number S. The data in decaying spheromaks are found to be consistent with the hypothesis and with magnetic fluctuations scaling as S -1/2 and determining the electron thermal conductivity. No conclusion is drawn from the data for sustained spheromaks, indicating the importance of a new experiment to determine core energy confinement while helicity is injected. The characteristics of such an experiment are discussed, including the importance of using modern vacuum and wall-conditioning techniques and of minimizing magnetic field errors. 44 refs., 7 figs., 1 tab

  9. Progress with energy confinement time in the CTX spheromak

    International Nuclear Information System (INIS)

    Jarboe, T.R.; Fernandez, J.C.; Wysocki, F.J.; Barnes, C.W.; Henins, I.; Knox, S.O.; Marklin, G.J.

    1990-01-01

    The 0.67 m radius mesh flux conserver (MFC) in CTX was replaced by a solid flux conserver (SFC), resulting in greatly reduced field errors. Decreased spheromak open flux led to vastly improved decaying discharged, including increased global energy confinement times, τ E (from 20 to 180 μs), and corresponding magnetic energy decay times, τ B 2 (from 0.7 to 2 ms). Improved confinement allowed the observation of the pressure-driven instability (predicted by Mercier) which ejects plasma from the spheromak interior to the wall

  10. Experimental Identification of the Kink Instability as a Poloidal Flux Amplification Mechanism for Coaxial Gun Spheromak Formation

    OpenAIRE

    Hsu, S. C.; Bellan, P. M.

    2003-01-01

    The magnetohydrodynamic kink instability is observed and identified experimentally as a poloidal flux amplification mechanism for coaxial gun spheromak formation. Plasmas in this experiment fall into three distinct regimes which depend on the peak gun current to magnetic flux ratio, with (I) low values resulting in a straight plasma column with helical magnetic field, (II) intermediate values leading to kinking of the column axis, and (III) high values leading immediately to a detached plasma...

  11. Simulation of Spheromak Evolution and Energy Confinement

    International Nuclear Information System (INIS)

    Cohen, B; Hooper, E; Cohen, R; Hill, D; McLean, H; Wood, R; Woodruff, S; Sovinec, C; Cone, G

    2004-01-01

    Simulation results are presented that illustrate the formation and decay of a spheromak plasma driven by a coaxial electrostatic plasma gun, and that model the energy confinement of the plasma. The physics of magnetic reconnection during spheromak formation is also illuminated. The simulations are performed with the three-dimensional, time-dependent, resistive magnetohydrodynamic NIMROD code. The simulation results are compared to data from the SSPX spheromak experiment at the Lawrence Livermore National Laboratory. The simulation results are tracking the experiment with increasing fidelity (e.g., improved agreement with measurements of the magnetic field, fluctuation amplitudes, and electron temperature) as the simulation has been improved in its representations of the geometry of the experiment (plasma gun and flux conserver), the magnetic bias coils, and the detailed time dependence of the current source driving the plasma gun, and uses realistic parameters. The simulations are providing a better understanding of the dominant physics in SSPX, including when the flux surfaces close and the mechanisms limiting the efficiency of electrostatic drive

  12. Spheromak Physics Development

    International Nuclear Information System (INIS)

    Hooper, E.B.

    1997-01-01

    The spheromak is a Magnetic Fusion Energy (MFE) configuration, which is a leading alternative to the tokamak. It has a simple geometry which offers an opportunity to achieve the promise of fusion energy if the physics of confinement, current drive, and pressure holding capability extrapolate favorably to a reactor. Recent changes in the US MFE program, taken in response to budget constraints and programmatic directions from Congress, include a revitalization of an experimental alternative concept effort. Detailed studies of the spheromak were consequently undertaken to examine the major physics issues which need to be resolved to advance it as a fusion plasma, the optimum configuration for an advanced experiment, and its potential as a reactor. As a result of this study, we conclude that it is important to evaluate several physics issues experimentally. Such an experiment might be appropriately be named the Sustained Spheromak Physics Experiment (SSPX). It would address several critical issues, the solution to which will provide the physics basis to enable an advanced experiment. The specific scientific goals of SSPX would be to: * Demonstrate that electron and ion temperatures of a few hundred electron volts can be achieved in a steady-state spheromak plasma sustained by a magnetic dynamo (''helicity injection''). * Relate energy confinement quantitatively to the magnetic turbulence accompanying the dynamo and use this knowledge to optimize performance. * Measure the magnetic field profiles and magnetic turbulence in the plasma and relate these to the science of the magnetic dynamo which drives the current in the plasma. * Examine experimentally the pressure holding capability (''beta limit'') of the spheromak. * Understand the initial phases of the transition of the plasma from an equilibrium supported by a magnetic-flux conserving wall to one supported by external coils

  13. Pressure effect on equilibrium configuration of CTCC-2 spheromak

    International Nuclear Information System (INIS)

    Nishikawa, M.; Kato, Y.; Satomi, N.; Watanabe, K.

    1990-01-01

    In CTCC-2 experiment, the initial plasma is produced by a magnetized gun and ejected into a metallic aluminum flux conserver (FC) with thickness of 15 mm. The spheromak is formed in the FC during a life time of 1.5 ms, in which the plasma is isolated from any external feeder. A choking-field-generating coil is equipped on the entrance of the spheroidal FC. The choking field is suppressing some leakage of spheromak field along the entrance duct, which is made of thin stainless steel plate (0.8 mm) for rapid penetration of the choking magnetic field. This resistive part acts as an effective plasma current limiter, which produces stable currentless region (flux hole). The flux hole increases magnetic shear without inserting a central conducting pole along the symmetric axis and is controlled to decrease with the choking field strength. Thus, in CTCC-2 spheromak, a stable oblate spheroidal boundary is rigidly fixed by the metal wall of FC and the entrance hole of FC is effectively closed by choking magnetic field, so that it is suitable to investigate precisely a fine structure of configuration. In spheromak configuration whose aspect ratio is near one, the ratio of the magnetic strength at the inner part to that at the outer part on equi-flux surface (mirror ratio) becomes very large in comparison with that of a large aspect ratio. This extreme configuration with a high mirror ratio may be associated with an anisotropic pressure effect even in collisional state like as our experimental condition. They have investigated the pressure effect on spheromak configuration in more detail. The obtained equilibrium profile is grossly explained by a theoretical profile on assuming low beta limit until now. However, the authors observe a systematic discrepancy between a measured poloidal profile and a theoretical one as mentioned

  14. Simulation of Spheromak Evolution and Energy Confinement

    International Nuclear Information System (INIS)

    Cohen, B.; Hooper, E.; Cohen, R.; Hill, D.; McLean, H.; Wood, R.; Woodruff, S.

    2004-01-01

    Simulation results are presented that illustrate the formation and decay of a spheromak plasma driven by a coaxial electrostatic plasma gun, and that model the energy confinement of the plasma. The physics of magnetic reconnection during spheromak formation is also illuminated. The simulations are performed with the three-dimensional, time-dependent, resistive magnetohydrodynamic NIMROD code. The dimensional, simulation results are compared to data from the SSPX spheromak experiment at the Lawrence Livermore National Laboratory. The simulation results are tracking the experiment with increasing fidelity (e.g., improved agreement with measurements of the magnetic field, fluctuation amplitudes, and electron temperature) as the simulation has been improved in its representations of the geometry of the experiment (plasma gun and flux conserver), the magnetic bias coils, and the detailed time dependence of the current source driving the plasma gun, and uses realistic parameters. The simulations are providing a better understanding of the dominant physics in SSPX, including when the flux surfaces close and the mechanisms limiting the efficiency of electrostatic drive

  15. Theoretical investigation of field-line quality in a driven spheromak

    International Nuclear Information System (INIS)

    Cohen, R.H.; Cohen, B.I.; Berk, H.L.

    2003-01-01

    Theoretical studies aimed at predicting and diagnosing field-line quality in a spheromak are described. These include nonlinear 3-D MHD simulations, stability studies, analyses of confinement in spheromaks dominated by either open (stochastic) field lines or approximate flux surfaces, and a theory of fast electrons as a probe of field-line length. (author)

  16. Current drive by spheromak injection into a tokamak

    International Nuclear Information System (INIS)

    Brown, M.R.; Bellan, P.M.

    1990-01-01

    The authors report the first observation of current drive by spheromak injection into a tokamak due to the process of helicity injection. Current drive is observed in Caltech's ENCORE tokamak (30% increase, ΔI > 1 kA) only when both the tokamak and injected spheromak have the same sign of helicity (where helicity is defined as positive if current flows parallel to magnetic field lines and negative if anti-parallel). The initial increase (decrease) in current is accompanied by a sharp decrease (increase) in loop voltage and the increase in tokamak helicity is consistent with the helicity content of the injected spheromak. In addition, the injection of the spheromak raises the tokamak central density by a factor of six. The introduction of cold spheromak plasma causes sudden cooling of the tokamak discharge from 12 eV to 4 eV which results in a gradual decline in tokamak plasma current by a factor of three. In a second experiment, the authors inject spheromaks into the magnetized toroidal vacuum vessel (with no tokamak plasma). An m = 1 magnetic structure forms in the vessel after the spheromak undergoes a double tilt; once in the cylindrical entrance between gun and tokamak, then again in the tokamak vessel. A horizontal shift of the spheromak equilibrium is observed in the direction opposite that of the static toroidal field. In the absence of net toroidal flux, the structure develops a helical pitch as predicted by theory. Experiments with a number of refractory metal coatings have shown that tungsten and chrome coatings provide some improvement in spheromak parameters. They have also designed and will soon construct a larger, higher current spheromak gun with a new accelerator section for injection experiments on the Phaedrus-T tokamak

  17. Spheromak experiment using separate guns for formation and sustainment

    International Nuclear Information System (INIS)

    Brown, M.R.; Martin, A.

    1996-01-01

    An experiment is described that incorporates the use of separate magnetized plasma guns for formation and sustainment of a spheromak. It is shown that energy coupling efficiency approaches unity if the gun and spheromak are of comparable size. A large gun should be able to operate at lower current and therefore lower voltage. In addition, it is expected that a gun matched to the size of the spheromak will cause less perturbation to the equilibrium. It is proposed to use a smaller gun for spheromak formation and a large, efficient gun for sustainment. The theoretical basis for the experiment is developed, and the details of the experiment are described. A prediction of the equilibrium magnetic flux surfaces using the EFIT code is presented. 28 refs., 3 figs., 1 tab

  18. Ohmic heating of a spheromak to 100 eV

    Energy Technology Data Exchange (ETDEWEB)

    Jarboe, T.R.; Barnes, C.W.; Henins, I.; Hoida, H.W.; Knox, S.O.; Linford, R.K.; Sherwood, A.R.

    1984-01-01

    The first spheromaks with Thomson-scattering-measured electron temperatures of over 100 eV are described. The spheromak is generated by a magnetized coaxial plasma source in a background gas of 30 mTorr of H/sub 2/, and it is stably confined in an oblate 80 cm diam copper mesh flux conserver. The open mesh design allows rapid impurity transport out of the spheromak. The peak temperature, measured using multipoint Thomson scattering, is observed to rise from approximately 25 eV to over 100 eV in about 0.2 msec due to Ohmic heating from the decaying magnetic fields. Density (approx.5 x 10/sup 13/ cm/sup -3/) and magnetic fields (approximately 2 kG) are measured using interferometry and magnetic probes.

  19. Ohmic heating of a spheromak to 100 eV

    International Nuclear Information System (INIS)

    Jarboe, T.R.; Barnes, C.W.; Henins, I.; Hoida, H.W.; Knox, S.O.; Linford, R.K.; Sherwood, A.R.

    1984-01-01

    The first spheromaks with Thomson-scattering-measured electron temperatures of over 100 eV are described. The spheromak is generated by a magnetized coaxial plasma source in a background gas of 30 mTorr of H 2 , and it is stably confined in an oblate 80 cm diam copper mesh flux conserver. The open mesh design allows rapid impurity transport out of the spheromak. The peak temperature, measured using multipoint Thomson scattering, is observed to rise from approximately 25 eV to over 100 eV in about 0.2 msec due to Ohmic heating from the decaying magnetic fields. Density (approx.5 x 10 13 cm -3 ) and magnetic fields (approximately 2 kG) are measured using interferometry and magnetic probes

  20. Simulation of multi-pulse coaxial helicity injection in the Sustained Spheromak Physics Experiment

    Science.gov (United States)

    O'Bryan, J. B.; Romero-Talamás, C. A.; Woodruff, S.

    2018-03-01

    Nonlinear, numerical computation with the NIMROD code is used to explore magnetic self-organization during multi-pulse coaxial helicity injection in the Sustained Spheromak Physics eXperiment. We describe multiple distinct phases of spheromak evolution, starting from vacuum magnetic fields and the formation of the initial magnetic flux bubble through multiple refluxing pulses and the eventual onset of the column mode instability. Experimental and computational magnetic diagnostics agree on the onset of the column mode instability, which first occurs during the second refluxing pulse of the simulated discharge. Our computations also reproduce the injector voltage traces, despite only specifying the injector current and not explicitly modeling the external capacitor bank circuit. The computations demonstrate that global magnetic evolution is fairly robust to different transport models and, therefore, that a single fluid-temperature model is sufficient for a broader, qualitative assessment of spheromak performance. Although discharges with similar traces of normalized injector current produce similar global spheromak evolution, details of the current distribution during the column mode instability impact the relative degree of poloidal flux amplification and magnetic helicity content.

  1. Sustained spheromak technology

    Energy Technology Data Exchange (ETDEWEB)

    Platts, D.A.; Sherwood, A.R.; Jarboe, T.R.; Linford, R.K.; Hoida, H.W.; Henins, I.

    1984-01-01

    The goal of these experiments is to devise a technique for driving a spheromak using dc-powered electrodes. The reduction or elimination of pulsed power components in the spheromak source would result in more attractive reactors, and simpler, cheaper experiments. This is important as experiments get larger and approach reactor size. According to some concepts, the dc spheromak would operate with plasma injection so that it would clean up any impurities produced during its formation. These features make the investigation of dc-powered spheromaks interesting. The questions that need to be answered in this investigation are: (1) can a spheromak be sustained by a dc source; and (2) can a practical source be designed to produce a hot clean plasma. After summarizing the evidence which suggests an answer to question one, the approach being taken to answer question two is discussed.

  2. Sustained spheromak technology

    International Nuclear Information System (INIS)

    Platts, D.A.; Sherwood, A.R.; Jarboe, T.R.; Linford, R.K.; Hoida, H.W.; Henins, I.

    1984-01-01

    The goal of these experiments is to devise a technique for driving a spheromak using dc-powered electrodes. The reduction or elimination of pulsed power components in the spheromak source would result in more attractive reactors, and simpler, cheaper experiments. This is important as experiments get larger and approach reactor size. According to some concepts, the dc spheromak would operate with plasma injection so that it would clean up any impurities produced during its formation. These features make the investigation of dc-powered spheromaks interesting. The questions that need to be answered in this investigation are: (1) can a spheromak be sustained by a dc source; and (2) can a practical source be designed to produce a hot clean plasma. After summarizing the evidence which suggests an answer to question one, the approach being taken to answer question two is discussed

  3. Route to High Temperatures by Current Amplification in the Sustained Spheromak Physics Experiment (SSPX)

    International Nuclear Information System (INIS)

    Woodruff, S.; Holbomb, C. T.; Stallard, B. W.; Hill, D. N.; Hooper, E. B.; McLean, H. S.; Wood, R. D.; Bulmer, R.; Cohen, B.; Sovinec, C.; Pearlstein, L. D.

    2002-01-01

    For the spheromak to be attractive as a reactor concept it would be necessary to sustain the configuration with a low recycling power, reflected in the current amplification factor: A 1 = I tor /I gun , where I tor is the toroidal current and I gun is the gun current. It is understood that A 1 needs to be around 60 for a reactor [1], although the highest obtained so far in the spheromak has been ∼3 [2]. The spheromak is a simply connected toroidal confinement device related to the reversed field pinch in that the q-profile falls at the edge and the first wall is conducting, although the central solenoid is absent. In the spheromak, the paradigm for field generation (and hence current amplification) is the injection of helicity, K = ∫A.BdV = 2ΦΨ where φ and Ψ are linked fluxes. Helicity is additive in the process of electrostatic injection by a coaxial gun [3]: K = 2V gunΨgun , where V gun is the voltage applied between two coaxial electrodes (giving the rate of toroidal flux injection) and Ψ gun is the poloidal vacuum flux connecting them. SSPX [4] is a 1m wide coaxial-gun-driven spheromak with W-coated copper electrodes (FIGURE 1) and a uniquely programmable vacuum field configuration. SSPX was built to assess if confinement can be reasonably preserved during injection, and to address the specific physics of the processes governing helicity injection

  4. Field and current amplification in the SSPX spheromak

    International Nuclear Information System (INIS)

    Hill, D.N. . hilld@llnl.gov; Bulmer, R.H.; Cohen, B.I.

    2003-01-01

    Results are presented from experiments relating to magnetic field generation and current amplification in the SSPX spheromak. The SSPX spheromak plasma is driven by DC coaxial helicity injection using a 2MJ capacitor bank. Peak toroidal plasma currents of up to 0.7MA and peak edge poloidal fields of 0.3T are produced; lower current discharges can be sustained up to 3.5msec. When edge magnetic fluctuations are reduced below 1% by driving the plasma near threshold, it is possible to produce plasmas with Te > 150eV, e >∼4% and core χ e ∼30m 2 /s. Helicity balance for these plasmas suggests that sheath dissipation can be significant, pointing to the importance of maximizing the voltage on the coaxial injector. For most operational modes we find a stiff relationship between peak spheromak field and injector current, and little correlation with plasma temperature, which suggests that other processes than ohmic dissipation may limit field amplification. However, slowing spheromak buildup by limiting the initial current pulse increases the ratio of toroidal current to injected current and points to new operating regimes with more favorable current amplification. (author)

  5. Los Alamos Spheromak Program

    International Nuclear Information System (INIS)

    Knox, S.O.; Barnes, C.W.; Fernandez, J.C.

    1985-01-01

    The Los Alamos Spheromak Program consists of two experimental facilities. The confinement physics of sustained and decaying spheromaks are being studied in CTX, which has an extensive array of diagnostics. Experiments are directed towards extending the physics understanding of the spheromak as a magnetic confinement concept. Electrodes for the production of clean sustained spheromaks are developed on the Electrode Facility, which is more flexible in terms of experimental modifications. Improvements to helicity sources and elecrodes which are proven on the Electrode Facility are then considered for incorporation onto CTX

  6. Spheromak Buildup in SSPX using a Modular Capacitor Bank

    International Nuclear Information System (INIS)

    Wood, R D; McLean, H S; Hill, D N; Hooper, E B; Romero-Talamas, C A

    2006-01-01

    The Sustained Spheromak Physics Experiment (SSPX) [1] was designed to address both magnetic field generation and confinement. The SSPX produces 1.5-3.5msec, spheromak plasmas with a 0.33m major radius and a minor radius of ∼0.23m. DC coaxial helicity injection is used to build and sustain the spheromak plasma within the flux conserver. Optimal operation is obtained by flattening the profile of λ = μ 0 j/B, consistent with reducing the drive for tearing and other MHD modes, and matching of edge current and bias flux to minimize |(delta)B/B| rms . With these optimizations, spheromak plasmas with central T e >350eV and β e ∼ 5% with toroidal fields of 0.6T [3] have been obtained. If a favorable balance between current drive efficiency and energy confinement can be shown, the spheromak has the potential to yield an attractive magnetic fusion concept [4]. The original SSPX power system consists of two lumped-circuit capacitor banks with fixed circuit parameters. This power system is used to produce an initial fast formation current pulse (10kV, 0.5MJ formation bank), followed by a lower current, 3.5ms flattop sustainment pulse (5kV, 1.5MJ sustainment bank). Experimental results indicate that a variety of injected current pulses, such as a longer sustainment flattop [5], higher and longer fast formation [6], and multiple current pulses [7], might further our understanding of magnetic field generation. Although the formation bank can be split into two independent banks capable of producing other injected current waveforms, the variety of current waveforms produced by this power system is limited. Thus, to extend the operating range of the SSPX, a new pulsed-power system has been designed and partially constructed. In this paper, we discuss the design of the programmable bank and present first results from using the bank to increase the magnetic field in SSPX

  7. Fundamental Magnetofluid Physics Studies on the Swarthmore Spheromak Experiment: Reconnection and Sustainment

    International Nuclear Information System (INIS)

    Brown, M.R.

    2001-01-01

    The general goal of the Magnetofluids Laboratory at Swarthmore College is to understand how magnetofluid kinetic energy can be converted to magnetic energy as it is in the core of the earth and sun (the dynamo problem) and to understand how magnetic energy can be rapidly converted back to kinetic energy and heat as it is in solar flares (the magnetic reconnection problem). Magnetic reconnection has been studied using the Swarthmore Spheromak Experiment (SSX) which was designed and built under this Junior Faculty Grant. In SSX we generate and merge two rings of magnetized plasma called spheromaks and study their interaction. The spheromaks have many properties similar to solar flares so this work is directly relevant to basic solar physics. In addition, since the spheromak is a magnetic confinement fusion configuration, issues of formation and stability have direct impact on the fusion program

  8. Progress in the SSPX Spheromak

    International Nuclear Information System (INIS)

    McLean, H S; Woodruff, S; Hill, D N; Bulmer, R H; Cohen, B I; Hooper, E B; Moller, J; Ryutov, D D; Stallard, B W; Wood, R D; Holcomb, C T; Jarboe, T R; Romero-Talamas, C

    2003-01-01

    The spheromak, with its simply connected geometry, holds promise as a less expensive fusion reactor. It has reasonably good plasma beta and can be formed and sustained in steady state with a magnetized coaxial plasma gun. The Sustained Spheromak Physics Experiment (SSPX) shown in Fig. 1 was constructed to investigate the key issues of magnetic field generation and energy confinement. In addition to the coaxial gun, nine magnetic field coils are utilized to shape the vacuum magnetic flux. This flexibility allows operation in many different regimes producing very different plasma characteristics. Pulse length is extended and magnetic field strength is increased. Improved surface conditioning produces plasmas with low impurity content, and higher electron temperature, T e . Electron heat transport within the separatrix is reduced by a factor of 4. The results strongly suggest the existence of closed flux surfaces even though the plasma is connected to the coaxial source. The CORSICA Grad-Shafranov 2-d equilibrium code with data from edge magnetic probes along with T e and electron density ne from Thomson scattering is used to calculate internal profiles: normalized current γ = μ 0 J/B, safety factor = q, ohmic heating, thermal energy density, and thermal diffusivity = ξ e . Ohmic heating is calculated by assuming spatially constant Spitzer resistivity with Z eff =2.3 estimated by VUV spectroscopy

  9. Simulation study of stepwise relaxation in a spheromak plasma

    International Nuclear Information System (INIS)

    Horiuchi, Ritoku; Uchida, Masaya; Sato, Tetsuya.

    1991-10-01

    The energy relaxation process of a spheromak plasma in a flux conserver is investigated by means of a three-dimensional magnetohydrodynamic simulation. The resistive decay of an initial force-free profile brings the spheromak plasma to an m = 1/n = 2 ideal kink unstable region. It is found that the energy relaxation takes place in two steps; namely, the relaxation consists of two physically distinguished phases, and there exists an intermediate phase in between, during which the relaxation becomes inactive temporarily. The first relaxation corresponds to the transition from an axially symmetric force-free state to a helically symmetric one with an n = 2 crescent magnetic island structure via the helical kink instability. The n = 2 helical structure is nonlinearly sustained in the intermediate phase. The helical twisting of the flux tube creates a reconnection current in the vicinity of the geometrical axis. The second relaxation is triggered by the rapid growth of the n = 1 mode when the reconnection current exceeds a critical value. The helical twisting relaxes through magnetic reconnection toward an axially symmetric force-free state. It is also found that the poloidal flux reduces during the helical twisting in the first relaxation and the generation of the toroidal flux occurs through the magnetic reconnection process in the second relaxation. (author)

  10. Multi-pulse power injection and spheromak sustainment in SSPX

    Science.gov (United States)

    Stallard, B. W.; Hill, D. N.; Hooper, E. B.; Bulmer, R. H.; McLean, H. S.; Wood, R. D.; Woodruff, S.; Sspx Team

    2000-10-01

    Lawrence Livermore National Laboratory, Livermore, CA 94550, USA. Spheromak formation (gun injection phase) and sustainment experiments are now routine in SSPX using a multi-bank power system. Gun voltage, impedance, and power coupling show a clear current threshold dependence on gun flux (I_th~=λ_0φ_gun/μ_0), increasing with current above the threshold, and are compared with CTX results. The characteristic gun inductance, L_gun~=0.6 μH, derived from the gun voltage dependence on di/dt, is larger than expected from Corsica modeling of the spheromak equilibrium. It’s value is consistent with the n=1 ‘doughook’ mode structure reported in SPHEX and believed important for helicity injection and toroidal current drive. Results of helicity and power balance calculations of spheromak poloidal field buildup are compared with experiment and used to project sustainment with a future longer pulse power supply. This work was performed under the auspices of US DOE by the University of California Lawrence Livermore National Laboratory under Contract No. W-7405-ENG-48.

  11. The spheromak as a prototype for ultra-high-field superconducting magnets

    International Nuclear Information System (INIS)

    Furth, H.P.; Jardin, S.C.

    1987-08-01

    In view of current progress in the development of superconductor materials, the ultimate high-field limit of superconducting magnets is likely to be set by mechanical stress problems. Maximum field strength should be attainable by means of approximately force-free magnet windings having favorable ''MHD'' stability properties (so that small winding errors will not grow). Since a low-beta finite-flux-hole spheromak configuration qualifies as a suitable prototype, the theoretical and experimental spheromak research effort of the past decade has served to create a substantial technical basis for the design of ultra-high-field superconducting coils. 11 refs

  12. Equilibrium and stability of the Los Alamos spheromak

    International Nuclear Information System (INIS)

    Marklin, G.

    1984-01-01

    The open mesh flux conserver (MFC) on the Los Alamos spheromak (CTX) has been equipped with a large number of Rogowski loops measuring the current in the individual segments of the MFC, providing a complete picture of the surface current pattern induced by the equilibrium and oscillations of the confined plasma. An analysis was made of the data from these Rogowski loops

  13. Controlled and spontaneous magnetic field generation in a gun-driven spheromak

    International Nuclear Information System (INIS)

    Woodruff, S.; Cohen, B.I.; Hooper, E.B.; Mclean, H.S.; Stallard, B.W.; Hill, D.N.; Holcomb, C.T.; Romero-Talamas, C.; Wood, R.D.; Cone, G.; Sovinec, C.R.

    2005-01-01

    In the Sustained Spheromak Physics Experiment, SSPX [E. B. Hooper, D. Pearlstein, and D. D. Ryutov, Nucl. Fusion 39, 863 (1999)], progress has been made in understanding the mechanisms that generate fields by helicity injection. SSPX injects helicity (linked magnetic flux) from 1 m diameter magnetized coaxial electrodes into a flux-conserving confinement region. Control of magnetic fluctuations (δB/B∼1% on the midplane edge) yields T e profiles peaked at >200 eV. Trends indicate a limiting beta (β e ∼4%-6%), and so we have been motivated to increase T e by operating with stronger magnetic field. Two new operating modes are observed to increase the magnetic field: (A) Operation with constant current and spontaneous gun voltage fluctuations. In this case, the gun is operated continuously at the threshold for ejection of plasma from the gun: stored magnetic energy of the spheromak increases gradually with δB/B∼2% and large voltage fluctuations (δV∼1 kV), giving a 50% increase in current amplification, I tor /I gun . (B) Operation with controlled current pulses. In this case, spheromak magnetic energy increases in a stepwise fashion by pulsing the gun, giving the highest magnetic fields observed for SSPX (∼0.7 T along the geometric axis). By increasing the time between pulses, a quasisteady sustainment is produced (with periodic good confinement), comparing well with resistive magnetohydrodynamic simulations. In each case, the processes that transport the helicity into the spheromak are inductive and exhibit a scaling of field with current that exceeds those previously obtained. We use our newly found scaling to suggest how to achieve higher temperatures with a series of pulses

  14. Controlled and Spontaneous Magnetic Field Generation in a Gun-Driven Spheromak

    International Nuclear Information System (INIS)

    Woodruff, S; Cohen, B I; Hooper, E B; McLean, H S; Stallard, B W; Hill, D N; Holcomb, C T; Romero-Talamas, C; Wood, R D; Cone, G; Sovinec, C R

    2005-04-01

    In the Sustained Spheromak Physics Experiment, SSPX, progress has been made in understanding the mechanisms that generate fields by helicity injection. SSPX injects helicity (linked magnetic flux) from 1-m diameter magnetized coaxial electrodes into a flux-conserving confinement region. Control of magnetic fluctuations ((delta)B/B∼1% on the midplane edge) yields T e profiles peaked at > 200eV. Trends indicate a limiting beta (β e ∼ 4-6%), and so we have been motivated to increase T e by operating with stronger magnetic field. Two new operating modes are observed to increase the magnetic field: (A) Operation with constant current and spontaneous gun voltage fluctuations. In this case, the gun is operated continuously at the threshold for ejection of plasma from the gun: stored magnetic energy of the spheromak increases gradually with (delta)B/B ∼2% and large voltage fluctuations ((delta)V ∼ 1kV), giving a 50% increase in current amplification, I tor /I gun . (B) Operation with controlled current pulses. In this case, spheromak magnetic energy increases in a stepwise fashion by pulsing the gun, giving the highest magnetic fields observed for SSPX (∼0.7T along the geometric axis). By increasing the time between pulses, a quasi-steady sustainment is produced (with periodic good confinement), comparing well with resistive MHD simulations. In each case, the processes that transport the helicity into the spheromak are inductive and exhibit a scaling of field with current that exceeds those previously obtained. We use our newly found scaling to suggest how to achieve higher temperatures with a series of pulses

  15. Formation of a field-reversed configuration by coalescence of spheromaks

    International Nuclear Information System (INIS)

    Dasgupta, B.; Sato, Tetsuya; Hayashi, Takaya; Watanabe, Kunihiko; Watanabe, Tomohiko

    1995-01-01

    We present a numerical simulation of the slow formation of FRC by the merging of two spheromaks with opposite toroidal fluxes. A rather important feature of such a method of formation of FRC should be made explicit. A spheromak is basically a Taylor minimum energy state. On the other hand the FRC with its single component poloidal magnetic field and high plasma beta is decidedly far away from a Taylor state. So a numerical simulation of this process, besides demonstrating the feasibility of such FRC formation, is expected to show the traits in the process of transition from a Taylor state to a non-Taylor state. 5 refs., 2 figs., 1 tab

  16. Spheromak type plasma experiment apparatus

    International Nuclear Information System (INIS)

    Odagiri, Kiyoyuki; Miyauchi, Yasuyuki; Oomura, Hiroshi

    1985-01-01

    The fusion power reactor which is expected to be the most promising energy has been developed for several plasma confinement systems. Under these circumstances, Spheromak configuration has recently attracted attention because of its simple structure and efficient plasma confinement. This apparatus was ordered by the Engineering Department of University of Tokyo for basic studies of the Spheromak plasma confinement technologies. This forms Spheromak plasma according to the induction discharge system which injects this plasma with magnetic energy generated by a toroidal current in the plasma and discharges the current through the electrical feed through. Toroidal current is induced by the poloidal coil in the vessel. We worked together with the researchers of University of Tokyo to conduct experiments and confirmed the formation and confinement of Spheromak plasma in the initial test. (author)

  17. Particle confinement and fueling effects on the Maryland spheromak

    International Nuclear Information System (INIS)

    Filuk, A.B.

    1991-01-01

    The spheromak plasma confinement concept provides the opportunity to study the evolution of a nearly force-free magnetic field configuration. The plasma currents and magnetic fields are produced self-consistently, making this type of device attractive as a possible fusion reactor. At present, spheromaks are observed to have poorer particle and magnetic confinement than expected from simple theory. The purpose of this study is to examine the role of plasma density in the decay of spheromaks produced in the Maryland Spheromak experiment. Density measurements are made with an interferometer and Langmuir probe, and results are correlated with those of other plasma diagnostics to understand the sources of plasma, the spheromak formation effects on the density, and the magnitude of particle loss during the spheromak decay. A power and particle balance computer model is constructed and applied to the spheromaks studied in order to assess the impact of high density and particle loss rate on the spheromak decay. The observations and model indicate that the decay of the spheromaks is at present dominated by impurity radiation loss. The model also predicts that high density and short particle confinement time play a critical role in the spheromak power balance when the impurity levels are reduced

  18. Amplification of S-1 Spheromak current by an inductive current transformer

    International Nuclear Information System (INIS)

    Jardin, S.C.; Janos, A.; Yamada, M.

    1985-11-01

    We attempt to predict the consequences of adding an inductive current transformer (OH Transformer) to the present S-1 Spheromak experiment. Axisymmetric modeling with only classical dissipation shows an increase of toroidal current and a shrinking and hollowing of the current channel, conserving toroidal flux. These unstable profiles will undergo helical reconnection, conserving helicity K = ∫ A-vector x B-vector d tau while increasing the toroidal flux and decreasing the poloidal flux so that the plasma relaxes toward the Taylor state. This flux rearrangement is modeled by a new current viscosity term in the mean-field Ohm's law which conserves helicity and dissipates energy

  19. Spheromak Impedance and Current Amplification

    International Nuclear Information System (INIS)

    Fowler, T K; Hua, D D; Stallard, B W

    2002-01-01

    It is shown that high current amplification can be achieved only by injecting helicity on the timescale for reconnection, τ REC , which determines the effective impedance of the spheromak. An approximate equation for current amplification is: dI TOR 2 /dt ∼ I 2 /τ REC - I TOR 2 /τ closed where I is the gun current, I TOR is the spheromak toroidal current and τ CLOSED is the ohmic decay time of the spheromak. Achieving high current amplification, I TOR >> I, requires τ REC CLOSED . For resistive reconnection, this requires reconnection in a cold zone feeding helicity into a hot zone. Here we propose an impedance model based on these ideas in a form that can be implemented in the Corsica-based helicity transport code. The most important feature of the model is the possibility that τ REC actually increases as the spheromak temperature increases, perhaps accounting for the ''voltage sag'' observed in some experiments, and a tendency toward a constant ratio of field to current, B ∝ I, or I TOR ∼ I. Program implications are discussed

  20. Conceptual design of a cassette compact toroid reactor (the zero-phase study) - Quick replacement of the reactor core

    International Nuclear Information System (INIS)

    Nishikawa, M.; Narikawa, T.; Iwamoto, M.; Watanabe, K.

    1986-01-01

    A study of a conceptual design for a ''cassette'' compact toroid reactor has been performed that emphasizes quick replacement handling. The core plasma, spheromak, is ohmically heated in a merging process between the core plasma and the gun-produced spheromak. The quick handling of replacement accomplished by using a functional material, a shape memory alloy (SMA) joint, which is proposed for release from first-wall high neutron loading in a newly devised mechanical and structural method. The SMA joint can be used for connecting or disconnecting the coupling by simply controlling the SMA temperature without the need for a robot system. Effective heat removal from the first wall and thermal and electromagnetic stress in a fusion core with very high heat flux are discussed from an engineering standpoint

  1. 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, ...

  2. Stellarmak a hybrid stellarator: Spheromak

    International Nuclear Information System (INIS)

    Hartman, C.W.

    1980-01-01

    This paper discusses hybridization of modified Stellarator-like transform windings (T-windings) with a Spheromak or Field-Reversed-Mirror configuration. This configuration, Stellarmak, retains the important topological advantage of the Spheromak or FRM of having no plasma linking conductors or blankets. The T-windings provide rotational transformation in toroidal angle of the outer poloidal field lines, in effect creating a reversed B/sub Toroidal/ Spheromak or adding average B/sub T/ to the FRM producing higher shear, increased limiting β, and possibly greater stability to kinks and tilt. The presence of field ripple in the toroidal direction may be sufficient to inhibit cancellation of directed ion current by electron drag to allow steady state operation with the toroidal as well as poloidal current maintained by neutral beams

  3. Ion temperature measurements in the Maryland Spheromak

    International Nuclear Information System (INIS)

    Gauvreau, J.L.

    1992-01-01

    Initial spectroscopic data from MS showed evidence of ion heating as deduced from the line widths of different ion species. Detailed measurements of OIV spectral emission line profiles in space and time revealed that heating takes place at early time, before spheromak formation and is occurring within the current discharge. The measured ion temperature is several times the electron temperature and cannot be explained by classical (Spitzer) resistivity. Classically, ions are expected to have lower temperatures than the electrons and therefore, lower temperatures than observed. High ion temperatures have been observed in different RFP's and Spheromaks but are usually associated with relaxation to the Taylor state and occur in the sustainment phase. During formation, the current delivered to start the discharge is not axisymmetric and as a consequence, X-points appear in the magnetic flux. A two dimensional analysis predicts that magnetic reconnection occurring at an X-point can give rise to high ion heating rates. A simple 0-dimensional calculation showed that within the first 20 μs, a conversion of mass flow kinetic energy into ion temperature could take place due to viscosity

  4. Development of the STPX Spheromak System

    Science.gov (United States)

    Williams, R. L.; Clark, J.; Weatherford, C. A.

    2015-11-01

    The progress made in starting up the STPX Spheromak system, which is now installed at the Florida A&M University, is reviewed. Experimental, computational and theoretical activities are underway. The control system for firing the magnetized coaxial plasma gun and for collecting data from the diagnostic probes, based on LabView, is being tested and adapted. Preliminary results of testing the installed magnetic field probes, Langmuir triple probes, cylindrical ion probes, and optical diagnostics will be discussed. Progress in modeling this spheromak using simulation codes, such as NIMROD, will be discussed. Progress in investigating the use of algebraic topology to describe this spheromak will be reported.

  5. Instantaneous current and field structure of a gun-driven spheromak for two gun polarities

    International Nuclear Information System (INIS)

    Woodruff, S; Nagata, M

    2002-01-01

    The instantaneous plasma structure of the SPHEX spheromak is determined here by numerically processing data from insertable Rogowski and magnetic field probes. Data is presented and compared for two modes of gun operation: with the central electrode biased positively and negatively. It is found that while the mean-, or even instantaneous-, field structure would give the impression of a roughly axisymmetric spheromak, the instantaneous current structure does not. Hundred per cent variations in J measured at the magnetic axis can be explained by the rotation of a current filament that has a width equal to half of the radius of the flux-conserving first wall. In positive gun operation, current leaves the filament in the confinement region leading to high wall current there. In negative gun operation, wall current remains low as all injected current returns to the gun through the plasma. The plasma, in either instance, is strongly asymmetric. We discuss evidence for the existence of the current filament in other gun-driven spheromaks and coaxial plasma thrusters

  6. The spheromak as a compact fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hagenson, R.L.; Krakowski, R.A.

    1987-03-01

    After summarizing the economic and utility-based rationale for compact, higher-power-density fusion reactors, the gun-sustained spheromak concept is explored as one of a number of poloidal-field-dominated confinement configurations that might improve the prospects for economically attractive and operationally simplified fusion power plants. Using a comprehensive physics/engineering/costing model for the spheromak, guided by realistic engineering constraints and physics extrapolation, a range of cost-optimized reactor design points is presented, and the sensitivity of cost to key physics, engineering, and operational variables is reported. The results presented herein provide the basis for conceptual engineering designs of key fusion-power-core (FPC) subsystems and more detailed plasma modeling of this promising, high mass-power-density concept, which stresses single-piece FPC maintenance, steady-state current drive through electrostatic magnetic helicity injection, a simplified co-axial electrode-divertor, and efficient resistive-coal equilibrium-field coils. The optimal FPC size and the cost estimates project a system that competes aggressively with the best offered by alternative energy sources while simplifying considerably the complexity that has generally been associated with most approaches to magnetic fusion energy.

  7. The spheromak as a compact fusion reactor

    International Nuclear Information System (INIS)

    Hagenson, R.L.; Krakowski, R.A.

    1987-03-01

    After summarizing the economic and utility-based rationale for compact, higher-power-density fusion reactors, the gun-sustained spheromak concept is explored as one of a number of poloidal-field-dominated confinement configurations that might improve the prospects for economically attractive and operationally simplified fusion power plants. Using a comprehensive physics/engineering/costing model for the spheromak, guided by realistic engineering constraints and physics extrapolation, a range of cost-optimized reactor design points is presented, and the sensitivity of cost to key physics, engineering, and operational variables is reported. The results presented herein provide the basis for conceptual engineering designs of key fusion-power-core (FPC) subsystems and more detailed plasma modeling of this promising, high mass-power-density concept, which stresses single-piece FPC maintenance, steady-state current drive through electrostatic magnetic helicity injection, a simplified co-axial electrode-divertor, and efficient resistive-coal equilibrium-field coils. The optimal FPC size and the cost estimates project a system that competes aggressively with the best offered by alternative energy sources while simplifying considerably the complexity that has generally been associated with most approaches to magnetic fusion energy

  8. Computational studies of ohmic heating in the spheromak

    International Nuclear Information System (INIS)

    Olson, R.E.

    1983-01-01

    Time-dependent computational simulations using both single-fluid O-D and two-fluid 1 1/2-D models are developed for and utilized in an investigation of the ohmic heating of a spheromak plasma. The plasma density and composition, the applied magnetic field strength, the plasma size, and the plasma current density profile are considered for their effects on the spheromak heating rate and maximum achievable temperature. The feasibility of ohmic ignition of a reactor-size spheromak plasma is also contemplated

  9. Final Report Sustained Spheromak Physics Project FY 1997 - FY 1999

    International Nuclear Information System (INIS)

    Hooper, E.B.; Hill, D.N.

    2000-01-01

    This is the final report on the LDRD SI-funded Sustained Spheromak Physics Project for the years FY1997-FY1999, during which the SSPX spheromak was designed, built, and commissioned for operation at LLNL. The specific LDRD project covered in this report concerns the development, installation, and operation of specialized hardware and diagnostics for use on the SSPX facility in order to study energy confinement in a sustained spheromak plasma configuration. The USDOE Office of Fusion Energy Science funded the construction and routine operation of the SSPX facility. The main distinctive feature of the spheromak is that currents in the plasma itself produce the confining toroidal magnetic field, rather than external coils, which necessarily thread the vacuum vessel. There main objective of the Sustained Spheromak Physics Project was to test whether sufficient energy confinement could be maintained in a spheromak plasma sustained by DC helicity injection. Achieving central electron temperatures of several hundred eV would indicate this. In addition, we set out to determine how the energy confinement scales with T c and to relate the confinement time to the level of internal magnetic turbulence. Energy confinement and its scaling are the central technical issues for the spheromak as a fusion reactor concept. Pending the outcome of energy confinement studies now under way, the spheromak could be the basis for an attractive fusion reactor because of its compact size, simply-connected magnetic geometry, and potential for steady-state current drive

  10. On the jets, kinks, and spheromaks formed by a planar magnetized coaxial gun

    International Nuclear Information System (INIS)

    Hsu, S.C.; Bellan, P.M.

    2005-01-01

    Measurements of the various plasma configurations produced by a planar magnetized coaxial gun provide insight into the magnetic topology evolution resulting from magnetic helicity injection. Important features of the experiments are a very simple coaxial gun design so that all observed geometrical complexity is due to the intrinsic physical dynamics rather than the source shape and use of a fast multiple-frame digital camera which provides direct imaging of topologically complex shapes and dynamics. Three key experimental findings were obtained: (1) formation of an axial collimated jet [Hsu and Bellan, Mon. Not. R. Astron. Soc. 334, 257 (2002)] that is consistent with a magnetohydrodynamic description of astrophysical jets (2) identification of the kink instability when this jet satisfies the Kruskal-Shafranov limit, and (3) the nonlinear properties of the kink instability providing a conversion of toroidal to poloidal flux as required for spheromak formation by a coaxial magnetized source [Hsu and Bellan, Phys. Rev. Lett. 90, 215002 (2003)]. An interpretation is proposed for how the n=1 central column instability provides flux amplification during spheromak formation and sustainment, and it is shown that jet collimation can occur within one rotation of the background poloidal field

  11. Studies of conceptual spheromak fusion reactors

    International Nuclear Information System (INIS)

    Katsurai, M.; Yamada, M.

    1982-01-01

    Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain the characteristic relations among various parameters of the spheromak configuration for an aspect ratio of A >or approx. 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) the DT reactor with two-component-type operation, (2) the ignited DT reactor, and (3) the ignited catalysed-type DD reactor. With a total wall loading of approx. 4 MW.m -2 , it is found that edge magnetic fields of only approx. 4 T (DT) and approx. 9 T (Cat. DD) are required for ignited reactors of 1 m plasma (minor) radius with output powers in the gigawatt range. An assessment of various schemes of generation, compression and translation of spheromak plasmas is presented. (author)

  12. HIFLUX: OBLATE FRCS, DOUBLE HELICES,SPHEROMAKS AND RFPS IN ONE SYSTEM

    International Nuclear Information System (INIS)

    SCHAFFER, M.J.; BOEDO, J.A.

    2003-01-01

    OAK-B135 High magnetic flux is required for thermonuclear FRC reactors and, more immediately, to advance the FRC experimental program in general. Oblate FRCs are of special interest because they are predicted to have certain improved MHD stability over elongated FRCs, and oblate FRCs may yield the most compact, magnetically confined fusion reactors. Neither oblate nor high-flux FRCs have been investigated experimentally to date. Our presently proposed technique is to make two high-flux, oppositely-handed plasmas by a pair of large, external, reversed-field pinch (RFP) sources. The plasmas would propagate as two Taylor-relaxed double-helix plasmas, to an oblate main plasma chamber, where they would relax further to a counter-helicity pair of spheromaks, which would finally merge into a single high-flux FRC. A concept for a new experimental facility, HIFLUX, to make and study high-magnetic-flux oblate Field-Reversed Configuration (FRC) plasmas, is described. Similar principles might also enable high flux non-inductive startup of other plasma devices

  13. Advanced spheromak fusion reactor

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1996-01-01

    The spheromak has no toroidal magnetic field coils or other structure along its geometric axis, and is thus more attractive than the leading magnetic fusion reactor concept, the tokamak. As a consequence of this and other attributes, the spheromak reactor may be compact and produce a power density sufficiently high to warrant consideration of a liquid 'blanket' that breeds tritium, converts neutron kinetic energy to heat, and protects the reactor vessel from severe neutron damage. However, the physics is more complex, so that considerable research is required to learn how to achieve the reactor potential. Critical physics problems and possible ways of solving them are described. The opportunities and issues associated with a possible liquid wall are considered to direct future research

  14. The impedance of energy efficiency of a coaxial magnetized plasma source used for spheromak formation and sustainment

    International Nuclear Information System (INIS)

    Barnes, C.W.; Jarboe, T.R.; Marklin, G.J.; Knox, S.O.; Henins, I.

    1989-01-01

    Electrostatic (dc) helicity injection has previously been shown to successfully sustain the magnetic fields of spheromaks and tokamaks. The magnitude of the injected magnetic helicity balances (within experimental error) the flux lost be resistive decay of the toroidal equilibrium. The problem of optimizing this current drive scheme hence involves maximizing the injected helicity (the voltage-connecting-flux product) while minimizing the current (which multiplied by the voltage represents the energy input and also possible damage to the electrodes). The impedance (voltage-to-current ratio) and energy efficiency of a dc helicity injection experiment are studied on the CTX spheromak. Over several years changes were made in the physical geometry of the coaxial magnetized plasmas source as well as changes in the external electrical circuit. The source could be operated over a wide range of external charging voltage (and hence current), applied axial flux, and source gas flow rate. A database of resulting voltage, helicity injection, efficiency, electron density, and rotation has been created. These experimental results are compared to an ideal magnetohydrodynamic theory of magnetic flux flow. The theory is parameterized by the dimensionless Hall parameter, the ratio of electric to mass current. For a constant Hall parameter the theory explains why the voltage depends quadratically on the current at constant flux. The theory also explains the approximately linear dependence of the impedance-to-current ratio on the current-to-flux ratio of the source. 9 refs., 6 figs

  15. Spheromak formation studies in SSPX

    International Nuclear Information System (INIS)

    Hill, D.N.; Bulmer, R.H.; Cohen, B.L.; Hooper, E.B.; LoDestro, L.L.; Mattor, N.; McLean, H.S.; Moller, J.; Pearlstein, L.D.; Ryutov, D.D.; Stallard, B.W.; Wood, R.D.; Woodruff, S.; Holcomb, C.T.; Jarboe, T.; Sovinec, C.R.; Wang, Z.; Wurden, G.

    2000-01-01

    We present results from the Sustained Spheromak Physics Experiment (SSPX) at LLNL, which has been built to study energy confinement in spheromak plasmas sustained for up to 2 ms by coaxial DC helicity injection. Peak toroidal currents as high as 600kA have been obtained in the 1m dia. (0.23m minor radius) device using injection currents between 200-400kA; these currents generate edge poloidal fields in the range of 0.2-0.4T. The internal field and current profiles are inferred from edge field measurements using the CORSICA code. Density and impurity control is obtained using baking, glow discharge cleansing, and titanium gettering, after which long plasma decay times (τ (ge) 1.5ms) are observed and impurity radiation losses are reduced from ∼50% to e (0)∼120eV and β e ∼7%. Edge field measurements show the presence of n=1 modes during the formation phase, as has been observed in other spheromaks. This mode dies away during sustainment and decay so that edge fluctuation levels as low as 1% have been measured. These results are compared with numerical simulations using the NIMROD code

  16. A new diagnostic for spheromaks

    International Nuclear Information System (INIS)

    Boyd, D.A.

    1986-01-01

    Electron cyclotron emission from a spheromak plasma may be able to provide information about the confining magnetic field of the system. Emission generated in the extraordinary mode wit hits electric vector perpendicular to the local magnetic field at sufficiently high frequency will propagate out of the plasma while retaining the original orientation if its electric vector. Thus, a measurement of the orientation of the emergent electric vector and the emission frequency will allow one to deduce the orientation and strength of the magnetic field at the radiation source. In this paper, simple models of the Maryland spheromak are used to examine the practicality of such a diagnostic

  17. All plasma spheromak: the plasmak

    International Nuclear Information System (INIS)

    Koloc, P.; Ogden, J.

    1981-01-01

    There has been an evolutionary pattern established in magnetic fusion concepts. The flow in ideas follows three directions. By extrapolating this evolutionary movement, we have anticipated the concept called Spheromak and have predicted the omega of this evolution which is called PLASMAK, or Plasma Spheromak. The evolutionary directions are from open systems to closed systems, from zero or low dimensional compression schemes to three dimensional compression, and finally from plasma configurations without any self confining currents to a plasma configuration which is completely self confined except for the mechanical pressure necessary to maintain the verticle field and hoop stress. Nevertheless, the plasma is imprisoned by heavy poloidal coils and a vacuum wall

  18. Design of a Modular E-Core Flux Concentrating Axial Flux Machine: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Husain, Tausif; Sozer, Yilmaz; Husain, Iqbal; Muljadi, Eduard

    2015-08-24

    In this paper a novel E-Core axial flux machine is proposed. The machine has a double-stator, single-rotor configuration with flux-concentrating ferrite magnets and pole windings across each leg of an E-Core stator. E-Core stators with the proposed flux-concentrating rotor arrangement result in better magnet utilization and higher torque density. The machine also has a modular structure facilitating simpler construction. This paper presents a single-phase and a three-phase version of the E-Core machine. Case studies for a 1.1-kW, 400-rpm machine for both the single-phase and three-phase axial flux machines are presented. The results are verified through 3D finite element analysis. facilitating simpler construction. This paper presents a single-phase and a three-phase version of the E-Core machine. Case studies for a 1.1-kW, 400-rpm machine for both the single-phase and three-phase axial flux machines are presented. The results are verified through 3D finite element analysis.

  19. Stellarator-Spheromak

    International Nuclear Information System (INIS)

    Moroz, P.E.

    1997-03-01

    A novel concept for magnetic plasma confinement, Stellarator-Spheromak (SSP), is proposed. Numerical analysis with the classical-stellarator-type outboard stellarator windings demonstrates a number of potential advantages of SSP for controlled nuclear fusion. Among the main ones are: simple and compact magnet coil configuration, absence of material structures (e.g. magnet coils or conducting walls) in the center of the torus, high rotational transform, and a possibility of MHD equilibria with very high β (pressure/magnetic pressure) of the confined plasma

  20. Flux distribution in single phase, Si-Fe, wound transformer cores

    International Nuclear Information System (INIS)

    Loizos, George; Kefalas, Themistoklis; Kladas, Antonios; Souflaris, Thanassis; Paparigas, Dimitris

    2008-01-01

    This paper shows experimental results of longitudinal flux density and its harmonics at the limb, the yoke and the corner as well as normal flux in the step lap joint of a single phase, Si-Fe, wound transformer core. Results show that the flux density as well as the harmonics content is higher in the inner (window) side of the core and reduces gradually towards the outer side. Variations of flux density distribution between the limb and the corner or the yoke of the core were observed. A full record of normal flux around the step lap region of the model core was also obtained. Longitudinal and normal flux findings will enable the development of more accurate numerical models that describe the magnetic behavior of magnetic cores

  1. Current drive by neutral beams, rotating magnetic fields and helicity injection in compact toroids

    International Nuclear Information System (INIS)

    Farengo, R.; Arista, N.R.; Lifschitz, A.F.; Clemente, R.A.

    2003-01-01

    The use of neutral beams (NB) for current drive and heating in spheromaks, the relaxed states of flux core spheromaks (FCS) sustained by helicity injection and the effect of ion dynamics on rotating magnetic field (RMF) current drive in spherical tokamaks (ST) are studied. (author)

  2. Selective decay in a helicity-injected spheromak

    International Nuclear Information System (INIS)

    MartInez, P L Garcia; Farengo, R

    2009-01-01

    The non-linear evolution of several unstable equilibria, representative of helicity-injected spheromak configurations inside a cylindrical flux conserver, is studied by means of three dimensional resistive MHD simulations. These equilibria are force-free (∇ x B = λ(ψ)B) but do not correspond to minimum energy states, having linear λ(ψ) profiles with negative slope. Several aspects of this process are studied (magnetic energy relaxation, selective helicity decay, relaxed profiles) for different initial A slopes. The stability threshold predicted by linear theory is recovered. The results show that complete plasma relaxation leading to a uniform A, is achieved only if the initial profile is hollow enough. The evolution for cases just above the stability threshold is more gentle and does not end in a Taylor state. The final state in these cases has a linear λ(ψ) profile, as the initial condition, but with a smaller slope.

  3. Global magnetic fluctuations in S-1 spheromak plasmas and relaxation toward a minimum-energy state

    International Nuclear Information System (INIS)

    Janos, A.; Hart, G.W.; Yamada, M.

    1986-01-01

    Globally coherent modes have been observed during formation in the S-1 Spheromak plasma. These modes play an important role in flux conversion and plasma relaxation toward a minimum-energy state. A significant finding is the temporal progression through the n = 5, 4, 3, 2; m = 1 mode sequence as q rises through rational fractions m/n. Peak amplitudes of the modes relative to the unperturbed field are typically less than 5%, while amplitudes as high as 20% have been observed

  4. 3-D MHD modeling and stability analysis of jet and spheromak plasmas launched into a magnetized plasma

    Science.gov (United States)

    Fisher, Dustin; Zhang, Yue; Wallace, Ben; Gilmore, Mark; Manchester, Ward; Arge, C. Nick

    2016-10-01

    The Plasma Bubble Expansion Experiment (PBEX) at the University of New Mexico uses a coaxial plasma gun to launch jet and spheromak magnetic plasma configurations into the Helicon-Cathode (HelCat) plasma device. Plasma structures launched from the gun drag frozen-in magnetic flux into the background magnetic field of the chamber providing a rich set of dynamics to study magnetic turbulence, force-free magnetic spheromaks, and shocks. Preliminary modeling is presented using the highly-developed 3-D, MHD, BATS-R-US code developed at the University of Michigan. BATS-R-US employs an adaptive mesh refinement grid that enables the capture and resolution of shock structures and current sheets, and is particularly suited to model the parameter regime under investigation. CCD images and magnetic field data from the experiment suggest the stabilization of an m =1 kink mode trailing a plasma jet launched into a background magnetic field. Results from a linear stability code investigating the effect of shear-flow as a cause of this stabilization from magnetic tension forces on the jet will be presented. Initial analyses of a possible magnetic Rayleigh Taylor instability seen at the interface between launched spheromaks and their entraining background magnetic field will also be presented. Work supported by the Army Research Office Award No. W911NF1510480.

  5. MHD stability analysis of axisymmetric surface current model tokamaks close to the spheromak regime

    International Nuclear Information System (INIS)

    Honma, Toshihisa; Kaji, Ikuo; Fukai, Ichiro; Kito, Masafumi.

    1984-01-01

    In the toroidal coordinates, a stability analysis is presented for very low-aspect-ratio tokamaks with circular cross section which is described by a surface current model (SCM) of axisymmetric equilibria. The energy principle determining the stability of plasma is treated without any expansion of aspect ratio. Numerical results show that, owing to the occurrence of the non-axisymmetric (n=1) unstable modes, there exists no MHD-stable ideal SCM spheromak characterized by zero external toroidal vacuum field. Instead, a stable spheromak-type plasma which comes to the ideal SCM spheromak is provided by the configuration with a very weak external toroidal field. Close to the spheromak regime (1.0 1 aspect ratio< = 1.1), the minimum safety factor and the critical β-values increase mo notonically with aspect ratio decreasing from a large value, and curves of βsub(p) versus β in the marginal stability approach to an ideal SCM spheromak line βsub(p)=β. (author)

  6. Steady-state spheromak reactor studies

    International Nuclear Information System (INIS)

    Krakowski, R.A.; Hagenson, R.L.

    1985-01-01

    After summarizing the essential elements of a gun-sustained spheromak, the potential for a steady-state is explored by means of a comprehensive physics/engineering/costing model. A range of cost-optimized reactor design points is presented, and the sensitivity of cost to key physics, engineering, and operational variables is reported

  7. Measurement and simulation of thermal neutron flux distribution in the RTP core

    Science.gov (United States)

    Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

    2018-01-01

    The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

  8. Recent results from the Los Alamos CTX spheromak

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, C.W.; Henins, I.; Hoida, H.W.; Jarboe, T.R.; Knox, S.O.; Linford, R.K.; Platts, D.A.; Sherwood, A.R.

    1982-01-01

    Continued discharge cleaning, improved vacuum practices, and optimized plasma formation operation have resulted in the Los Alamos CTX spheromak experiment achieving 1 millisecond plasma lifetimes with average temperatures of 20 to 40 eV. Impurity radiation power loss has been reduced significantly and the plasma behavior appears to be dominated by pressure-driven instabilities causing increased particle loss. The major advance in operation has been the use of a constant, uniform background of 5 to 20 mTorr of H/sub 2/ filling the vacuum tank, flux conserver, and plasma source. This fill operation directly reduces the impurities generated in the plasma source, allows operation of the source at parameters resulting in fewer impurities, and provides a neutral source to maintain the density for long lifetimes. In this paper we present data on the improved operation of CTX, and present evidence for its ..beta..-limited operation.

  9. Recent results from the Los Alamos CTX spheromak

    International Nuclear Information System (INIS)

    Barnes, C.W.; Henins, I.; Hoida, H.W.; Jarboe, T.R.; Knox, S.O.; Linford, R.K.; Platts, D.A.; Sherwood, A.R.

    1982-01-01

    Continued discharge cleaning, improved vacuum practices, and optimized plasma formation operation have resulted in the Los Alamos CTX spheromak experiment achieving 1 millisecond plasma lifetimes with average temperatures of 20 to 40 eV. Impurity radiation power loss has been reduced significantly and the plasma behavior appears to be dominated by pressure-driven instabilities causing increased particle loss. The major advance in operation has been the use of a constant, uniform background of 5 to 20 mTorr of H 2 filling the vacuum tank, flux conserver, and plasma source. This fill operation directly reduces the impurities generated in the plasma source, allows operation of the source at parameters resulting in fewer impurities, and provides a neutral source to maintain the density for long lifetimes. In this paper we present data on the improved operation of CTX, and present evidence for its β-limited operation

  10. Power balance and characterization of impurities in the Maryland Spheromak

    International Nuclear Information System (INIS)

    Cote, C.

    1993-01-01

    The Maryland Spheromak is a medium size magnetically confined plasma of toroidal shape. Low T e and higher n e than expected contribute to produce a radiation dominated short-lived spheromak configuration. A pyroelectric radiation detector and a VUV spectrometer have been used for space and time-resolved measurements of radiated power and impurity line emission. Results from the bolometry and VUV spectroscopy diagnostics have been combined to give the absolute concentrations of the major impurity species together with the electron temperature. The large amount of oxygen and nitrogen ions in the plasma very early in the discharge is seen to be directly responsible for the abnormally high electron density. The dominant power loss mechanisms are found to be radiation (from impurity line emission) and electron convection to the end walls during the formation phase of the spheromak configuration, and radiation only during the decay phase

  11. Power balance and characterization of impurities in the Maryland Spheromak

    Energy Technology Data Exchange (ETDEWEB)

    Cote, Claude [Univ. of Maryland, College Park, MD (United States)

    1993-01-01

    The Maryland Spheromak is a medium size magnetically confined plasma of toroidal shape. Low Te and higher ne than expected contribute to produce a radiation dominated short-lived spheromak configuration. A pyroelectric radiation detector and a VUV spectrometer have been used for space and time-resolved measurements of radiated power and impurity line emission. Results from the bolometry and VUV spectroscopy diagnostics have been combined to give the absolute concentrations of the major impurity species together with the electron temperature. The large amount of oxygen and nitrogen ions in the plasma very early in the discharge is seen to be directly responsible for the abnormally high electron density. The dominant power loss mechanisms are found to be radiation (from impurity line emission) and electron convection to the end walls during the formation phase of the spheromak configuration, and radiation only during the decay phase.

  12. Measurement of magnetic properties of confined compact toroid plasma (spheromak)

    International Nuclear Information System (INIS)

    Hwang, Fu-Kwun.

    1991-01-01

    The theoretical aspect of the spheromak is described in this paper. The MS machine hardware will be explored along with the formation scheme and diagnostic systems. The magnetic pickup probes, their calibration procedures and the data analysis methods will be discussed. Observations from the probe measurements and magnetic properties of the MS spheromak are considered. The axisymmetric Grad-Shafranov equilibrium code calculations are presented and compared with the measurements. Magnetic helicity and its correlation with the experimental observations is described

  13. Transport and fluctuations in high temperature spheromak plasmas

    International Nuclear Information System (INIS)

    McLean, H.S.; Wood, R.D.; Cohen, B.I.; Hooper, E.B.; Hill, D.N.; Moller, J.M.; Romero-Talamas, C.; Woodruff, S.

    2006-01-01

    Higher electron temperature (T e >350 eV) and reduced electron thermal diffusivity (χ e 2 /s) is achieved in the Sustained Spheromak Physics Experiment (SSPX) by increasing the discharge current=I gun and gun bias flux=ψ gun in a prescribed manner. The internal current and q=safety factor profile derived from equilibrium reconstruction as well as the measured magnetic fluctuation amplitude can be controlled by programming the ratio λ gun =μ 0 I gun /ψ gun . Varying λ gun above and below the minimum energy eigenvalue=λ FC of the flux conserver (∇xB-vector=λ FC B-vector) varies the q profile and produces the m/n=poloidal/toroidal magnetic fluctuation mode spectrum expected from mode-rational surfaces with q=m/n. The highest T e is measured when the gun is driven with λ gun slightly less than λ FC , producing low fluctuation amplitudes ( e as T e increases, differing from Bohm or open field line transport models where χ e increases with T e . Detailed resistive magnetohydrodynamic simulations with the NIMROD code support the analysis of energy confinement in terms of the causal link with the q profile, magnetic fluctuations associated with low-order mode-rational surfaces, and the quality of magnetic surfaces

  14. 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...

  15. Sustained spheromak coaxial gun operation in the presence of an n=1 magnetic distortion

    International Nuclear Information System (INIS)

    Holcomb, C.T.; Jarboe, T.R.; Hill, D.N.; Woodruff, S.; Wood, R.D.

    2006-01-01

    The Sustained Spheromak Physics Experiment (SSPX) [E. B. Hooper, L. D. Pearlstein, and R. H. Bulmer, Nucl. Fusion 39, 863 (1999)] uses a magnetized coaxial gun to form and sustain spheromaks by helicity injection. Internal probes give the magnetic profile within the gun. An analysis of these data show that a number of commonly applied assumptions are not completely correct, and some previously unrecognized processes may be at work. Specifically, the fraction of the available vacuum flux spanning the gun that is stretched out of the gun is variable and not usually 100%. The n=1 mode that is present during sustained discharges has its largest value of δB/B within the gun, so that instantaneously B within the gun is not axisymmetric. By applying a rigid-rotor model to account for the mode, the instantaneous field and current structure within the gun are determined. The current density is also highly nonaxisymmetric and the local value of λ≡μ 0 j parallel /B is not constant, although the global value λ g ≡μ 0 I g /ψ g closely matches that expected by axisymmetric models. The current distribution near the gun muzzle suggests a cross-field current exists, and this is explained as a line-tying reaction to plasma rotation

  16. Magnetic structure in the entrance region of spheromaks sustained by a magnetized coaxial plasma gun under long pulse operation

    International Nuclear Information System (INIS)

    Amemiya, Naoyuki; Takaichi, Kazuaki; Katsurai, Makoto

    1989-01-01

    The magnetic structure in coaxial-gun-sustained spheromaks has been investigated. The plasma gun has been operated with a small axial/radial bias magnetic flux as compared to the azimuthal magnetic flux produced by the discharge current. Stronger magnetic field is observed in the entrance region (ER) than in the flux conserver (FC). In both ER and FC, the magnetic structure is nearly axisymmetric. The axial magnetic field in ER is amplified up to about sixteen times as large as the bias magnetic field. This amplification is limited by the drastic change in the magnetic structure, which occurs when the discharge current becomes very large. The magnetic structure before the drastic change is interpreted with the Bessel function model. The μ estimation shows that the magnetic structure is mainly determined by the boundary geometry, not by the external magnetic flux and current. (author)

  17. Prospects for spheromak fusion reactors

    International Nuclear Information System (INIS)

    Fowler, T.K.; Hua, D.D.

    1995-01-01

    The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on physics principles confirmed in CTX experiments in many respects. Most uncertain was the energy confinement time and the role of magnetic turbulence inherent in the concept. In this paper, a one-dimensional model of heat confinement, calibrated by CTX, predicts negligible heat loss by magnetic turbulence at reactor scale

  18. Confinement requirements for OHMIC-compressive ignition of a Spheromak plasma

    International Nuclear Information System (INIS)

    Olson, R.; Gilligan, J.; Miley, G.

    1980-01-01

    The Moving Plasmoid Reactor (MPR) is an attractive alternative magnetic fusion scheme in which Spheromak plasmoids are envisioned to be formed, compressed, burned, and expanded as the plasmoids translate through a series of linear reactor modules. Although auxiliary heating of the plasmoids may be possible, the MPR scenario would be especially interesting if ohmic decay and compression along were sufficient to heat the plasmoids to an ignition temperature. In the present work, we will study the transport conditions under which a Spheromak plasmoid could be expected to reach ignition via a combination of ohmic and compression heating

  19. Confinement requirements for ohmic-compressive ignition of a Spheromak plasma

    International Nuclear Information System (INIS)

    Olson, R.E.; Miley, G.H.

    1981-01-01

    The Moving Plasmoid Reactor (MPR) is an attractive alternative magnetic fusion scheme in which Spheromak plasmoids are envisioned to be formed, compressed, burned, and expanded as the plasmoids translate through a series of linear reactor modules. Although auxiliary heating of the plasmoids may be possible, the MPR scenario would be especially interesting if ohmic decay and compression alone is sufficient to heat the plasmoids to an ignition temperature. In the present work, we examine the transport conditions under which a Spheromak plasmoid can be expected to reach ignition via a combination of ohmic and compression heating

  20. Relaxation and particle diffusion in the Proto S-1/C spheromak

    International Nuclear Information System (INIS)

    Meyerhofer, D.D.

    1987-01-01

    The relationship between relaxation and particle diffusion in the Proto S-1C spheromak has been studied. The plasma was formed in a magnetic configuration which was not the minimum-energy Taylor state, and went through a period of relaxation before its magnetic configuration was that of the Taylor state. Early in the relaxation phase, the internal and external magnetic fluctuations were correlated and it was found that, at the time of peak amplitude, they had a radial structure of a tearing mode. After the reconnection of these modes, the plasma continued to evolve towards the Taylor state with only small magnetic fluctuations at the center of plasma. The local particle diffusion coefficient was measured in these Proto S-1C discharges, the technique used was to inject a delta-function source of impurities into the plasma and observe the motion of the impurities relative to the flux surface. It was found that, during the decay phase of the spheromak discharge, when the plasma was in a Taylor state, the carbon diffusion coefficient was explained classically. While the plasma was relaxing towards the Taylor state, the diffusion coefficient was 2 ∼ 4a times larger than classical. At this time, the plasma was not yet force-free. This nonclassical diffusion appears to have been caused by v/sub ExB/ velocities due to correlations between the fluctuating electric field and density. Because the v/sub ExB/ velocity acts on all of the plasma species similarly, the anomalous hydrogen-particle diffusion coefficient should have been as large as that of carbon

  1. The development of ex-core neutron flux monitoring system for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S

    2004-12-01

    Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying

  2. The development of ex-core neutron flux monitoring system for integral reactor

    International Nuclear Information System (INIS)

    Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S.

    2004-12-01

    Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying the ex-core

  3. Reference equilibrium core with central flux irradiation facility for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Israr, M.; Shami, Qamar-ud-din; Pervez, S.

    1997-11-01

    In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

  4. Overview of the HIT-SI3 spheromak experiment

    Science.gov (United States)

    Hossack, A. C.; Jarboe, T. R.; Chandra, R. N.; Morgan, K. D.; Sutherland, D. A.; Everson, C. J.; Penna, J. M.; Nelson, B. A.

    2017-10-01

    The HIT-SI and HIT-SI3 spheromak experiments (a = 23 cm) study efficient, steady-state current drive for magnetic confinement plasmas using a novel method which is ideal for low aspect ratio, toroidal geometries. Sustained spheromaks show coherent, imposed plasma motion and low plasma-generated mode activity, indicating stability. Analysis of surface magnetic fields in HIT-SI indicates large n = 0 and 1 mode amplitudes and little energy in higher modes. Within measurement uncertainties all the n = 1 energy is imposed by the injectors, rather than being plasma-generated. The fluctuating field imposed by the injectors is sufficient to sustain the toroidal current through dynamo action whereas the plasma-generated field is not (Hossack et al., Phys. Plasmas, 2017). Ion Doppler spectroscopy shows coherent, imposed plasma motion inside r 10 cm in HIT-SI and a smaller volume of coherent motion in HIT-SI3. Coherent motion indicates the spheromak is stable and a lack of plasma-generated n = 1 energy indicates the maximum q is maintained below 1 for stability during sustainment. In HIT-SI3, the imposed mode structure is varied to test the plasma response (Hossack et al., Nucl. Fusion, 2017). Imposing n = 2, n = 3, or large, rotating n = 1 perturbations is correlated with transient plasma-generated activity. Work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Number DE-FG02-96ER54361.

  5. Critical heat flux experiments in tight lattice core

    Energy Technology Data Exchange (ETDEWEB)

    Kureta, Masatoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Fuel rods of the Reduced-Moderation Water Reactor (RMWR) are so designed to be in tight lattices as to reduce moderation and achieve higher conversion ratio. As for the BWR type reactor coolant flow rate is reduced small compared with the existing BWR, so average void fraction comes to be langer. In order to evaluate thermo hydraulic characteristics of designed cores, critical heat flux experiments in tight lattice core have been conducted using simulated high pressure coolant loops for both the PWR and BWR seven fuel rod bundles. Experimental data on critical heat flux for full bundles have been accumulated and applied to assess the critical power of designed cores using existing codes. Evaluated results are conservative enough to satisfy the limiting condition. Further experiments on axial power distribution effects and 37 fuel rod bundle tests will be performed to validate thermohydraulic characteristics of designed cores. (T. Tanaka)

  6. Critical heat flux experiments in tight lattice core

    International Nuclear Information System (INIS)

    Kureta, Masatoshi

    2002-01-01

    Fuel rods of the Reduced-Moderation Water Reactor (RMWR) are so designed to be in tight lattices as to reduce moderation and achieve higher conversion ratio. As for the BWR type reactor coolant flow rate is reduced small compared with the existing BWR, so average void fraction comes to be langer. In order to evaluate thermo hydraulic characteristics of designed cores, critical heat flux experiments in tight lattice core have been conducted using simulated high pressure coolant loops for both the PWR and BWR seven fuel rod bundles. Experimental data on critical heat flux for full bundles have been accumulated and applied to assess the critical power of designed cores using existing codes. Evaluated results are conservative enough to satisfy the limiting condition. Further experiments on axial power distribution effects and 37 fuel rod bundle tests will be performed to validate thermohydraulic characteristics of designed cores. (T. Tanaka)

  7. New mode of operating a magnetized coaxial plasma gun for injecting magnetic helicity into a spheromak

    International Nuclear Information System (INIS)

    Woodruff, S.; Hill, D.N.; Stallard, B.W.; Bulmer, R.; Cohen, B.; Holcomb, C.T.; Hooper, E.B.; McLean, H.S.; Moller, J.; Wood, R.D.

    2003-01-01

    By operating a magnetized coaxial plasma gun continuously with just sufficient current to enable plasma ejection, large gun-voltage spikes (∼1 kV) are produced, giving the highest sustained voltage ∼500 V and highest sustained helicity injection rate observed in the Sustained Spheromak Physics Experiment. The spheromak magnetic field increases monotonically with time, exhibiting the lowest fluctuation levels observed during formation of any spheromak (B-tilde)/B≥2%). The results suggest an important mechanism for field generation by helicity injection, namely, the merging of helicity-carrying filaments

  8. New mode of operating a magnetized coaxial plasma gun for injecting magnetic helicity into a spheromak.

    Science.gov (United States)

    Woodruff, S; Hill, D N; Stallard, B W; Bulmer, R; Cohen, B; Holcomb, C T; Hooper, E B; McLean, H S; Moller, J; Wood, R D

    2003-03-07

    By operating a magnetized coaxial plasma gun continuously with just sufficient current to enable plasma ejection, large gun-voltage spikes (approximately 1 kV) are produced, giving the highest sustained voltage approximately 500 V and highest sustained helicity injection rate observed in the Sustained Spheromak Physics Experiment. The spheromak magnetic field increases monotonically with time, exhibiting the lowest fluctuation levels observed during formation of any spheromak (B/B>/=2%). The results suggest an important mechanism for field generation by helicity injection, namely, the merging of helicity-carrying filaments.

  9. Relaxation phenomena in the high temperature S-1 spheromak

    International Nuclear Information System (INIS)

    Ono, Y.; Ellis, R.A. Jr.; Janos, A.C.; Levinton, F.M.; Mayo, R.M.; Motley, R.W.; Ueda, Y.; Yamada, M.

    1988-06-01

    Operation of the S-1 device in a high current density (j/n/sub e/ ≥ 2 /times/ 10 -14 A/center dot/m) regime has created high electron temperature spheromaks (50eV ≤ T/sub e/ ≤ 130eV). The mechanisms and causes of the periodic relaxation events often observed in these hotter spheromak plasmas were made clear. Also, a relationship between the MHD relaxation cycle and confinement characteristics was revealed for the first time. Resistive loss at the outer edge of the plasma causes a departure from the initial force-free minimum-energy Taylor state to a MHD profile unstable to low-n ideal MHD modes; a relaxation event then returns the configuration to nearly a Taylor state. 11 refs., 5 figs

  10. Theory of the evolution of the decaying spheromak

    International Nuclear Information System (INIS)

    Sgro, A.G.; Marklin, G.; Mirin, A.A.

    1986-01-01

    The strongly nonlinear dynamics present in decaying Spheromaks has been studied by various computational methods, demonstrating the tendency of the plasma to oscillate about or approach relaxed states and resulting in new insight into the significance of minimum energy states

  11. High aspect ratio spheromak experiments

    International Nuclear Information System (INIS)

    Robertson, S.; Schmid, P.

    1987-05-01

    The Reversatron RFP (R/a = 50cm/8cm) has been operated as an ohmically heated spheromak of high aspect ratio. We find that the dynamo can drive the toroidal field upward at rates as high as 10 6 G/sec. Discharges can be initiated and ramped upward from seed fields as low as 50 G. Small toroidal bias fields of either polarity (-0.2 < F < 0.2) do not significantly affect operation. 5 refs., 3 figs

  12. Online In-Core Thermal Neutron Flux Measurement for the Validation of Computational Methods

    International Nuclear Information System (INIS)

    Mohamad Hairie Rabir; Muhammad Rawi Mohamed Zin; Yahya Ismail

    2016-01-01

    In order to verify and validate the computational methods for neutron flux calculation in RTP calculations, a series of thermal neutron flux measurement has been performed. The Self Powered Neutron Detector (SPND) was used to measure thermal neutron flux to verify the calculated neutron flux distribution in the TRIGA reactor. Measurements results obtained online for different power level of the reactor. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and measured thermal neutron flux in the core are in very good agreement indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux distribution in the reactor core. Since the computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of RTP utilization. (author)

  13. Effect of the helicity injection rate and the Lundquist number on spheromak sustainment

    Science.gov (United States)

    García-Martínez, Pablo Luis; Lampugnani, Leandro Gabriel; Farengo, Ricardo

    2014-12-01

    The dynamics of the magnetic relaxation process during the sustainment of spheromak configurations at different helicity injection rates is studied. The three-dimensional activity is recovered using time-dependent resistive magnetohydrodynamic simulations. A cylindrical flux conserver with concentric electrodes is used to model configurations driven by a magnetized coaxial gun. Magnetic helicity is injected by tangential boundary flows. Different regimes of sustainment are identified and characterized in terms of the safety factor profile. The spatial and temporal behavior of fluctuations is described. The dynamo action is shown to be in close agreement with existing experimental data. These results are relevant to the design and operation of helicity injected devices, as well as to basic understanding of the plasma relaxation mechanism in quasi-steady state.

  14. Effect of the helicity injection rate and the Lundquist number on spheromak sustainment

    Energy Technology Data Exchange (ETDEWEB)

    García-Martínez, Pablo Luis, E-mail: pablogm@cab.cnea.gov.ar [Consejo Nacional de Investigaciones Científicas y Técnicas (CONICET) and Sede Andina—Universidad Nacional de Río Negro (UNRN), Av. Bustillo 9500, 8400 San Carlos de Bariloche, Río Negro (Argentina); Lampugnani, Leandro Gabriel; Farengo, Ricardo [Instituto Balseiro and Centro Atómico Bariloche (CAB-CNEA), Av. Bustillo 9500, 8400 San Carlos de Bariloche, Río Negro (Argentina)

    2014-12-15

    The dynamics of the magnetic relaxation process during the sustainment of spheromak configurations at different helicity injection rates is studied. The three-dimensional activity is recovered using time-dependent resistive magnetohydrodynamic simulations. A cylindrical flux conserver with concentric electrodes is used to model configurations driven by a magnetized coaxial gun. Magnetic helicity is injected by tangential boundary flows. Different regimes of sustainment are identified and characterized in terms of the safety factor profile. The spatial and temporal behavior of fluctuations is described. The dynamo action is shown to be in close agreement with existing experimental data. These results are relevant to the design and operation of helicity injected devices, as well as to basic understanding of the plasma relaxation mechanism in quasi-steady state.

  15. Structure of Maryland Spheromak plasmas

    International Nuclear Information System (INIS)

    Hess, R.; Chinfatt, C.; Cote, C.; DeSilva, A.; Filuk, A.; Goldenbaum, G.; Gauvreau, J.; Hwang, Fukwun

    1990-01-01

    Recent efforts on the Maryland Spheromak (MS) have concentrated on detailed measurement of magnetic field structures in order to better understand the formation and evolution of the spheromak configuration. These efforts were prompted by results showing a very rapid decay of the magnetic field under certain conditions. It was not known if this loss was a rapid movement of the plasma to the walls of the vacuum vessel, or by some mechanism causing a rapid decay of a more or less stationary field. To investigate the magnetic field structure in more detail, an array of magnetic probes was built that could be moved from shot to shot so as to acquire a complete map of the three magnetic field components in a plane containing the symmetry axis of the machine. Data taken with these probes in a case where the rapid loss of field occurs is given. Further analysis of the data shows that the instability that forms is a combination of tilt and shift. The initial asymmetry of the magnetic field is possibly due to the non-symmetric configuration of the reversal field coils, or the non-symmetric cabling to the I z electrodes. Future work will concentrate on eliminating the initial plasma asymmetry by eliminating any asymmetries in the machine, and on stopping the tilt/shift instability by different configurations for the passive stabilization coils

  16. 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)

  17. Fast Flux Test Facility core system

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.; Leggett, R.D.; Pitner, A.L.; Waltar, A.E.

    1990-11-01

    A review of Liquid Metal Reactor (LMR) core system accomplishments provides an excellent road map through the maze of issues that faced reactor designers 10 years ago. At that time relatively large uncertainties were associated with fuel pin and fuel assembly performance, irradiation of structural materials, and performance of absorber assemblies. The extensive core systems irradiation program at the US Department of Energy's Fast Flux Test Facility (FFTF) has addressed each of these principal issues. As a result of the progress made, the attention of long-range LMR planners and designers can shift away from improving core systems and focus on reducing capital costs to ensure the LMR can compete economically in the 21st century with other nuclear reactor concepts. 3 refs., 6 figs., 1 tab

  18. Effects of passive coils on spheromak gross MHD instabilities

    International Nuclear Information System (INIS)

    Munson, C.; Janos, A.; Paul, S.; Wysocki, F.; Yamada, M.

    1983-01-01

    The experimental investigation of the effectiveness of figure-8 coils in stabilizing the n=1 tilting mode of spheromak plasmas in Proto S-1 A/B is extended. In addition, another coil configuration, the saddle coil, is examined

  19. 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)

  20. Model predictions for auxiliary heating in spheromaks

    International Nuclear Information System (INIS)

    Fauler, T.K.; Khua, D.D.

    1997-01-01

    Calculations are presented of the plasma temperature waited for under auxiliary heating in spheromaks. A model, ensuring good agreement of earlier experiments with joule heating results, is used. The model includes heat losses due to magnetic fluctuations and shows that the plasma temperatures of the kilo-electron-volt order may be achieved in a small device with the radius of 0.3 m only

  1. Current drive by neutral beams, rotating magnetic fields and helicity injection in compact toroids

    International Nuclear Information System (INIS)

    Farengo, R.

    2002-01-01

    A Monte-Carlo code is used to study neutral beam current drive in Spheromaks. The exact particle trajectories are followed in the self-consistent equilibria calculated including the beam current. Reducing Z(eff) does not increase the current drive efficiency because the reduction of the stopping cross section is compensated by an increase in the electron canceling current. Significant changes in the safety factor profile can be produced with relatively low beam currents. Minimum dissipation states of a flux core spheromak sustained by helicity injection are presented. Helicity balance is used as a constraint and the resistivity is considered to be non-uniform. Two types of relaxed states are found; one has a central core of open flux surrounded by a toroidal region of closed flux surfaces and the other has the open flux wrapped around the closed flux surfaces. Non-uniform resistivity effects can be very important due to the changes they produce in the safety factor profile. A hybrid, fluid electrons particle ions, code is employed to study ion dynamics in FRCs sustained by rotating magnetic fields. (author)

  2. 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

  3. Particle diffusion in a spheromak

    International Nuclear Information System (INIS)

    Meyerhofer, D.D.; Levinton, F.M.; Yamada, M.

    1988-01-01

    The local carbon particle diffusion coefficient was measured in the Proto S-1/C spheromak using a test particle injection scheme. When the plasma was not in a force-free Taylor state, and when there were pressure gradients in the plasma, the particle diffusion was five times that predicted by Bohm and was consistent with collisional drift wave diffusion. The diffusion appears to be driven by correlations of the fluctuating electric field and density. During the decay phase of the discharge when the plasma was in the Taylor state, the diffusion coefficient of the carbon was classical. 23 refs., 4 figs

  4. E-core transverse flux machine with integrated fault detection system

    DEFF Research Database (Denmark)

    Rasmussen, Peter Omand; Runólfsson, Gunnar; Thorsdóttir, Thórunn Ágústa

    2011-01-01

    extent also thermal. Since the E-core transverse flux-machine belongs to the family of the SRMs it has unique properties of intervals without current in the windings. By careful investigation of the voltage and current in these intervals a very simple method to detect single and partial turn short...... circuit faults have been developed. For other types of machines the single and partial turn short circuit is very difficult to deal with and requires normally very comprehensive detection and calculation schemes. The developed detection algorithm combined with the E-core transverse flux machine...

  5. Excitation of neutron flux waves in reactor core transients

    International Nuclear Information System (INIS)

    Carew, J.F.; Neogy, P.

    1983-01-01

    An analysis of the excitation of neutron flux waves in reactor core transients has been performed. A perturbation theory solution has been developed for the time-dependent thermal diffusion equation in which the absorption cross section undergoes a rapid change, as in a PWR rod ejection accident (REA). In this analysis the unperturbed reactor flux states provide the basis for the spatial representation of the flux solution. Using a simplified space-time representation for the cross section change, the temporal integrations have been carried out and analytic expressions for the modal flux amplitudes determined. The first order modal excitation strength is determined by the spatial overlap between the initial and final flux states, and the cross section perturbation. The flux wave amplitudes are found to be largest for rapid transients involving large reactivity perturbations

  6. Critical beta for analytical spheromak equilibria

    International Nuclear Information System (INIS)

    Freire, E.M.; Clemente, R.A.

    1985-01-01

    The Mercier criterion is applied to two analytical spheromak equilibria, one with a spherical separatrix and the other with a cylindrical one of variable elongation. The maximum beta, defined as the ratio between the plasma pressure and the magnetic pressure averaged over the plasma volume, for which the criterion is satisfied on every magnetic surface, has been obtained. In the spherical model the critical beta is 0.003, while in the cylindrical case it is a function of the elongation of the separatrix with a maximum of 0.083. (author)

  7. Fast Flux Test Facility core restraint system performance

    International Nuclear Information System (INIS)

    Hecht, S.L.; Trenchard, R.G.

    1990-02-01

    Characterizing Fast Flux Test Facility (FFTF) core restraint system performance has been ongoing since the first operating cycle. Characterization consists of prerun analysis for each core load, in-reactor and postirradiation measurements of subassembly withdrawal loads and deformations, and using measurement data to fine tune predictive models. Monitoring FFTF operations and performing trend analysis has made it possible to gain insight into core restraint system performance and head off refueling difficulties while maximizing component lifetimes. Additionally, valuable information for improved designs and operating methods has been obtained. Focus is on past operating experience, emphasizing performance improvements and avoidance of potential problems. 4 refs., 12 figs., 2 tabs

  8. Numerical Analysis on Heat Flux Distribution through the Steel Liner of the Ex-vessel Core Catcher

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Hong; Choi, Choeng Ryul [ELSOLTEC, Yongin (Korea, Republic of); Kim, Byung Jo; Lee, Kyu Bok [KEPCO, Gimcheon (Korea, Republic of); Hwang, Do Hyun [KHNP-CRI, Daejeon (Korea, Republic of)

    2016-05-15

    In order to prevent material failure of steel container of the core catcher system due to high temperatures, heat flux through the steel liner wall must be kept below the critical heat flux (CHF), and vapor dry-out of the cooling channel must be avoided. In this study, CFD methodology has been developed to simulate the heat flux distribution in the core catcher system, involving following physical phenomena: natural convection in the corium pool, boiling heat transfer and solidification/melting of the corium. A CFD methodology has been developed to simulate the thermal/hydraulic phenomena in the core catcher system, and a numerical analysis has been carried out to estimate the heat flux through the steel liner of the core catcher. High heat flux values are formed at the free surface of the corium pool. However, the heat flux through the steel liner is maintained below the critical heat flux.

  9. Field-Reversed Configuration Formation Scheme Utilizing a Spheromak and Solenoid Induction

    International Nuclear Information System (INIS)

    Gerhardt, S.P.; Belova, E.V.; Yamada, M.; Ji, H.; Ren, Y.; McGeehan, B.; Inomoto, M.

    2008-01-01

    A new field-reversed configuration (FRC) formation technique is described, where a spheromak transitions to a FRC with inductive current drive. The transition is accomplished only in argon and krypton plasmas, where low-n kink modes are suppressed; spheromaks with a lighter majority species, such as neon and helium, either display a terminal tilt-mode, or an n=2 kink instability, both resulting in discharge termination. The stability of argon and krypton plasmas through the transition is attributed to the rapid magnetic diffusion of the currents that drive the kink-instability. The decay of helicity during the transition is consistent with that expected from resistivity. This observation indicates a new scheme to form a FRC plasma, provided stability to low-n modes is maintained, as well as a unique situation where the FRC is a preferred state

  10. Investigating the Effects of I-Shaped Cores in an Outer-Rotor Transverse Flux Permanent Magnet Generator

    DEFF Research Database (Denmark)

    Hosseini, Seyedmohsen; Moghani, Javad Shokrollahi; Jensen, Bogi Bech

    2011-01-01

    This paper deals with the effects of I-shaped cores in an outer-rotor transverse flux permanent magnet generator. Performance characteristics of a typical outer-rotor transverse flux permanent magnet generator are obtained in two cases; with and without I-shaped cores. The results show that altho...... the advantages and disadvantage of using I-shaped cores and emphasizes the necessity of performing a tradeoff study between using and not using I-shaped cores in practical transverse flux permanent magnet generators....

  11. Heat loss by helicity injection in spheromaks

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1994-01-01

    A model is presented for spheromak buildup and decay including thermal diffusivity associated with magnetic turbulence during helicity injection. It is shown that heat loss by magnetic turbulence scales more favorably than gyroBohm transport. Thus gyroBohm scaling for the proposed ignition experiment would be the conservative choice, though present experiments may be dominated by magnetic turbulence. Because of a change in boundary conditions when the gun is turned off, the model may account for the observed increase in electron temperature in CTX after turnoff

  12. Performance analysis of a new radial-axial flux machine with SMC cores and ferrite magnets

    Science.gov (United States)

    Liu, Chengcheng; Wang, Youhua; Lei, Gang; Guo, Youguang; Zhu, Jianguo

    2017-05-01

    Soft magnetic composite (SMC) is a popular material in designing of new 3D flux electrical machines nowadays for it has the merits of isotropic magnetic characteristic, low eddy current loss and high design flexibility over the electric steel. The axial flux machine (AFM) with the extended stator tooth tip both in the radial and circumferential direction is a good example, which has been investigated in the last years. Based on the 3D flux AFM and radial flux machine, this paper proposes a new radial-axial flux machine (RAFM) with SMC cores and ferrite magnets, which has very high torque density though the low cost low magnetic energy ferrite magnet is utilized. Moreover, the cost of RAFM is quite low since the manufacturing cost can be reduced by using the SMC cores and the material cost will be decreased due to the adoption of the ferrite magnets. The 3D finite element method (FEM) is used to calculate the magnetic flux density distribution and electromagnetic parameters. For the core loss calculation, the rotational core loss computation method is used based on the experiment results from previous 3D magnetic tester.

  13. Neutron energy spectrum flux profile of Ghana's miniature neutron source reactor core

    International Nuclear Information System (INIS)

    Sogbadji, R.B.M.; Abrefah, R.G.; Ampomah-Amoako, E.; Agbemava, S.E.; Nyarko, B.J.B.

    2011-01-01

    Highlights: → The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was studied. → Using 20,484 energy grids, the thermal, slowing down and fast neutron energy regions were studied. - Abstract: The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) x 10 12 n/cm 2 s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) x 10 11 n/cm 2 s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) x 10 11 n/cm 2 s. The peak values of the thermal energy range occurred in the energy range (1.8939-3.7880) x 10 -08 MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) x 10 09 n/cm 2 s at the lower energy end of the slowing down region between 8.2491 x 10 -01 MeV and 8.2680 x 10 -01 MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) x 10 08 n/cm 2 s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.

  14. Experiments of spheromak and reversed field configuration in 2m theta pinch

    International Nuclear Information System (INIS)

    Nogi, Y.; Shimamura, S.; Ogura, H.; Osanai, Y.; Saito, K.; Shiina, S.; Yoshimura, H.

    1981-01-01

    Since the z-current produces the paramagnetic field near the electrodes, the spheromak formation is more difficult in the straight bias field. In order to help the reconnection at the coil ends, the cusp bias coils are added to both ends of the straight coil. Then the spheromak configuration is formed and the plasma is confined for 5 to 10 μs. On the other hand, the RFC continues for about 30 μs in case of the straight bias field. The confinement time is limited by the rotational instability. Although the start time of the instability is not clear, the elongation of the plasma is detected in 15 to 20 μs after the RFC is formed. The period of the rotation is slightly different every shot. Detailed study of the instability is being pursued

  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. Detection of flux perturbations in pebble bed HTGRs by near core instrumentation

    International Nuclear Information System (INIS)

    Neef, R.D.; Basse, W.; Carlson, D.E.; Knob, P.; Schaal, H.; Wilhelm, H.; Stroemich, A.

    1982-06-01

    For pebble bed reactors an incore monitoring system cannot be utilized during normal operation, mainly for two reasons: 1) The necessary instrumentation cannot withstand possible coolant gas temperatures of up to 1150 deg. C. 2) The detector guide structures cannot withstand the continuous downward movement of the fuel elements in the core and would perturb the loading scheme. Therefore a near-core detector system is necessary which can be used to monitor the power distribution and to recognise perturbations in the neutron flux distribution. This helps guarantee that temperature limits in the core (fuel elements, absorber rods) and in the heat removal systems (steam generators) will not be exceeded. For this purpose an instrumentation system of the following kind is planned (and at least for a prototype reactor no part of it should be omitted): 1) Fast fission chambers in the top reflector for measuring the fast neutron flux distribution; 2) Self powered neutron detectors (SPNDs) in the radial reflector for thermal flux mapping; 3) Thermocouples in the bottom reflector for measuring the profile of the outlet gas temperature

  17. 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%

  18. A spheromak ignition experiment reusing Mirror Fusion Test Facility (MFTF) equipment

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1993-01-01

    Based on available experimental results and theory, a scenario is presented to achieve ohmic ignition in a spheromak by slow (∼ 10 sec.) helicity injection using power from the Mirror Fusion Test Facility (MFTF) substation. Some of the other parts needed (vacuum vessel, coils, power supplies, pumps, shielded building space) might also be obtained from MFTF or other salvage, as well as some components needed for intermediate experiments for additional verification of the concept (especially confinement scaling). The proposed ignition experiment would serve as proof-of-principle for the spheromak DT fusion reactor design published by Hagenson and Krakowski, with a nuclear island cost about ten times less than a tokamak of comparable power. Designs at even higher power density and lower cost might be possible using Christofilos' concept of a liquid lithium blanket. Since all structures would be protected from neutrons by the lithium blanket and the tritium inventory can be reduced by continuous removal from the liquid blanket, environmental and safety characteristics appear to be favorable

  19. Axisymmetric force-free states and relaxation of a spheroidal spheromak

    International Nuclear Information System (INIS)

    Throumoulopoulos, G.N.; Pantis, G.

    1990-01-01

    Axisymmetric force-free equilibrium eigenstates for a prolate as well as an oblate spheroidal Spheromak with arbitrary elongation are obtained. In the framework of the Woltjer-Taylor relaxation theory the relaxed states are also identified. A simple hypothesis for the relaxation process is introduced, which implies that the plasma relaxes from multitoroidal formations to a singly toroidal configuration, in qualitative agreement with experimental results. (author)

  20. Axisymmetric force-free states and relaxation of a spheroidal spheromak

    International Nuclear Information System (INIS)

    Throumoulopoulos, G.N.; Pantis, G.

    1990-01-01

    Axisymmetric force-free equilibrium eigenstates for a prolate as well as an oblate spheroidal spheromak with arbitrary elongation are obtained. In the framework of the Woltjer-Taylor relaxation theory the relaxed states are also identified. A simple hypothesis for the relaxation process is introduced which implies that the plasma relaxes from multitoroidal formations to a singly toroidal configuration in qualitative agreement with experimental results. (Author)

  1. 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)

  2. Nonlinear saturation of non-resonant internal instabilities in a straight spheromak

    International Nuclear Information System (INIS)

    Park, W.; Jardin, S.C.

    1982-04-01

    An initial value numerical solution of the time dependent nonlinear ideal magnetohydrodynamic equations demonstrates that spheromak equilibria which are linearly unstable to nonresonant helical internal perturbations saturate at low amplitude without developing singularities. These instabilities thus represent the transition from an axisymmetric to a non-axisymmetric equilibrium state, caused by a peaking of the current density

  3. Process and equipment for monitoring flux distribution in a nuclear reactor outside the core

    International Nuclear Information System (INIS)

    Graham, K.F.; Gopal, R.

    1977-01-01

    This concerns the monitoring system for axial flux distribution during the whole load operating range lying outside the core of, for example, a PWR. Flux distribution cards can be produced continuously. The core is divided into at least three sections, which are formed by dividing it at right angles to the longitudinal axis, and the flux is measured outside the core using adjacent detectors. Their output signals are calibrated by amplifiers so that the load distribution in the associated sections is reproduced. A summation of the calibrated output signals and the formation of a mean load signal takes place in summing stages. For monitoring, this is compared with a value which corresponds to the maximum permissible load setting. Apart from this the position of the control rods in the core can be taken into account by multiplication of the mean load signals by suitable peak factors. The distribution of monitoring positions or the position of the detectors can be progressive or symmetrical along the axis. (DG) 891 HP [de

  4. Design and analysis of EI core structured transverse flux linear reluctance actuator

    OpenAIRE

    FENERCİOĞLU, AHMET; AVŞAR, YUSUF

    2015-01-01

    In this study, an EI core linear actuator is proposed for horizontal movement systems. It is a transverse flux linear switched reluctance motor designed with an EI core structure geometrically. The actuator is configured into three phases and at a 6/4 pole ratio, and it has a stationary active stator along with a sliding passive translator. The stator consists of E cores and the translator consists of I cores. The actuator has a yokeless design because the stator and translator have no back i...

  5. Neutron flux measurement in the central channel (XC-1) of TRIGA 14 MW LEU core

    International Nuclear Information System (INIS)

    BARBOS, D.; BUSUIOC, P.; ROTH, Cs.; PAUNOIU, C.

    2008-01-01

    The TRIGA 14 MW reactor, operated by Institute for Nuclear Research Pitesti, Romania, is a pool type reactor, and has a rectangular shape which holds fuel bundles and is surrounded with beryllium reflectors. Each fuel bundle is composed of 25 nuclear fuel rods. The TRIGA 14 MW reactor was commissioned 28 years ago with HEU fuel rods. The conversion was gradually achieved, starting in February 1992 and completed in March 2006. The full conversion of the 14 MW TRIGA Research Reactor was completed in May 2006 and each step of the conversion was achieved by removal of HEU fuel, replaced by LEU fuel, accompanied by a large set of theoretical evaluation and physical measurements intended to confirm the performances of gradual conversion. After the core full conversion, a program of measurements and comparisons with previous results of core physics and measurements is underway, allowing data acquisition for normal operation, demonstration of safety and economics of the converted core. Neutron flux spectrum measurements in the XC in the XC-1 water 1 water-filled channel were performed using multi multi-foil activation techniques. The neutron spectra and flux are obtained by unfolding from measured reaction rates using SAND II computer code. The integral neutron flux value for LEU core is greater of 13% than for the standard HEU core. Also thermal neutron flux value for converted LEU core is smaller by 0.38% than for the standard HEU core. These differences appear because the foil activation detectors have been irradiated using a pneumatic rabbit having a diameter of 32 mm, whereas foil irradiations in standard HEU core has been performed with a pneumatic rabbit having a diameter of 14 mm, and therefore the neutron spectra in LEU core is less thermalized and the weight of fast neutron is greater

  6. Correlation and flux tilt measurements of coupled-core reactor assemblies

    International Nuclear Information System (INIS)

    Harries, J.R.

    1976-01-01

    The systematics of coupling reactivity and time delay between cores have been investigated with a series of coupled-core assemblies on the AAEC Split-table Critical Facility. The assemblies were similar to the Universities' Training Reactor (UTR), but had graphite coupling region thickness of 450 mm, 600 mm and 800 mm. The coupling reactivity measured by both the cross-correlation of reactor noise and the flux tilt methods was stronger than for the UTRs, but showed a similar trend with core spacing. The cross-correlograms were analysed using the two-node model to derive the time delays between the cores. The time delays were compared with thermal neutron wave propagation, and found to be consistent when the time delays were added to the individual node response-function delays. (author)

  7. Current-driven instabilities of the kinetic shear Alfven wave: Application to reversed field pinches and spheromaks

    International Nuclear Information System (INIS)

    Meyerhofer, D.D.; Perkins, F.W.

    1984-01-01

    The kinetic Alfven wave is studied in a cylindrical force-free plasma with self-consistent magnetic fields. This equilibrium represents a reversed field pinch or a spheromak. The stability of the wave is found to depend on the ratio of the electron drift velocity to the Alfven velocity. This ratio varies inversely with the square root of the plasma line density. The critical line density using the Spitzer--Harm electron distribution function is found for reversed field pinches with deuterium plasmas to be approximately 2 x 10 18 and is 5 x 10 17 m -1 in spheromaks with hydrogen plasmas. The critical line density is in reasonable agreement with experimental data for reversed field pinches

  8. Current driven instabilities of the kinetic shear Alfven wave: application to reversed field pinches and spheromaks

    International Nuclear Information System (INIS)

    Meyerhofer, D.D.; Perkins, F.W.

    1984-04-01

    The kinetic Alfven wave is studied in a cylindrical force-free plasma with self-consistent magnetic fields. This equilibrium represents a reversed field pinch or a spheromak. The stability of the wave is found to depend on the ratio of the electron drift velocity to the Alfven velocity. This ratio varies inversely with the square root of the plasma line density. The critical line density using the Spitzer-Harm electron distribution function is found for reversed field pinches with deuterium plasmas to be approximately 2 x 10 18 m -1 and is 5 x 10 17 m -1 in spheromaks with hydrogen plasmas. The critical line density is in reasonable agreement with experimental data for reversed field pinches

  9. Measurements of neutron flux distributions in the core of the Ljubljana TRIGA Mark II Reactor

    International Nuclear Information System (INIS)

    Rant, J.; Ravnik, M.; Mele, I.; Dimic, V.

    2008-01-01

    Recently the Ljubljana TRIGA Mark II Reactor has been refurbished and upgraded to pulsed operation. To verify the core design calculations using TRIGAP and PULSTR1 codes and to obtain necessary data for future irradiation and neutron beam experiments, an extensive experimental program of neutron flux mapping and neutron field characterization was carried out. Using the existing neutron measuring thimbles complete axial and radial distributions in two radial directions were determined for two different core configurations. For one core configuration the measurements were also carried out in the pulsed mode. For flux distributions thin Cu (relative measurements) and diluted Au wires (absolute values) were used. For each radial position the cadmium ratio was determined in two axial levels. The core configuration was rather uniform, well defined (fresh fuel of a single type, including fuelled followers) and compact (no irradiation channels or gaps), offering unique opportunity to test the computer codes for TRIGA reactor calculations. The neutron flux measuring procedures and techniques are described and the experimental results are presented. The agreement between the predicted and measured power peaking factors are within the error limits of the measurements (<±5%) and calculations (±10%). Power peaking occurs in the B ring, and in the A ring (centre) there is a significant flux depression. (authors)

  10. 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.

  11. Pipeline welding with Flux Cored and Metal Cored Wire; Soldagem de dutos com processos Arame Tubular e de Alma Metalica

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Ubirajara Pereira da [ITW Soldagem Brasil Miller-Hobart, Sao Paulo, SP (Brazil)

    2003-07-01

    Different welding process like SMAW, Semi-Automatic FCAW Gas-shielded and Self-shielded and Mechanized GMAW-MAG with Solid Wire are suggested to weld Transmission Pipelines. Presently, the largest extensions of Transmission Pipelines under construction, are in China like Lines West-East, Zong-Wu, Shan-Jing Fuxian and some others, totalizing about 8.000 km, and all using Semi-Automatic Self Shielded Flux Cored Arc Welding Process. Also, several papers and magazines that covers Transmission Pipelines Welding, not frequently mention Operational aspects of the process and some other variables like environment and site geography. This presentation intends to cover some of the Operational aspects of the Flux Cored Arc Welding and GMAW-Metal Cored in order to give sufficient information for Construction, Engineering, Projects e Contractors so they can evaluate these Process against the SMAW or even Mechanized Systems, considering the Operation Factor, Efficiency and Deposition Rate. We will not cover operational details of the GMAW Mechanized Systems but only suggest that be evaluated the possibility to replace the GMAW-Solid Wire by the GMAW-Metal Cored Wire. (author)

  12. Properties of spheromaks generated by a magnetized coaxial source

    International Nuclear Information System (INIS)

    Hoida, H.W.; Henins, I.; Jarboe, T.R.; Linford, R.K.; Lipson, J.; Marshall, J.; Platts, D.A.; Sherwood, A.R.; Tuszewski, M.

    1981-01-01

    In gun-generated spheromaks impurity contamination plays an important role in determining the energy loss. Metallic impurities can be reduced by an appropriate change of source parameters. The reduction of the level of metal impurities results in a spectrum showing a preponderance of oxygen and carbon lines and OIV radiation is observed to increase indicating a warmer plasma. However, the plasma lifetime is not changed. Discharge cleaning techniques appear to be necessary. It is still possible that electron heat conduction during the reconnection processs will be found to be important once the impurities are reduced

  13. Properties of spheromaks generated by a magnetized coaxial source

    Energy Technology Data Exchange (ETDEWEB)

    Hoida, H.W.; Henins, I.; Jarboe, T.R.; Linford, R.K.; Lipson, J.; Marshall, J.; Platts, D.A.; Sherwood, A.R.; Tuszewski, M.

    1981-01-01

    In gun-generated spheromaks impurity contamination plays an important role in determining the energy loss. Metallic impurities can be reduced by an appropriate change of source parameters. The reduction of the level of metal impurities results in a spectrum showing a preponderance of oxygen and carbon lines and OIV radiation is observed to increase indicating a warmer plasma. However, the plasma lifetime is not changed. Discharge cleaning techniques appear to be necessary. It is still possible that electron heat conduction during the reconnection processs will be found to be important once the impurities are reduced.

  14. TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR

    International Nuclear Information System (INIS)

    Kurosawa, M.

    2005-01-01

    For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54 Mn and 60 Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. (authors)

  15. TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

    Science.gov (United States)

    Kurosawa, Masahiko

    2005-01-01

    For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

  16. Measurements of neutron fluxes and cadmium ratio at equilibrium core in JRR-3M

    International Nuclear Information System (INIS)

    Ohtomo, Akitoshi; Sasajima, Fumio; Ishida, Takuya; Shigemoto, Masamitsu; Takahashi, Hidetake; Maejima, Takeshi; Sekine, Katsunori.

    1993-08-01

    Construction and characteristics tests of JRR-3M (Modified JRR-3) had been completed on October 1990, and the reactor reached to equilibrium core in July 1991. Measurements of neutron flux and cadmium ratio in Hydraulic irradiation facility (HR) and Pneumatic irradiation facility (PN) at 20 MW reactor power were carried out for the equilibrium core from May to August 1991 and for the latest core in April 1993. The results at the equilibrium core and the latest core are described in this paper. (author)

  17. Test of In-core Flux Detectors in KNK II

    CERN Document Server

    Hoppe, P

    1979-01-01

    The development of in-core detectors for Liquid Metal Fast Breeder Reactors (LMFBRs) is still in an early stage, and little operation experience is available. Therefore self-powered neutron and gamma detectors and neutron sensitive ionization chambers -especially developed for LMFBRs- have been tested in the Fast Sodium Cooled Test Reactor KNK II. Seven flux detectors have been installed in the core of KNK II by means of a special test rig. Five of them failed already within the first week during operation in the reactor. Due to measurements of electrical resistances and capacities, sodium penetrating into the detectors or cables probably seems to be the cause. As tests prior to the installation in the core proved the tightness of all detectors, it is suspected that small cracks have developed in the detector casings or in the outer cable sheaths during their exposure to the hot coolant. Two ionization chambers did not show these faults. However, one of them failed because the saturation current plateau disap...

  18. Thermal neutron flux distribution in the ET R R-1 reactor core as experimentally measured and theoretically calculated by the code triton

    Energy Technology Data Exchange (ETDEWEB)

    Imam, M [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

    1995-10-01

    Thermal neutron flux distributions that were measured earlier at the ET-R R-1 reactor are compared with those calculated by the three dimensional diffusion code Triton. This comparison was made for the horizontal and vertical flux distributions. The horizontal thermal flux distributions considered in this comparison were along the core diagonals at two planes of different heights from core bottom, where one at a level passing through the control rod at core center and the other at a level below this control rod. In the meantime all the control rods were taken into consideration. The effect of the existence of a water cavity inside the core as well as the influence of the control rods on the thermal flux are illustrated in this work. The vertical thermal flux distributions considered in the comparison were at two positions in core namely; one along the core height the horizontal reactor power distribution along the core height and the horizontal reactor power distribution along the core diagonal as calculated by the code Triton are also given this work. 8 figs., 1 tab.

  19. Improved methodology for generation of axial flux shapes in digital core protection systems

    International Nuclear Information System (INIS)

    Lee, G.-C.; Baek, W.-P.; Chang, S.H.

    2002-01-01

    An improved method of axial flux shape (AFS) generation for digital core protection systems of pressurized water reactors is presented in this paper using an artificial neural network (ANN) technique - a feedforward network trained by backpropagation. It generates 20-node axial power shapes based on the information from three ex-core detectors. In developing the method, a total of 7173 axial flux shapes are generated from ROCS code simulation for training and testing of the ANN. The ANN trained 200 data predicts the remaining data with the average root mean square error of about 3%. The developed method is also tested with the real plant data measured during normal operation of Yonggwang Unit 4. The RMS errors in the range of 0.9∼2.1% are about twice as accurate as the cubic spline approximation method currently used in the plant. The developed method would contribute to solve the drawback of the current method as it shows reasonable accuracy over wide range of core conditions

  20. The University of Maryland spheromak fusion experiment: Final report

    International Nuclear Information System (INIS)

    Antoniades, J.A.; Chin-Fatt, C.; DeSilva, A.W.; Goldenbaum, G.C.; Hess, R.A.; Shaw, R.S.

    1986-01-01

    The spheromak is a magnetic plasma confinement configuration that features a simple magnetic structure free of coils that link the plasma torus. It offers the possibility of a simple and efficient confinement system for a fusion plasma. Design of the experimental apparatus occupied the first 15 months of the contract period. At the same time, computer studies of the formation of the spheromak plasma, using a two-dimensional MHD code were performed. After the first 12 months of the contract period, subcontracts were let for major components of the system, particularly for the liquid nitrogen cooled bias magnetic coils, the associated power supplies, and the capacitors for the reversal bank. When the design work was complete, the machining contract for the vacuum vessel was placed. At about this time, work on the operating system for the control computer was begun. The necessary hardware items for the data acquisition computer were decided upon and ordered at the end of the second year. The capacitor bank for the Z-directed current (I/sub z/ bank) was rebuilt from existing parts here, and construction of this bank and of the parts for the reversal bank was accomplished while the outside fabrication of other major parts was in progress. Switching hardware for the two capacitor banks was fabricated in house to reduce costs. As capacitors for the reversal bank were delivered, they were incorporated into the bank modules. A full description of the MS experimental hardware is described in this paper. 2 refs., 9 figs., 1 tab

  1. An evaluation of multigroup flux predictions in the EBR-II core

    International Nuclear Information System (INIS)

    Hill, R.N.; Fanning, T.H.; Finck, P.J.

    1991-01-01

    The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required

  2. An evaluation of multigroup flux predictions in the EBR-II core

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.N.; Fanning, T.H.; Finck, P.J.

    1991-12-31

    The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required.

  3. An evaluation of multigroup flux predictions in the EBR-II core

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.N.; Fanning, T.H.; Finck, P.J.

    1991-01-01

    The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required.

  4. Split core experiments; Part I. Axial neutron flux distribution measurements in the reactor core with a central horizontal reflector

    Energy Technology Data Exchange (ETDEWEB)

    Strugar, P; Raisic, N; Obradovic, D; Jovanovic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1965-05-01

    A series of critical experiments were performed on the RB reactor in order to determine the thermal neutron flux increase in the central horizontal reflector formed by a split reactor core. The objectives of these experiments were to study the possibilities of improving the thermal neutron flux characteristics of the neutron beam in the horizontal beam tube of the RA research reactor. The construction of RA reactor enables to split the core in two, to form a central horizontal reflector in front of the beam tube. This is achieved by replacing 2% enriched uranium slugs in the fuel channel by dummy aluminium slugs. The purpose of the first series of experiments was to study the gain in thermal neutron component inside the horizontal reflector and the loss of reactivity as a function of the lattice pitch and central reflector thickness.

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

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1986-04-01

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

  6. Influence of fuel assembly loading pattern and fuel burnups upon leakage neutron flux spectra from light water reactor core (Joint research)

    International Nuclear Information System (INIS)

    Kojima, Kensuke; Okumura, Keisuke; Kosako, Kazuaki; Torii, Kazutaka

    2016-01-01

    At the decommissioning of light water reactors (LWRs), it is important to evaluate an amount of radioactivity in the ex-core structures such as a reactor containment vessel, radiation shieldings, and so on. It is thought that the leakage neutron spectra in these radioactivation regions, which strongly affect the induced radioactivity, would be changed by different reactor core configurations such as fuel assembly loading pattern and fuel burnups. This study was intended to evaluate these effects. For this purpose, firstly, partial neutron currents on the core surfaces were calculated for some core configurations. Then, the leakage neutron flux spectra in major radioactivation regions were calculated based on the provided currents. Finally, influence of the core configurations upon the neutron flux spectra was evaluated. As a result, it has been found that the influence is small on the spectrum shapes of neutron fluxes. However, it is necessary to pay attention to the facts that intensities of the leakage neutron fluxes are changed by the configurations and that intensities and spectrum shapes of the leakage neutron fluxes are changed depending on the angular direction around the core. (author)

  7. A mechanism for the dynamo terms to sustain closed-flux current, including helicity balance, by driving current which crosses the magnetic field

    Energy Technology Data Exchange (ETDEWEB)

    Jarboe, T. R.; Nelson, B. A.; Sutherland, D. A. [University of Washington, Seattle, Washington 98195 (United States)

    2015-07-15

    An analysis of imposed dynamo current drive (IDCD) [T.R. Jarboe et al., Nucl. Fusion 52 083017 (2012)] reveals: (a) current drive on closed flux surfaces seems possible without relaxation, reconnection, or other flux-surface-breaking large events; (b) the scale size of the key physics may be smaller than is often computationally resolved; (c) helicity can be sustained across closed flux; and (d) IDCD current drive is parallel to the current which crosses the magnetic field to produce the current driving force. In addition to agreeing with spheromak data, IDCD agrees with selected tokamak data.

  8. Variations of the core luminosity and solar neutrino fluxes

    Science.gov (United States)

    Grandpierre, Attila

    The aim of the present work is to analyze the geological and astrophysical data as well as presenting theoretical considerations indicating the presence of dynamic processes present in the solar core. The dynamic solar model (DSM) is suggested to take into account the presence of cyclic variations in the temperature of the solar core. Comparing the results of calculations of the CO2 content, albedo and solar evolutionary luminosity changes with the empirically determined global earthly temperatures, and taking into account climatic models, I determined the relation between the earthly temperature and solar luminosity. These results indicate to the observed maximum of 10o change on the global terrestrial surface temperature a related solar luminosity change around 4-5 % on a ten million years timescale, which is the timescale of heat diffusion from the solar core to the surface. The related solar core temperature changes are around 1 % only. At the same time, the cyclic luminosity changes of the solar core are shielded effectively by the outer zones since the radiation diffusion takes more than 105 years to reach the solar surface. The measurements of the solar neutrino fluxes with Kamiokande 1987-1995 showed variations higher than 40 % around the average, at the Super-Kamiokande the size of the apparent scatter decreased to 13 %. This latter scatter, if would be related completely to stochastic variations of the central temperature, would indicate a smaller than 1 % change. Fourier and wavelet analysis of the solar neutrino fluxes indicate only a marginally significant period around 200 days (Haubold, 1998). Helioseismic measurements are known to be very constraining. Actually, Castellani et al. (1999) remarked that the different solar models lead to slightly different sound speeds, and the different methods of regularization yield slightly different sound speeds, too. Therefore, they doubled the found parameter variations, and were really conservative assuming

  9. Two-fluid (plasma-neutral) Extended-MHD simulations of spheromak configurations in the HIT-SI experiment with PSI-Tet

    Science.gov (United States)

    Sutherland, D. A.; Hansen, C. J.; Jarboe, T. R.

    2017-10-01

    A self-consistent, two-fluid (plasma-neutral) dynamic neutral model has been implemented into the 3-D, Extended-MHD code PSI-Tet. A monatomic, hydrogenic neutral fluid reacts with a plasma fluid through elastic scattering collisions and three inelastic collision reactions: electron-impact ionization, radiative recombination, and resonant charge-exchange. Density, momentum, and energy are evolved for both the plasma and neutral species. The implemented plasma-neutral model in PSI-Tet is being used to simulate decaying spheromak configurations in the HIT-SI experimental geometry, which is being compare to two-photon absorption laser induced fluorescence measurements (TALIF) made on the HIT-SI3 experiment. TALIF is used to measure the absolute density and temperature of monatomic deuterium atoms. Neutral densities on the order of 1015 m-3 and neutral temperatures between 0.6-1.7 eV were measured towards the end of decay of spheromak configurations with initial toroidal currents between 10-12 kA. Validation results between TALIF measurements and PSI-Tet simulations with the implemented dynamic neutral model will be presented. Additionally, preliminary dynamic neutral simulations of the HIT-SI/HIT-SI3 spheromak plasmas sustained with inductive helicity injection will be presented. Lastly, potential benefits of an expansion of the two-fluid model into a multi-fluid model that includes multiple neutral species and tracking of charge states will be discussed.

  10. Effects of moderation level on core reactivity and. neutron fluxes in natural uranium fueled and heavy water moderated reactors

    International Nuclear Information System (INIS)

    Khan, M.J.; Aslam; Ahmad, N.; Ahmed, R.; Ahmad, S.I.

    2005-01-01

    The neutron moderation level in a nuclear reactor has a strong influence on core multiplication, reactivity control, fuel burnup, neutron fluxes etc. In the study presented in this article, the effects of neutron moderation level on core reactivity and neutron fluxes in a typical heavy water moderated nuclear research reactor is explored and the results are discussed. (author)

  11. Critical heat flux predictions for the Sandia Annular Core Research Reactor

    International Nuclear Information System (INIS)

    Rao, D.V.; El-Genk, M.S.

    1994-08-01

    This study provides best estimate predictions of the Critical Heat Flux (CHF) and the Critical Heat Flux Ratio (CHFR) to support the proposed upgrade of the Annual Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) from its present value of 2 MWt to 4 MWt. These predictions are based on the University of New Mexico (UNM) - CHF correlation, originally developed for uniformly heated vertical annuli. The UNM-CHF correlation is applicable to low-flow and low-pressure conditions, which are typical of those in the ACRR. The three hypotheses that examined the effect of the nonuniform axial heat flux distribution in the ACRR core are (1) the local conditions hypotheses, (2) the total power hypothesis, and (3) the global conditions hypothesis. These hypotheses, in conjunction with the UNM-CHF correlation, are used to estimate the CHF and CHFR in the ACRR. Because the total power hypothesis predictions of power per rod at CHF are approximately 15%-20% lower than those corresponding to saturation exit conditions, it can be concluded that the total power hypothesis considerably underestimates the CHF for nonuniformly heated geometries. This conclusion is in agreement with previous experimental results. The global conditions hypothesis, which is more conservative and more accurate of the other two, provides the most reliable predictions of CHF/CHFR for the ACRR. The global conditions hypothesis predictions of CHFR varied between 2.1 and 3.9, with the higher value corresponding to the lower water inlet temperature of 20 degrees C

  12. Comparison between different flux traps assembled in the core of the nuclear reactor IPEN/MB-01 by measuring of the thermal and epithermal neutron fluxes using activation foils

    International Nuclear Information System (INIS)

    Mura, Luiz Ernesto Credidio; Bitelli, Ulysses d'Utra; Mura, Luis Felipe Liambos; Carluccio, Thiago; Andrade, Graciete Simoes de

    2011-01-01

    The production of radioisotopes is one of the most important applications of nuclear research reactors. This study investigated a method called Flux Trap, which is used to increase the yield of production of radioisotopes in nuclear reactors. The method consists in the rearrangement of the fuel rods to allow the increase of the thermal neutron flux in the irradiation region inside the reactor core, without changing the standard reactor power level. Various configurations were assembled with the objective of finding the configuration with the highest thermal neutron flux in the region of irradiation. The method of activation analysis was used to measure the thermal neutron flux and determine the most efficient reactor core configuration . It was found that there was an increase in the thermal neutron flux of 337% in the most efficient configuration, which demonstrates the effectiveness of the method. (author)

  13. 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

  14. Test of In-core Flux Detectors in KNK II

    International Nuclear Information System (INIS)

    Hoppe, P.; Mitzel, F.

    1979-10-01

    The development of in-core detectors for Liquid Metal Fast Breeder Reactors (LMFBRs) is still in an early stage, and little operation experience is available. Therefore self-powered neutron and gamma detectors and neutron sensitive ionization chambers -especially developed for LMFBRs- have been tested in the Fast Sodium Cooled Test Reactor KNK II. Seven flux detectors have been installed in the core of KNK II by means of a special test rig. Five of them failed already within the first week during operation in the reactor. Due to measurements of electrical resistances and capacities, sodium penetrating into the detectors or cables probably seems to be the cause. As tests prior to the installation in the core proved the tightness of all detectors, it is suspected that small cracks have developed in the detector casings or in the outer cable sheaths during their exposure to the hot coolant. Two ionization chambers did not show these faults. However, one of them failed because the saturation current plateau disappeared and the other one's sensitivity decreased by a factor of five during the test period. It is suspected that in both cases changes of the filling gas might be involved

  15. Local carbon diffusion coefficient measurement in the S-1 spheromak

    International Nuclear Information System (INIS)

    Mayo, R.M.; Levinton, F.M.; Meyerhofer, D.D.; Chu, T.K.; Paul, S.F.; Yamada, M.

    1988-10-01

    The local carbon diffusion coefficient was measured in the S - 1 spheromak by detecting the radial spread of injected carbon impurity. The radial impurity density profile is determined by the balance of ionization and diffusion. Using measured local electron temperature T/sub e/ and density n/sub e/, the ionization rate is determined from which the particle diffusion coefficient is inferred. The results found in this work are consistent with Bohm diffusion. The absolute magnitude of D/sub /perpendicular// was determined to be (4/approximately/6) /times/ D/sub Bohm/. 25 refs., 13 figs., 2 tabs

  16. SORO post-simulations of Bruce A Unit 4 in-core flux detector verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Braverman, E.; Nainer, O. [Bruce Power, Nuclear Safety Analysis and Support Dept., Toronto, Ontario (Canada)]. E-mail: Evgeny.Braverman@brucepower.com; Ovidiu.Nainer@brucepower.com

    2004-07-01

    During the plant equipment assessment prior to requesting approval for restart of Bruce A Units 3 and 4 it was determined that all in-core flux detectors needed to be replaced. Flux detector verification tests were performed to confirm that the newly installed detectors had been positioned according to design specifications and that their response closely follows the calculated flux shape changes caused by selected reactivity mechanism movements. By comparing the measured and post-simulated RRS and NOP detector responses to various perturbations, it was confirmed that the new detectors are wired and positioned correctly. (author)

  17. Planetary cores, their energy flux relationship, and its implications

    Science.gov (United States)

    Johnson, Fred M.

    2018-02-01

    Integrated surface heat flux data from each planet in our solar system plus over 50 stars, including our Sun, was plotted against each object's known mass to generate a continuous exponential curve at an R-squared value of 0.99. The unexpected yet undeniable implication of this study is that all planets and celestial objects have a similar mode of energy production. It is widely accepted that proton-proton reactions require hydrogen gas at temperatures of about 15 million degrees, neither of which can plausibly exist inside a terrestrial planet. Hence, this paper proposes a nuclear fission mechanism for all luminous celestial objects, and uses this mechanism to further suggest a developmental narrative for all celestial bodies, including our Sun. This narrative was deduced from an exponential curve drawn adjacent to the first and passing through the Earth's solid core (as a known prototype). This trend line was used to predict the core masses for each planet as a function of its luminosity.

  18. Vertical and lateral flux on the continental slope off Pakistan: correlation of sediment core and trap results

    Science.gov (United States)

    Schulz, H.; von Rad, U.

    2014-06-01

    Due to the lack of bioturbation, the varve-laminated muds from the oxygen minimum zone (OMZ) off Pakistan provide a unique opportunity to precisely determine the vertical and lateral sediment fluxes in the nearshore part of the northeastern Arabian Sea. West of Karachi (Hab area), the results of two sediment trap stations (EPT and WPT) were correlated with 16 short sediment cores on a depth transect crossing the OMZ. The top of a distinct, either reddish- or light-gray silt layer, 210Pb-dated as AD 1905 ± 10, was used as an isochronous stratigraphic marker bed to calculate sediment accumulation rates. In one core, the red and gray layer were separated by a few (5-10) thin laminae. According to our varve model, this contributes water column above. All traps on the steep Makran continental slope show exceptionally high, pulsed winter fluxes of up to 5000 mg m-2 d-1. Based on core results, the flux at the seafloor amounts to 4000 mg m-2 d-1 and agrees remarkably well with the bulk winter flux of material, as well as with the flux of the individual bulk components of organic carbon, calcium carbonate and opal. However, due to the extreme mass of remobilized matter, the high winter flux events exceeded the capacity of the shallow traps. Based on our comparisons, we argue that high-flux events must occur regularly during winter within the upper OMZ off Pakistan to explain the high accumulations rates. These show distribution patterns that are a negative function of water depth and distance from the shelf. Some of the sediment fractions show marked shifts in accumulation rates near the lower boundary of the OMZ. For instance, the flux of benthic foraminifera is lowered but stable below ~1200-1300 m. However, flux and sedimentation in the upper eastern Makran area are dominated by the large amount of laterally advected fine-grained material and by the pulsed nature of the resuspension events at the upper margin during winter.

  19. Resistive stability of the cylindrical spheromak

    International Nuclear Information System (INIS)

    DeLucia, J.; Jardin, S.C.; Glasser, A.H.

    1983-11-01

    The growth rates for resistive instabilities in a straight circular cylinder with spheromak profiles are computed by using two complementary methods. The first method employs boundary layer analysis and asymptotic matching, most valid for values of the magnetic Reynolds number S greater than or equal to 10 5 . The second method solved the full linearized resistive MHD equations as an initial value problem, utilizing zone packing around the mode rational surface. Resolution requirements limit this to S less than or equal to 10 7 . The results from these two methods agree to better than 1 in 10 3 in the overlap region 10 7 greater than or equal to S greater than or equal to 10 5 . A scan of parameter space reveals that for parabolic q-profiles, the least unstable configurations have q 0 R/a approx. 0.67. The Hall term in Ohm's Law is easily incorporated into both methods. Recalculating the resistive MHD growth rates in the presence of this term shows that the resistive interchange mode is completely stabilized for a large enough value of the ion cyclotron time

  20. Plasma response to sustainment with imposed-dynamo current drive in HIT-SI and HIT-SI3

    Science.gov (United States)

    Hossack, A. C.; Jarboe, T. R.; Chandra, R. N.; Morgan, K. D.; Sutherland, D. A.; Penna, J. M.; Everson, C. J.; Nelson, B. A.

    2017-07-01

    The helicity injected torus—steady inductive (HIT-SI) program studies efficient, steady-state current drive for magnetic confinement plasmas using a novel experimental method. Stable, high-beta spheromaks have been sustained using steady, inductive current drive. Externally induced loop voltage and magnetic flux are oscillated together so that helicity and power injection are always positive, sustaining the edge plasma current indefinitely. Imposed-dynamo current drive (IDCD) theory further shows that the entire plasma current is sustained. The method is ideal for low aspect ratio, toroidal geometries with closed flux surfaces. Experimental studies of spheromak plasmas sustained with IDCD have shown stable magnetic profiles with evidence of pressure confinement. New measurements show coherent motion of a stable spheromak in response to the imposed perturbations. On the original device two helicity injectors were mounted on either side of the spheromak and the injected mode spectrum was predominantly n  =  1. Coherent, rigid motion indicates that the spheromak is stable and a lack of plasma-generated n  =  1 energy indicates that the maximum q is maintained below 1 during sustainment. Results from the HIT-SI3 device are also presented. Three inductive helicity injectors are mounted on one side of the spheromak flux conserver. Varying the relative injector phasing changes the injected mode spectrum which includes n  =  2, 3, and higher modes.

  1. In core fuel management optimization by varying the equilibrium cycle average flux shape for batch refuelled reactors

    International Nuclear Information System (INIS)

    Jong, A.J. de.

    1992-12-01

    We suggest a method to overcome this problem of optimization by varying reloading patterns by characterizing each particular reloading pattern by a set of intermediate parameters that are numbers. Plots of the objective function versus the intermediate parameters can be made. When the intermediate parameters represent the reloading patterns in a unique way, the optimum of the objective function can be found by interpolation within such plots and we can find the optimal reloading pattern in terms of intermediate parameters. These have to be transformed backwards to find an optimal reloading pattern. The intermediate parameters are closely related to the time averaged neutron flux shape in the core during an equilibrium cycle. This flux shape is characterized by a set of ratios of the space averaged fluxes in the fuel zones and the space averaged flux in the zone with the fresh fuel elements. An advantage of this choice of intermediate parameters is that it permits analytical calculation of equilibrium cycle fuel densities in the fuel zones for any applied reloading patten characterized by a set of equilibrium cycle average flux ratios and thus, provides analytical calculations of fuel management objective functions. The method is checked for the burnup of one fissile nuclide in a reactor core with the geometry of the PWR at Borssele. For simplicity, neither the conversion of fuel, nor the buildup of fission products were taken into account in this study. Since these phenomena can also be described by the equilibrium cycle average flux ratios, it is likely that this method can be extended to a more realistic method for global in core fuel management optimization. (orig./GL)

  2. Theory of edge plasma in a spheromak

    International Nuclear Information System (INIS)

    Hooper, E.B.

    1998-01-01

    Properties of the edge plasma in the SSPX spheromak during the plasma formation and sustainment phases are discussed. For the breakdown and formation phase, the main emphasis is on the analysis of possible plasma contamination by impurities from the electrodes of the plasma gun (helicity injector). The issue of an azimuthally uniform breakdown initiation is also discussed. After the plasma settles down in the main vacuum chamber, one has to sustain the current between the electrodes, in order to continuously inject helicity. We discuss properties of the plasma on the field lines intersecting the electrodes. We conclude that the thermal balance of this plasma is maintained by Joule heating competing with parallel heat losses to the electrodes. The resulting plasma temperature is in the range of 15 - 30 eV. Under the expected operational conditions, the ''current'' velocity of the electrons is only slightly below their thermal velocity. Implications of this observation are briefly discussed

  3. Design of spheromak injector using conical accelerator for large helical device

    Energy Technology Data Exchange (ETDEWEB)

    Miyazawa, J.; Yamada, H.; Yasui, K.; Kato, S. [National Inst. for Fusion Science, Toki, Gifu (Japan); Fukumoto, N.; Nagata, M.; Uyama, T. [Himeji Inst. of Tech., Hyogo (Japan)

    1999-11-01

    Optimization of CT injector for LHD has been carried out and conical electrode for adiabatic CT compression is adopted in the design. Point-model of CT acceleration in a co-axial electrode is solved to optimize the electrode geometry and the power supplies. Large acceleration efficiency of 34% is to be obtained with 3.2 m long conical accelerator and 40 kV - 42 kJ power supply. The operation scenario of a CT injector named SPICA mk. I (SPheromak Injector using Conical Accelerator) consisting of 0.8 m conical accelerator is discussed based on this design. (author)

  4. Local drift parameter, j/n/sub e/ and resistivity anomaly measurements in CTX spheromaks

    International Nuclear Information System (INIS)

    Hoida, H.W.; Barnes, C.W.; Henins, I.; Jarboe, T.R.; Marklin, G.; Buchenauer, C.J.; Knox, S.O.

    1985-01-01

    In a spheromak, the magnetic fields confining the plasma are generated primarily by internal currents rather than external coils. In order to provide information on the possible existence of current-driven microinstabilities, localized measurements of the ratio of the drift velocity of the electrons generating the internal current to their thermal velocity, V/sub d//V/sub th/ proportional to j/n/sub e/√T/sub e/ (known as the drift or streaming parameter), and j/n/sub e/ (proportional to V/sub d/) are needed. These microinstabilities are in some theories associated with an increase in the resistivity anomaly factor (eta/eta/sub Spitzer/). We present results on local measurements (at the magnetic axis) of the values of V/sub d//V/sub th/ and eta/eta/sub Spitzer/ by combining data from the spatially-resolved diagnostics employed on the CTX spheromak experiment, coupled with current density profile information from equilibrium measurements. The values of V/sub d//V/sub th/ and j/n/sub e/ appear to be correlated with local variations in eta/eta/sub Spitzer/, and can be changed by varying the plasma density. Data sets are presented for three values of n/sub e/

  5. Measurement and calculation of spatial and energetic neutron flux in the IEA-R1 reactor core

    International Nuclear Information System (INIS)

    Bittelli, U.D.

    1988-01-01

    This work presents spatial and energetic flux distribution measured in the IEA-R1 reactor core. The thermal neutron flux was measured by gold activation foils (bare and covered with cadmium) in the fuel element number 108 (reaction: 197 Au(n,γ) 198 Au) at 451W overall reactor power. The fast neutron flux was measured by indium activation foils (reaction: 115 In(n,n') 115m In) in the fuel elements number 94 at 4510W overall reactor power. The neutron energy spectrum was adjusted by SAND II code with the data produced by the irradiation of seven activation detectors in the fuel element number 94 at 4510 W overall reactor power. The following reactions were used: 58 Fe(n,γ) 59 Fe, 232 Th(n,γ) 233 Th, 197 Au(n,γ) 198 Au, 59 Co(n,γ) 60 Co, 54 Fe(n,p) 54 Mn, 24 Mg(n,p) 24 Na, 47 Ti(n,p) 47 Sc, 48 Ti(n,p) 48 Sc and 115 In(n,n') 115m In. The experimental results compared to those obtained by CITATION (spatial distribution flux) and HAMMER (energetic distribution flux) code, showed good agreement. The results presented in this work are a good contribution for a better knowledge of spatial and energetic neutron flux distribution in the IEA-R1 reactor core, besides that the experimental procedure is easily applicable to another situations. (autor) [pt

  6. Linking lowermost mantle structure, core-mantle boundary heat flux and mantle plume formation

    Science.gov (United States)

    Li, Mingming; Zhong, Shijie; Olson, Peter

    2018-04-01

    The dynamics of Earth's lowermost mantle exert significant control on the formation of mantle plumes and the core-mantle boundary (CMB) heat flux. However, it is not clear if and how the variation of CMB heat flux and mantle plume activity are related. Here, we perform geodynamic model experiments that show how temporal variations in CMB heat flux and pulses of mantle plumes are related to morphologic changes of the thermochemical piles of large-scale compositional heterogeneities in Earth's lowermost mantle, represented by the large low shear velocity provinces (LLSVPs). We find good correlation between the morphologic changes of the thermochemical piles and the time variation of CMB heat flux. The morphology of the thermochemical piles is significantly altered during the initiation and ascent of strong mantle plumes, and the changes in pile morphology cause variations in the local and the total CMB heat flux. Our modeling results indicate that plume-induced episodic variations of CMB heat flux link geomagnetic superchrons to pulses of surface volcanism, although the relative timing of these two phenomena remains problematic. We also find that the density distribution in thermochemical piles is heterogeneous, and that the piles are denser on average than the surrounding mantle when both thermal and chemical effects are included.

  7. Thermal neutron flux measurement using self-powered neutron detector (SPND) at out-core locations of TRIGA PUSPATI Reactor (RTP)

    Science.gov (United States)

    Ali, Nur Syazwani Mohd; Hamzah, Khaidzir; Mohamad Idris, Faridah; Hairie Rabir, Mohamad

    2018-01-01

    The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND’s signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

  8. Confinement and power balance in the S-1 spheromak

    Energy Technology Data Exchange (ETDEWEB)

    Levinton, F.M.; Meyerhofer, D.D.; Mayo, R.M.; Janos, A.C.; Ono, Y.; Ueda, Y.; Yamada, M.

    1989-07-01

    The confinement and scaling features of the S-1 spheromak have been investigated using magnetic, spectroscopic, and Thomson scattering data in conjunction with numerical modeling. Results from the multipoint Thomson scattering diagnostic shows that the central beta remains constant (/beta//sub to/ /approximately/ 5%) as the plasma current density increases from 0.68--2.1 MA/m/sup 2/. The density is observed to increase slowly over this range, while the central electron temperature increases much more rapidly. Analysis of the global plasma parameters shows a decrease in the volume average beta and energy confinement as the total current is increased. The power balance has been modeled numerically with a 0-D non-equilibrium time-dependent coronal model and is consistent with the experimental observations. 20 refs., 12 figs., 2 tabs.

  9. Confinement and power balance in the S-1 spheromak

    International Nuclear Information System (INIS)

    Levinton, F.M.; Meyerhofer, D.D.; Mayo, R.M.; Janos, A.C.; Ono, Y.; Ueda, Y.; Yamada, M.; Rochester Univ., NY; Los Alamos National Lab., NM; Princeton Univ., NJ

    1989-07-01

    The confinement and scaling features of the S-1 spheromak have been investigated using magnetic, spectroscopic, and Thomson scattering data in conjunction with numerical modeling. Results from the multipoint Thomson scattering diagnostic shows that the central beta remains constant (β to ∼ 5%) as the plasma current density increases from 0.68--2.1 MA/m 2 . The density is observed to increase slowly over this range, while the central electron temperature increases much more rapidly. Analysis of the global plasma parameters shows a decrease in the volume average beta and energy confinement as the total current is increased. The power balance has been modeled numerically with a 0-D non-equilibrium time-dependent coronal model and is consistent with the experimental observations. 20 refs., 12 figs., 2 tabs

  10. Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

    International Nuclear Information System (INIS)

    D'Utra Bitelli, U.

    1993-01-01

    This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)

  11. 0-D study of the compression of low temperature spheromaks

    International Nuclear Information System (INIS)

    Meyerhofer, D.D.; Hulse, R.A.; Zweibel, E.G.

    1985-09-01

    Compression of low temperature spheromak plasmas has been studied with the aid of a O-D two-fluid computer code. It is found that in a plasma which is radiation dominated, the electron temperature can be increased by up to a factor of seven for a compression of a factor of two, provided the temperature is above some critical value (approx.25eV) and the electron density particle confinement time product n/sub e/tau/sub p/ greater than or equal to 1 x 10 9 s/cm 3 . If the energy balance is dominated by particle confinement losses rather than radiation losses, the effect of compression is to raise the temperature as T/sub e/ approx.C/sup 6/5/, for constant tau/sub p/

  12. Performance Improvement of a Magnetized Coaxial Plasma Gun by adopting Iron-core Bias Coil and New Pre-Ionization System

    Science.gov (United States)

    Edo, Takahiro; Asai, T.; Tanaka, F.; Yamada, S.; Hosozawa, A.; Gota, H.; Roche, T.; Allfrey, I.; Matsumoto, T.

    2017-10-01

    A magnetized coaxial plasma gun (MCPG) is a device used to generate a compact toroid (CT), which has a spheromak-like configuration. A typical MCPG consists of a set of axisymmetric cylindrical electrodes, bias coil, and gas-puff valves. In order to expand the CT operating range, the distributions of the bias magnetic field and neutral gas have been investigated. We have developed a new means of generating stuffing flux. By inserting an iron core into the bias coil, the magnetic field increases dramatically; even a small current of a few Amps produces a sufficient bias field. According to a simulation result, it was also suggested that the radial distribution of the bias field is easily controlled. The ejected CT and the target FRC are cooled by excess neutral gas that typical MCPGs require to initiate a breakdown; therefore, we have adopted a miniature gun as a new pre-ionization (PI) system. By introducing this PI system, the breakdown occurs at lower neutral gas density so that the amount of excess neutral gas can be reduced.

  13. Energy conversion and concentration in a high-current gaseous discharge: Dense plasma spheromak in plasma focus experiments

    International Nuclear Information System (INIS)

    Kukushkin, A.B.; Rantsev-Kartinov, V.A.; Terentiev, A.R.

    1995-01-01

    Experimental results are presented which verify the possibility of the self-generated transformation of the magnetic field in plasma focus discharges to give a closed, spheromak-like magnetic configuration (SLMC). The energy conversion mechanism suggests a possibility of further concentrating the plasma power density by means of natural compressing the SLMC-trapped plasma by the residual magnetic field of the plasma focus discharge

  14. Reactivity And Neutron Flux At Silicide Fuel Element In The Core Of RSG-GAS

    International Nuclear Information System (INIS)

    Hamzah, Amir

    2000-01-01

    In order to 4.8 and 5.2 gr U/cm exp 3 loading of U 3 Si 2 --Al fuel plates characterization, he core reactivity change and neutron flux depression had been done. Control rod calibration method was used to reactivity change measurement and neutron flux distribution was measured using foil activation method. Measurement of insertion of A-type of testing fuel element with U-loading above cannot be done due to technical reason, so the measurement using full type silicide fuel element of 2.96 gr U/cm exp 3 loading. The reactivity change measurement result of insertion in A-9 and C-3 is + 2.67 cent. The flux depression at silicide fuel in A-9 is 1.69 times bigger than oxide and in C-3 is 0.68 times lower than oxide

  15. Development of SiC Neutron Detector Assembly to Measure the Neutron Flux of the Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Park, Se Hwan; Park, June Sic; Shin, Hee Sung; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yong Kyun [Hanyang University, Seoul (Korea, Republic of)

    2012-05-15

    At present, the conventional detector to measure the neutron at harsh environment is a Self Powered Neutron Detector (SPND). Rhodium(Rh)-103 is in the SPND. When neutron is incident on the Rhodium, the neutron capture reaction occurs, and the Rh-103 is converted to Rh-104. The Rh-104 is decayed to Pd-104 by {beta}-decay, and electrons are generated as the decay products. Because of the half life of Rh-104, approximately 5 minutes are required for the SPND output to reach the equilibrium condition. Therefore the on-line monitoring of the nuclear reactor state is limited if the neutron flux in the reactor core is monitored with the SPND. Silicon carbide (SiC) has the possibility to be developed as neutron detector at harsh environment, because the SiC can be operative at high temperature and high neutron flux conditions. Previously, the basic operation properties of the SiC detector were studied. Also, the radiation response of the SiC detector was studied at high neutron and gamma dose rate. The measurement results for an ex-core neutron flux monitor or a neutron flux monitor of the spent fuel were published. The SiC detector was also developed as neutron detector to measure the fissile material with active interrogation method. However, the studies about the development of SiC detector are still limited. In the present work, the radiation damage effect of the SiC detector was studied. The detector structure was determined based on the study, and a neutron detector assembly was made with the SiC detectors. The neutron and gamma-ray response of the detector assembly is presented in this paper. The detector assembly was positioned in the HANARO research reactor core, the performance test was done. The preliminary results are also included in this paper

  16. Application of AC servo motor on the in-core neutron flux instrumentation system

    International Nuclear Information System (INIS)

    Du Xiaoguang; Wang Mingtao

    2010-01-01

    The application of ac servo motor in the In-Core Neutron Flux Instrumentation System is described. The hardware component of ac servo motor control system is different from the dc motor control system. The effect of two control system on the instrumentation system is compared. The ac servo motor control system can improve the accuracy of the motion control, optimize the speed control and increase the reliability. (authors)

  17. Verification of the Taylor (minimum energy) state in the S-1 Spheromak

    International Nuclear Information System (INIS)

    Hart, G.W.; Janos, A.; Meyerhofer, D.D.; Yamada, M.

    1985-09-01

    Experimental measurements of the equilibrium in the S-1 Spheromak by use of magnetic probes inside the plasma show that the final magnetic equilibrium is one which has relaxed close to the Taylor (minimum-energy) state, even though the plasma is far from that state during formation. The comparison is made by calculating the two-dimensional μ profile of the plasma from the probe data, where μ is defined as μ 0 j/sub parallel//B. Measurements using a triple Langmuir probe provide evidence to support the conclusion that the pressure gradients in the relaxed state are confined to the edge region of the plasma

  18. Using robustness and preferred locations of archeomagnetic flux patches to constrain the physics of the core

    Science.gov (United States)

    Terra-Nova, F.; Amit, H.; Hartmann, G. A.; Trindade, R. I. F.

    2017-12-01

    Archaeomagnetic field models cover longer timescales than historical models and may therefore resolve the motion of geomagnetic features on the core-mantle boundary (CMB) in a more meaningful statistical sense. Here we perform a detailed appraisal of archaeomagnetic field models to infer some aspects of the physics of the outer core. We characterize and compare the identification and tracking of reversed flux patches (RFPs) in order to assess the RFPs robustness. We find similar behaviour within a family of models but differences among different families, suggesting that modelling strategy is more influential than data set. Similarities involve recurrent positions of RFPs, but no preferred direction of motion is found. The tracking of normal flux patches shows similar qualitative behaviour confirming that RFPs identification and tracking is not strongly biased by their relative weakness. We also compare the tracking of RFPs with that of the historical field model gufm1 and with seismic anomalies of the lowermost mantle to explore the possibility that RFPs have preferred locations prescribed by lower mantle lateral heterogeneity. The archaeomagnetic field model that most resembles the historical field is interpreted in terms of core dynamics and core-mantle thermal interactions. This model exhibits correlation between RFPs and low seismic shear velocity in co-latitude and a shift in longitude. These results shed light on core processes, in particular we infer toroidal field lines with azimuthal orientation below the CMB and large fluid upwelling structures with a width of about 80° (Africa) and 110° (Pacific) at the top of the core. Finally, similar preferred locations of RFPs in the past 9 and 3 kyr of the same archaeomagnetic field model suggest that a 3 kyr period is sufficiently long to reliably detect mantle control on core dynamics. This allows estimating an upper bound of 220-310 km for the magnetic boundary layer thickness below the CMB.

  19. Measurement and analysis of neutron flux distribution of STACY heterogeneous core by position sensitive proportional counter. Contract research

    CERN Document Server

    Murazaki, M; Uno, Y

    2003-01-01

    We have measured neutron flux distribution around the core tank of STACY heterogeneous core by position sensitive proportional counter (PSPC) to develop the method to measure reactivity for subcritical systems. The neutron flux distribution data in the position accuracy of +-13 mm have been obtained in the range of uranium concentration of 50g/L to 210g/L both in critical and in subcritical state. The prompt neutron decay constant, alpha, was evaluated from the measurement data of pulsed neutron source experiments. We also calculated distribution of neutron flux and sup 3 He reaction rates at the location of PSPC by using continuous energy Monte Carlo code MCNP. The measurement data was compared with the calculation results. As results of comparison, calculated values agreed generally with measurement data of PSPC with Cd cover in the region above half of solution height, but the difference between calculated value and measurement data was large in the region below half of solution height. On the other hand, ...

  20. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  1. Torque Characteristic Analysis of a Transverse Flux Motor Using a Combined-Type Stator Core

    Directory of Open Access Journals (Sweden)

    Xiaobao Yang

    2016-11-01

    Full Text Available An external rotor transverse flux motor using a combined-type stator core is proposed for a direct drive application in this paper. The stator core is combined by two kinds of components that can both be manufactured conveniently by generic laminated silicon steel used in traditional motors. The motor benefits from the predominance of low manufacturing cost and low iron loss by using a silicon-steel sheet. Firstly, the basic structure and operation principles of the proposed motor are introduced. Secondly, the expressions of the electromagnetic torque and the cogging torque are deduced by theoretical analysis. Thirdly, the basic characteristics such as permanent magnet flux linkage, no-load back electromotive force, cogging torque and electromagnetic torque are analyzed by a three-dimensional finite element method (3D FEM. Then, the influence of structure parameters on the torque density is investigated, which provides a useful foundation for optimum design of the novel motor. Finally, the torque density of the proposed motor is calculated and discussed, and the result shows that the proposed motor in this paper can provide considerable torque density by using few permanent magnets.

  2. EXPERIMENTAL DETERMINATION OF LONGITUDINAL COMPONENT OF MAGNETIC FLUX IN FERROMAGNETIC WIRE OF SINGLE-CORE POWER CABLE ARMOUR

    Directory of Open Access Journals (Sweden)

    I.A. Kostiukov

    2014-12-01

    Full Text Available A problem of determination of effective longitudinal magnetic permeability of single core power cable armour is defined. A technique for experimental determination of longitudinal component of magnetic flux in armour spiral ferromagnetic wire is proposed.

  3. The dynomak: An advanced spheromak reactor concept with imposed-dynamo current drive and next-generation nuclear power technologies

    Energy Technology Data Exchange (ETDEWEB)

    Sutherland, D.A., E-mail: das1990@uw.edu; Jarboe, T.R.; Morgan, K.D.; Pfaff, M.; Lavine, E.S.; Kamikawa, Y.; Hughes, M.; Andrist, P.; Marklin, G.; Nelson, B.A.

    2014-04-15

    A high-β spheromak reactor concept has been formulated with an estimated overnight capital cost that is competitive with conventional power sources. This reactor concept utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt (FLiBe) blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER-developed cryogenic pumping systems were implemented in this concept from the basis of technological feasibility. A tritium breeding ratio (TBR) of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%.

  4. Measurement of the energy spectrum of the neutrons inside the neutron flux trap assembled in the center of the reactor core IPEN/MB-01

    Energy Technology Data Exchange (ETDEWEB)

    Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto Credidio; Santos, Diogo Feliciano dos; Jerez, Rogerio; Mura, Luis Felipe Liamos, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    This paper presents the neutron energy spectrum in the central position of a neutron flux trap assembled in the core center of the research nuclear reactor IPEN/MB-01 obtained by an unfolding method. To this end, have been used several different types of activation foils (Au, Sc, Ti, Ni, and plates) which have been irradiated in the central position of the reactor core (setting number 203) at a reactor power level of 64.57 ±2.91 watts . The activation foils were counted by solid-state detector HPGe (gamma spectrometry). The experimental data of nuclear reaction rates (saturated activity per target nucleus) and a neutron spectrum estimated by a reactor physics computer code are the main input data to get the most suitable neutron spectrum in the irradiation position obtained through SANDBP code: a neutron spectra unfolding code that use an iterative adjustment method. The adjustment resulted in 3.85 ± 0.14 10{sup 9} n cm{sup -2} s{sup -1} for the integral neutron flux, 2.41 ± 0.01 10{sup 9} n cm{sup -2} s{sup -1} for the thermal neutron flux, 1.09 ± 0.02 10{sup 9} n cm{sup -2} s{sup -1} for intermediate neutron flux and 3.41± 0.02 10{sup 8} n cm{sup -2} s{sup -1} for the fast neutrons flux. These results can be used to verify and validate the nuclear reactor codes and its associated nuclear data libraries, besides show how much is effective the use of a neutron flux trap in the nuclear reactor core to increase the thermal neutron flux without increase the operation reactor power level. The thermal neutral flux increased 4.04 ± 0.21 times compared with the standard configuration of the reactor core. (author)

  5. Development of Eddy Current Technique for Reactor In-Core Flux Thimble Wear

    International Nuclear Information System (INIS)

    Park, S. S.; Jang, Y. Y.; Yim, C. Y.; Park, K. H.

    1990-01-01

    Since in-core flux thimble tube wear the due to flow-induced vibration could degrade the integrity of nuclear reactor, the effective detection and interpretation of the wear is important. In order to establish an inspection technique for thimble tubes, an eddy current experiment was performed to determine the optimum test frequency, defect sensitivity and evaluation accuracy. Eddy current probes were designed and fabricated with a theory. Specimens with artificial defects were fabricated using electro discharge machining method. The results from inspection technique developed and on-site inspection showed good applicability

  6. Measurement and analysis of neutron flux distribution of STACY heterogeneous core by position sensitive proportional counter. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Murazaki, Minoru; Uno, Yuichi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    We have measured neutron flux distribution around the core tank of STACY heterogeneous core by position sensitive proportional counter (PSPC) to develop the method to measure reactivity for subcritical systems. The neutron flux distribution data in the position accuracy of {+-}13 mm have been obtained in the range of uranium concentration of 50g/L to 210g/L both in critical and in subcritical state. The prompt neutron decay constant, {alpha}, was evaluated from the measurement data of pulsed neutron source experiments. We also calculated distribution of neutron flux and {sup 3}He reaction rates at the location of PSPC by using continuous energy Monte Carlo code MCNP. The measurement data was compared with the calculation results. As results of comparison, calculated values agreed generally with measurement data of PSPC with Cd cover in the region above half of solution height, but the difference between calculated value and measurement data was large in the region below half of solution height. On the other hand, calculated value agreed well with measurement data of PSPC without Cd cover. (author)

  7. Energy efficiency of the CTX magnetized coaxial plasma source

    International Nuclear Information System (INIS)

    Fernandez, J.C.; Barnes, C.W.; Jarboe, T.R.; Knox, S.O.; Platts, D.A.; McKenna, K.F.

    1985-01-01

    The energy efficiency of the CTX coaxial plasma source in creating spheromaks is determined experimentally to be in agreement with the theoretical prediction of lambda/sub sp//lambda/sub g/, where del x B = lambda/sub sp/ B in the spheromak, and lambda/sub g/ identical with μ 0 I/sub g//phi/sub g/ with I/sub g/ the source current and phi/sub g/ the magnetic flux through either source electrode. This is shown to be equivalent to magnetic helicity conservation. The spheromak impurity radiation was measured using an absolutely calibrated single chord bolometer system. The theoretical efficiency is within the experimental uncertainty of the ratio of spheromak radiated energy to source input energy in a group of ''dirty'' discharges. But the radiation measurement uncertainty is too large to determine whether a substantial part of the excess source energy not used in the production of spheromak magnetic energy is radiated from the spheromak volume

  8. Magnetohydrodynamic simulation of kink instability and plasma flow during sustainment of a coaxial gun spheromak

    International Nuclear Information System (INIS)

    Kanki, Takashi; Nagata, Masayoshi; Kagei, Yasuhiro

    2010-01-01

    Kink instability and the subsequent plasma flow during the sustainment of a coaxial gun spheromak are investigated by three-dimensional nonlinear magnetohydrodynamic simulations. Analysis of the parallel current density λ profile in the central open column revealed that the n = 1 mode structure plays an important role in the relaxation and current drive. The toroidal flow (v t ≈ 37 km/s) is driven by magnetic reconnection occurring as a result of the helical kink distortion of the central open column during repetitive plasmoid ejection and merging. (author)

  9. Influence of external toroidal flux on low-aspect-ratio toroidal plasma

    International Nuclear Information System (INIS)

    Ikuno, S.; Natori, M.; Kamitani, A.

    1999-01-01

    In the HIST device, the external flux is generated by two kinds of currents: the current I s flowing along the symmetry axis and the bias coil current I D . The influence of the external flux on the MHD equilibrium and stability of the low-aspect-ratio toroidal plasma in the HIST device is investigated numerically. Equilibrium configurations of the low-aspect-ratio toroidal plasma in the HIST device are numerically determined by means of the combination of FDM and BEM. The influence of I s and I D on their stability is also investigated by using the Mercier criterion. The results of computations show that the Mercier limit decreases to zero with increasing I s and with decreasing I D . Moreover, either a further increase in I s or a further decrease in I D raises the Mercier limit considerably. Besides, the equilibrium configuration in the HIST device changes its state from spheromak through ultra-low q to tokamak with increasing I s and with decreasing I D . (author)

  10. A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, E.A. [Department of Mechanical Engineering, University of Texas, Austin, TX (United States); Deinert, M.R. [Theoretical and Applied Mechanics, Cornell University, 219 Kimball Hall, Ithaca, NY 14853 (United States)]. E-mail: mrd6@cornell.edu; Cady, K.B. [Theoretical and Applied Mechanics, Cornell University, 219 Kimball Hall, Ithaca, NY 14853 (United States)

    2006-10-15

    The balance of isotopes in a nuclear reactor core is key to understanding the overall performance of a given fuel cycle. This balance is in turn most strongly affected by the time and energy-dependent neutron flux. While many large and involved computer packages exist for determining this spectrum, a simplified approach amenable to rapid computation is missing from the literature. We present such a model, which accepts as inputs the fuel element/moderator geometry and composition, reactor geometry, fuel residence time and target burnup and we compare it to OECD/NEA benchmarks for homogeneous MOX and UOX LWR cores. Collision probability approximations to the neutron transport equation are used to decouple the spatial and energy variables. The lethargy dependent neutron flux, governed by coupled integral equations for the fuel and moderator/coolant regions is treated by multigroup thermalization methods, and the transport of neutrons through space is modeled by fuel to moderator transport and escape probabilities. Reactivity control is achieved through use of a burnable poison or adjustable control medium. The model calculates the buildup of 24 actinides, as well as fission products, along with the lethargy dependent neutron flux and the results of several simulations are compared with benchmarked standards.

  11. Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of spatial and energetic distribution of neutron flux distribution

    International Nuclear Information System (INIS)

    Aredes, Vitor Ottoni Garcia

    2014-01-01

    In this work was conducted the mapping of the thermal and epithermal neutrons flux and the energy spectrum of the neutrons in the reactor core IPEN/MB-01 for a cylindrical core configuration with minor excess reactivity, which is 28 x 28 fuel rods arranged in north-south and east-west directions. The calibration of control rods for this configuration determined their excess reactivity. The lower excess reactivity in the core decreased neutron flux disturbance caused by the neutron absorbing rods , given that the nuclear reactor was operated with the rods almost completely removed . Was used the 'Activation Analysis Technique' with the thin foil activation detectors ( infinitely diluted and hyper-pure), of different materials that work in different energy ranges, to calculate the saturation activity, used for determining the neutron flux and in the SANDBP code as input for the calculation of the neutrons energy spectrum. To discriminate thermal and epithermal flux , was used the 'Cadmium RatioTechnique' . The activation detectors were distributed in a total of 140 radial and axial positions in the reactor core and 16 irradiation, with bare and covered with cadmium activation foils. A model of this configuration was simulated by MCNP-5 code to determine the cadmium correction factor and comparison of the results obtained experimentally. The cylindrical configuration desired, with 17% less fuel than the standard rectangular configuration (28 x 26 fuel rods), reached criticality with the control rods approximately 90% removed, which decreased considerably the disturbance in neutron flux. Given the highest power density of the 28 x 28 cylindrical core, the neutron flux increased by over 50% in the central regions of the core compared to the values of the 28 x 26 standard rectangular core. (author)

  12. The Influence of Heat Flux Boundary Heterogeneity on Heat Transport in Earth's Core

    Science.gov (United States)

    Davies, C. J.; Mound, J. E.

    2017-12-01

    Rotating convection in planetary systems can be subjected to large lateral variations in heat flux from above; for example, due to the interaction between the metallic cores of terrestrial planets and their overlying silicate mantles. The boundary anomalies can significantly reorganise the pattern of convection and influence global diagnostics such as the Nusselt number. We have conducted a suite of numerical simulations of rotating convection in a spherical shell geometry comparing convection with homogeneous boundary conditions to that with two patterns of heat flux variation at the outer boundary: one hemispheric pattern, and one derived from seismic tomographic imaging of Earth's lower mantle. We consider Ekman numbers down to 10-6 and flux-based Rayleigh numbers up to 800 times critical. The heterogeneous boundary conditions tend to increase the Nusselt number relative to the equivalent homogeneous case by altering both the flow and temperature fields, particularly near the top of the convecting region. The enhancement in Nusselt number tends to increase as the amplitude and wavelength of the boundary heterogeneity is increased and as the system becomes more supercritical. In our suite of models, the increase in Nusselt number can be as large as 25%. The slope of the Nusselt-Rayleigh scaling also changes when boundary heterogeneity is included, which has implications when extrapolating to planetary conditions. Additionally, regions of effective thermal stratification can develop when strongly heterogeneous heat flux conditions are applied at the outer boundary.

  13. Dynamics of spheromak-like compact toroids in a drift tube

    International Nuclear Information System (INIS)

    Suzuki, Y.; Kishimoto, Y.; Hayashi, T.

    2001-01-01

    In order to supply plasma fuel confined in spheromak-like compact toroids (SCTs) to a fusion device, the SCTs must be successfully guided through a drift tube region, in which they might be influenced by the magnetic field leaking from the fusion device. To reveal the SCT dynamics in a drift tube, MHD numerical simulations, where the SCTs are accelerated in a co-axial perfectly conducting cylinder with an external magnetic field, are carried out. In addition, the effect of an extended central electrode is examined by changing the length of the inner conducting cylinder. It is revealed that the SCT penetration depth is shorter than that estimated from the conventional conducting sphere model and that the SCTs are further decelerated by extending the inner conducting cylinder. These results are consistent with the results of the compact toroid injection experiment performed on the TEXT Upgrade tokamak. Finally, the deceleration mechanism of the SCTs is discussed by comparing the simulation result with the proposed theoretical model. (author)

  14. 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)

  15. FNR demonstration experiments Part II: Subcadmium neutron flux measurements

    International Nuclear Information System (INIS)

    Wehe, D.K.; King, J.S.

    1983-01-01

    The FNR HEU-LEU Demonstration Experiments include a comprehensive set of experiments to identify and quantify significant operational differences between two nuclear fuel enrichments. One aspect of these measurements, the subcadmium flux profiling, is the subject of this paper. The flux profiling effort has been accomplished through foil and wire activations, and by rhodium self-powered neutron detector (SPND) mappings. Within the experimental limitations discussed, the program to measure subcadmium flux profiles, lead to the following conclusions: (1) Replacement of a single fresh HEU element by a fresh LEU element at the center of an equilibrium HEU core produces a local flux depression. The ratio of HEU to LEU local flux is 1.19 ± .036, which is, well within experimental uncertainty, equal to the inverse of the U-235 masses for the two elements. (2) Whole core replacement of a large 38 element equilibrium HEU core by a fresh or nearly unburned LEU core reduces the core flux and raises the flux in both D 2 O and H 2 O reflectors. The reduction in the central core region is 40% to 10.0% for the small fresh 29 element LEU core, and 16% to 18% for a 31 element LEU core 482) with low average burnup 2 O reflector fluxes relative to core fluxes as measured by SPND with a fixed value of sensitivity, are in gross disagreement with the same flux ratios measured by Fe and Rh wire activations. Space dependent refinements of S are calculated to give some improvement in the discrepancy but the major part of the correction remains to be resolved

  16. Neutron flux distribution measurement in the Fort St. Vrain initial core (results of Fort St. Vrain start-up test A-7)

    International Nuclear Information System (INIS)

    Marshall, A.C.; Brown, J.R.

    1975-01-01

    A description is given of a test to measure the axial flux distribution at several radial locations in the Fort St. Vrain core representing unrodded, rodded, and partially rodded regions. The measurements were intended to verify the calculational accuracy of the three-dimensional calculational model used to compute axial power distributions for the Fort St. Vrain core. (U.S.)

  17. Nonlinear Model of Tape Wound Core Transformers

    Directory of Open Access Journals (Sweden)

    A. Vahedi

    2015-03-01

    Full Text Available Recently, tape wound cores due to their excellent magnetic properties, are widely used in different types of transformers. Performance prediction of these transformers needs an accurate model with ability to determine flux distribution within the core and magnetic loss. Spiral structure of tape wound cores affects the flux distribution and always cause complication of analysis. In this paper, a model based on reluctance networks method is presented for analysis of magnetic flux in wound cores. Using this model, distribution of longitudinal and transverse fluxes within the core can be determined. To consider the nonlinearity of the core, a dynamic hysteresis model is included in the presented model. Having flux density in different points of the core, magnetic losses can be calculated. To evaluate the validity of the model, results are compared with 2-D FEM simulations. In addition, a transformer designed for series-resonant converter and simulation results are compared with experimental measurements. Comparisons show accuracy of the model besides simplicity and fast convergence

  18. Effects of the current boundary conditions at the plasma-gun gap on density in SSPX

    Science.gov (United States)

    Kolesnikov, Roman; Lodestro, L. L.; Meyer, W. H.

    2012-10-01

    The Sustained Spheromak Physics Experiment (SSPX) was a toroidal magnetic-confinement device without toroidal magnetic-field coils or a central transformer but which generated core-plasma currents by dynamo processes driven by coaxial plasma-gun injection into a flux-conserving vessel. Record electron temperatures in a spheromak (Te˜500eV) were achieved, and final results of the SSPX program were reported in [1]. Plasma density, which depended strongly on wall conditions, was an important parameter in SSPX. It was observed that density rises with Igun and that confinement improved as the density was lowered. Shortly after the last experiments, a new feature was added to the Corsica code's solver used to reconstruct SSPX equilibria. Motivated by n=0 fields observed in NIMROD simulations of SSPX, an insulating boundary condition was implemented at the plasma-gun gap. Using this option we will perform new reconstructions of SSPX equilibria and look for correlations between the location of the separatrix (which moves up the gun wall and onto the insulating gap as Igun increases) and plasma density and magnetic-flux amplification [2].[4pt] [1] H. S. McLean, APS, DPP, Dallas, TX, 2008.[0pt] [2] E. B. Hooper et al., Nucl. Fusion 47, 1064 (2007).

  19. What does determine the sign of core in Magnetic Flux Rope structures of the Earth's magnetotail

    Directory of Open Access Journals (Sweden)

    D. V. Sarafopoulos

    2014-09-01

    Full Text Available This paper primarily examines the key factors being involved in precisely determining the sign of the core field in a magnetic flux rope (MFR like structure embedded in the tailward plasma flow associated with the Earth's magnetotail. Magnetic flux ropes are frequently detected by satellites moving smoothly northwards (upwards or southwards (downwards and crossing almost the whole plasma sheet; the sign of the rope's core is associated with the local tail's motion: If the tail is bending to an upward or downward direction, then the sign of the rope's core, being essentially an intense By deviation, will be positive or negative correspondingly. On the basis of this observational finding, a major question concerns the mechanism by which the tail's motion is dictated. The reconnection process acting in the tail will obviously produce symmetric structures of MFRs (with respect to the neutral sheet plane; therefore, the detected organized asymmetry may be an additional indication in the whole magnetotail' s dynamics. Moreover, we discuss the issue of the core's sign in cases without any significant magnetotail's motion. A model interpreting the diagnosed behavior is introduced: Once a tailward ion jet is produced in a thinned plasma sheet, it might form clockwise or counterclockwise ion vortices (i.e., loop-like ion currents providing the "magnetic core" with the appropriate sign. The crucial role of the interplanetary By deviation of the magnetic field (IMF is scrutinized and taken into account. The whole model is tested under the condition of long-lasting extraordinary events characterized by a persistent-intense By deviation with a duration up to 34 min. This work, based on Geotail single-satellite measurements, is not a statistical one; it is a first approach allowing the reconstruction of measurements in the whole range of the magnetotail's deflections, from negligible up to stronger significant magnetotail movements, and should be therefore

  20. Magnetohydrodynamic equilibrium and stability of spheromak with spheroidal plasma-vacuum interface

    International Nuclear Information System (INIS)

    Kaneko, Shobu; Kamitani, Atsushi; Takimoto, Akio.

    1985-05-01

    The analytic solutions to the Grad-Shafranov equation are obtained for a prolate and an oblate spheroidal plasma by using Hill's vortex model. Effects of a toroidal magnetic field Bsub(phi) on the MHD equilibrium configurations are investigated by using these analytic solutions. When Bsub(phi) is larger than that of the force-free configuration, the spheroidal plasmas in a vacuum magnetic field are shown to be unable in the MHD equilibrium. The several physical quantities on the equilibrium configuration are evaluated. The spheromak plasma is proved to be unstable if dp/d psi not equal 0 and d 2 V/d psi 2 >= 0 on the magnetic axis. Here p is the pressure and V(psi) the volume surrounded by a magnetic surface of psi=const. The equilibrium configurations of the spheroidal plasmas by using Hill's vortex model are shown to satisfy the above conditions, i.e., to be unstable. (author)

  1. Magnetohydrodynamic equilibrium and stability of spheromak with spheroidal plasma-vacuum interface

    International Nuclear Information System (INIS)

    Kaneko, Shobu; Kamitani, Atsushi; Takimoto, Akio

    1985-01-01

    The analytic solutions to the Grad-Shafranov equation are obtained for a prolate and an oblate spheroidal plasma by using Hill's vortex model. Effects of a toroidal magnetic field Bsub(phi) on the MHD equilibrium configurations are investigated by using these analytic solutions. When Bsub(phi) is stronger than that of the force-free configuration, the spheroidal plasmas in a vacuum magnetic field are shown to be unable in the MHD equilibrium. The several physical quantities on the equilibrium configuration are evaluated. The spheromak plasma is proved to be unstable if dp/d psi not equal 0 and d 2 V/d psi 2 >= 0 on the magnetic axis. Here p is the pressure and V(psi) the volume surrounded by a magnetic surface of psi = const. The equilibrium configurations of the spheroidal plasmas by using Hill's vortex model are shown to satisfy the above conditions, i.e., to be unstable. (author)

  2. Specification of ROP flux shape

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Gray, A [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1997-06-01

    The CANDU 9 480/SEU core uses 0.9% SEU (Slightly Enriched Uranium) fuel. The use f SEU fuel enables the reactor to increase the radial power form factor from 0.865, which is typical in current natural uranium CANDU reactors, to 0.97 in the nominal CANDU 9 480/SEU core. The difference is a 12% increase in reactor power. An additional 5% increase can be achieved due to a reduced refuelling ripple. The channel power limits were also increased by 3% for a total reactor power increase of 20%. This report describes the calculation of neutron flux distributions in the CANDU 9 480/SEU core under conditions specified by the C and I engineers. The RFSP code was used to calculate of neutron flux shapes for ROP analysis. Detailed flux values at numerous potential detector sites were calculated for each flux shape. (author). 6 tabs., 70 figs., 4 refs.

  3. Specification of ROP flux shape

    International Nuclear Information System (INIS)

    Min, Byung Joo; Gray, A.

    1997-06-01

    The CANDU 9 480/SEU core uses 0.9% SEU (Slightly Enriched Uranium) fuel. The use f SEU fuel enables the reactor to increase the radial power form factor from 0.865, which is typical in current natural uranium CANDU reactors, to 0.97 in the nominal CANDU 9 480/SEU core. The difference is a 12% increase in reactor power. An additional 5% increase can be achieved due to a reduced refuelling ripple. The channel power limits were also increased by 3% for a total reactor power increase of 20%. This report describes the calculation of neutron flux distributions in the CANDU 9 480/SEU core under conditions specified by the C and I engineers. The RFSP code was used to calculate of neutron flux shapes for ROP analysis. Detailed flux values at numerous potential detector sites were calculated for each flux shape. (author). 6 tabs., 70 figs., 4 refs

  4. Real time magnetic field and flux measurements for tokamak control using a multi-core PCI Express system

    International Nuclear Information System (INIS)

    Giannone, L.; Schneider, W.; McCarthy, P.J.; Sips, A.C.C.; Treutterer, W.; Behler, K.; Eich, T.; Fuchs, J.C.; Hicks, N.; Kallenbach, A.; Maraschek, M.; Mlynek, A.; Neu, G.; Pautasso, G.; Raupp, G.; Reich, M.; Schuhbeck, K.H.; Stober, J.; Volpe, F.; Zehetbauer, T.

    2009-01-01

    The existing real time system for the position and shape control in ASDEX Upgrade has been extended to calculate magnetic flux surfaces in real time using a multi-core PCI Express system running LabVIEW RT. The availability of reflective memory for LabVIEW RT will allow this system to be connected to the control system and other diagnostics in a multi-platform real time network. The measured response of each magnetic probe to the individual poloidal field coil currents in the absence of plasma current is compared to the calculated value. Prior to a tokamak discharge this comparison can be used to check for failure of the magnetic probe, flux loop or integrator.

  5. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    International Nuclear Information System (INIS)

    Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F.; Vermeeren, L.; Lopez, A. Legrand

    2011-01-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10 20 n/cm 2 . A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  6. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    Science.gov (United States)

    Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

    2011-03-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  7. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F. [CEA, DEN, Cadarache, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Lopez, A. Legrand [CEA, DEN, Saclay, SIREN/LECSI, F-91400 Saclay (France)

    2011-03-15

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10{sup 20} n/cm{sup 2}. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  8. Effect of inclusions on microstructure and toughness of deposited metals of self-shielded flux cored wires

    International Nuclear Information System (INIS)

    Zhang, Tianli; Li, Zhuoxin; Kou, Sindo; Jing, Hongyang; Li, Guodong; Li, Hong; Jin Kim, Hee

    2015-01-01

    The effect of inclusions on the microstructure and toughness of the deposited metals of self-shielded flux cored wires was investigated by optical microscopy, electron microscopy and mechanical testing. The deposited metals of three different wires showed different levels of low temperature impact toughness at −40 °C mainly because of differences in the properties of inclusions. The inclusions formed in the deposited metals as a result of deoxidation caused by the addition of extra Al–Mg alloy and ferromanganese to the flux. The inclusions, spherical in shape, were mixtures of Al 2 O 3 and MgO. Inclusions predominantly Al 2 O 3 and 0.3–0.8 μm in diameter were effective for nucleation of acicular ferrite. However, inclusions predominantly MgO were promoted by increasing Mg in the flux and were more effective than Al 2 O 3 inclusions of the same size. These findings suggest that the control of inclusions can be an effective way to improve the impact toughness of the deposited metal

  9. Core calculations of JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)

  10. An Equation Governing Ultralow-Velocity Zones: Implications for Holes in the ULVZ, Lateral Chemical Reactions at the Core-Mantle Boundary, and Damping of Heat Flux Variations in the Core

    Science.gov (United States)

    Hernlund, J. W.; Matsui, H.

    2017-12-01

    Ultralow-velocity zones (ULVZ) are increasingly illuminated by seismology, revealing surprising diversity in size, shape, and physical characteristics. The only viable hypotheses are that ULVZs are a compositionally distinct FeO-enriched dense material, which could have formed by fractional crystallization of a basal magma ocean, segregation of subducted banded iron formations, precipitation of solids from the outer core, partial melting and segregation of iron-rich melts from subducted basalts, or most likely a combination of many different processes. But many questions remain: Are ULVZ partially molten in some places, and not in others? Are ULVZ simply the thicker portions of an otherwise global thin layer, covering the entire CMB and thus blocking or moderating chemical interactions between the core and overlying mantle? Is such a layer inter-connected and able to conduct electrical currents that allow electro-magnetic coupling of core and mantle angular momentum? Are they being eroded and shrinking in size due to viscous entrainment, or is more material being added to ULVZ over time? Here we derive an advection-diffusion-like equation that governs the dynamical evolution of a chemically distinct ULVZ. Analysis of this equation shows that ULVZ should become readily swept aside by viscous mantle flows at the CMB, exposing "ordinary mantle" to the top of the core, thus inducing chemical heterogeneity that drives lateral CMB chemical reactions. These reactions are correlated with heat flux, thus maintaining large-scale pressure variations atop the core that induce cyclone-like flows centered around ULVZ and ponded subducted slabs. We suggest that turbulent diffusion across adjacent cyclone streams inside a stratified region atop the core readily accommodates lateral transport and re-distribution of components such as O and Si, in addition to heat. Our model implies that the deeper core is at least partly shielded from the influence of strong heat flux variations at

  11. Rational mapping (RAM) of in-core data

    International Nuclear Information System (INIS)

    Bonalumi, R.A.; Kherani, N.P.

    1985-01-01

    A unique processing of in-core flux detector data is described and demonstrated, such that the detailed in-core power distribution can be derived with great accuracy by combining a speciall ''smoothed-out'' set of in-core data with neutron diffusion theory. Rational Mapping (RAM) is designed in such a way that erratic detector signals are recognized very efficiently and can be eliminated from the experimental data set: This is achieved by modal expansion of the difference between theoretical fluxes and experimental fluxes at the detector sites. Sensitivity studies have shown that RAM is quite stable, does not absorb the ''wild'' detector error in the mapping procedure, and results in mapped fluxes with errors about three times smaller than would be obtained by direct interpolation of detector readings. A new method is described to infer corrections to theoretical core parameters based on the difference between the RAM fluxes and the theoretical fluxes

  12. Responses of platinum, vanadium and cobalt self-powered flux detectors near simulated booster rods in a ZED-2 mockup of a Bruce reactor core

    International Nuclear Information System (INIS)

    French, P.M.; Shields, R.B.; Kroon, J.C.

    1978-02-01

    The static responses of Pt, V and Co self-powered detectors have been compared with copper-foil neutron activation profiles in reference and perturbed Bruce reactor core mockups assembled in the ZED-2 test reactor at Chalk River Nuclear Laboratories. The results indicate that the normalized response of each self-powered detector is an accurate measure of the thermal-neutron flux at locations greater than one lattice pitch from either a booster rod or the core boundary. They indicate that, in the Bruce booster/detector configuration, the normalized static Pt response overestimates the neutron flux by less than 3.5% upon full booster-rod insertion. (author)

  13. Effect of texture and grain size on magnetic flux density and core loss in non-oriented electrical steel containing 3.15% Si

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K.M.; Park, S.Y. [Department of Materials Science and Engineering, Korea University, 5-1, Anam-dong, Sungbuk-Gu, Seoul 136-701 (Korea, Republic of); Huh, M.Y., E-mail: myhuh@korea.ac.kr [Department of Materials Science and Engineering, Korea University, 5-1, Anam-dong, Sungbuk-Gu, Seoul 136-701 (Korea, Republic of); Kim, J.S. [Electrical Steel Sheet Research Group, Technical Research Laboratories, POSCO, Goedong-dong, Pohang (Korea, Republic of); Engler, O. [Hydro Aluminium Rolled Products GmbH, R and D Center Bonn, P.O. Box 2468, D-53014 Bonn (Germany)

    2014-03-15

    In an attempt to differentiate the impact of grain size and crystallographic texture on magnetic properties of non-oriented (NO) electrical steel sheets, samples with different grain sizes and textures were produced and analyzed regarding magnetic flux density B and core loss W. The textures of the NO electrical steel samples could be precisely quantified with the help of elliptical Gaussian distributions. In samples with identical textures, small grain sizes resulted in about 15% higher core loss W than larger grains, whereas grain size only moderately affected the magnetic flux density B. In samples having nearly the same grain size, a correlation of the magneto-crystalline anisotropic properties of B and W with texture was obtained via the anisotropy parameter A(h{sup →}). With increasing A(h{sup →}) a linear decrease of B and a linear increase of W were observed. - Highlights: • We produced electrical steel sheets having different grain size and texture. • Magnetic flux density B and core loss W were varied with grain size and texture. • Correlation of B and W with texture was established via anisotropy parameter A(h{sup →}). • With increasing A(h{sup →}) a linear decrease of B and a linear increase of W were observed. • Grain size mainly affected W with only minor impact on B.

  14. Neutron flux distribution inside the cylindrical core of minor excess of reactivity in the IPEN/MB-01 reactor and comparison with citation code and MCNP- 5 code

    International Nuclear Information System (INIS)

    Aredes, Vitor Ottoni; Bitelli, Ulysses d'Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza

    2015-01-01

    This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10 8 ± 5.25% n/cm 2 s. (author)

  15. Neutron flux distribution inside the cylindrical core of minor excess of reactivity in the IPEN/MB-01 reactor and comparison with citation code and MCNP- 5 code

    Energy Technology Data Exchange (ETDEWEB)

    Aredes, Vitor Ottoni; Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10{sup 8} ± 5.25% n/cm{sup 2}s. (author)

  16. Neutron flux measurements in PUSPATI Triga Reactor

    International Nuclear Information System (INIS)

    Gui Ah Auu; Mohamad Amin Sharifuldin Salleh; Mohamad Ali Sufi.

    1983-01-01

    Neutron flux measurement in the PUSPATI TRIGA Reactor (PTR) was initiated after its commissioning on 28 June 1982. Initial measured thermal neutron flux at the bottom of the rotary specimen rack (rotating) and in-core pneumatic terminus were 3.81E+11 n/cm 2 sec and 1.10E+12n/cm 2 sec respectively at 100KW. Work to complete the neutron flux data are still going on. The cadmium ratio, thermal and epithermal neutron flux are measured in the reactor core, rotary specimen rack, in-core pneumatic terminus and thermal column. Bare and Cadmium covered gold foils and wires are used for the above measurement. The activities of the irradiated gold foils and wires are determined using Ge(Li) and hyperpure germinium detectors. (author)

  17. The S-1 Spheromak Control System

    International Nuclear Information System (INIS)

    Mathe, P.; Mika, R.; Oliaro, G.

    1983-01-01

    The use of a CAMAC based DEC LSI-11/23 microcomputer to perform all control functions for the S-1 Spheromak is described. The system monitors and controls the three coil systems, Toroidal, Poloidal, and Equilibrium field coils and their associated power sources, the water cooling system, the personnel and machine safety system, the machine and diagnostic timing system and the control room display and operator interface. Future requirements include control of the vacuum system, the gas injection system and interface to the PPPL Data Acquisition System DEC10. The computer is connected to five remotely located CAMAC crates by a fiber-optic serial highway operating at five megahertz. These crates contain interface modules required to control the S-1 experiment. These modules include: D/A and A/D converters, fast transient digitizers, timing modules, temperature sensing modules, CRT alphanumeric display drivers, watchdog timers, and relay and TTL parallel I/O ports. The computer itself resides in crate number0 and consists of an LSI-11/23 with hardware floating post processor, memory management, 256K bytes of memory, four RS-232 serial ports and a 30 megabyte hard disk with a one megabyte floppy disk backup. The majority of software is written in FORTRAN with a few speed critical programs written in PDP-11 MACRO assembly language. The software simulates a sequential state machine which allows easily changeable logic since all logic is represented by standard Boolean Fortran statements. The RSX-11/m operating system allows multiple tasks to be active simultaneously. This provides computing time for operator interactions, editing of critical machine parameters, data analysis and transmission of data to other computers while still maintaining the scan activity which constantly monitors machine parameters

  18. Comparison between measured and computed magnetic flux density distribution of simulated transformer core joints assembled from grain-oriented and non-oriented electrical steel

    Directory of Open Access Journals (Sweden)

    Hamid Shahrouzi

    2018-04-01

    Full Text Available The flux distribution in an overlapped linear joint constructed in the central region of an Epstein Square was studied experimentally and results compared with those obtained using a computational magnetic field solver. High permeability grain-oriented (GO and low permeability non-oriented (NO electrical steels were compared at a nominal core flux density of 1.60 T at 50 Hz. It was found that the experimental results only agreed well at flux densities at which the reluctance of different paths of the flux are similar. Also it was revealed that the flux becomes more uniform when the working point of the electrical steel is close to the knee point of the B-H curve of the steel.

  19. KSI's Cross Insulated Core Transformer Technology

    International Nuclear Information System (INIS)

    Uhmeyer, Uwe

    2009-01-01

    Cross Insulated Core Transformer (CCT) technology improves on Insulated Core Transformer (ICT) implementations. ICT systems are widely used in very high voltage, high power, power supply systems. In an ICT transformer ferrite core sections are insulated from their neighboring ferrite cores. Flux leakage is present at each of these insulated gaps. The flux loss is raised to the power of stages in the ICT design causing output voltage efficiency to taper off with increasing stages. KSI's CCT technology utilizes a patented technique to compensate the flux loss at each stage of an ICT system. Design equations to calculate the flux compensation capacitor value are presented. CCT provides corona free operation of the HV stack. KSI's CCT based High Voltage power supply systems offer high efficiency operation, high frequency switching, low stored energy and smaller size over comparable ICT systems.

  20. Determination of PWR core water level using ex-core detectors signals

    International Nuclear Information System (INIS)

    Bernal, Alvaro; Abarca, Agustin; Miro, Rafael; Verdu, Gumersindo

    2013-01-01

    The core water level provides relevant neutronic and thermalhydraulic information of the reactor such as power, k eff and cooling ability; in fact, core water level monitoring could be used to predict LOCA and cooling reduction which may deal with core damage. Although different detection equipment is used to monitor several parameters such as the power, core water level monitoring is not an evident task. However, ex-core detectors can measure the fast neutrons leaking the core and several studies demonstrate the existence of a relationship between fast neutron leakage and core water level due to the shielding effect of the water. In addition, new ex-core detectors are being developed, such as silicon carbide semiconductor radiation detectors, monitoring the neutron flux with higher accuracy and in higher temperatures conditions. Therefore, a methodology to determine this relationship has been developed based on a Monte Carlo calculation using MCNP code and applying variance reduction with adjoint functions based on the adjoint flux obtained with the discrete ordinates code TORT. (author)

  1. MAGNETIC FLUX EXPULSION IN STAR FORMATION

    International Nuclear Information System (INIS)

    Zhao Bo; Li Zhiyun; Nakamura, Fumitaka; Krasnopolsky, Ruben; Shang, Hsien

    2011-01-01

    Stars form in dense cores of magnetized molecular clouds. If the magnetic flux threading the cores is dragged into the stars, the stellar field would be orders of magnitude stronger than observed. This well-known 'magnetic flux problem' demands that most of the core magnetic flux be decoupled from the matter that enters the star. We carry out the first exploration of what happens to the decoupled magnetic flux in three dimensions, using a magnetohydrodynamic (MHD) version of the ENZO adaptive mesh refinement code. The field-matter decoupling is achieved through a sink particle treatment, which is needed to follow the protostellar accretion phase of star formation. We find that the accumulation of the decoupled flux near the accreting protostar leads to a magnetic pressure buildup. The high pressure is released anisotropically along the path of least resistance. It drives a low-density expanding region in which the decoupled magnetic flux is expelled. This decoupling-enabled magnetic structure has never been seen before in three-dimensional MHD simulations of star formation. It generates a strong asymmetry in the protostellar accretion flow, potentially giving a kick to the star. In the presence of an initial core rotation, the structure presents an obstacle to the formation of a rotationally supported disk, in addition to magnetic braking, by acting as a rigid magnetic wall that prevents the rotating gas from completing a full orbit around the central object. We conclude that the decoupled magnetic flux from the stellar matter can strongly affect the protostellar collapse dynamics.

  2. Neutron flux measurement by mobile detectors

    International Nuclear Information System (INIS)

    Verchain, M.

    1987-01-01

    Various incore instrumentation systems and their technological evolution are first reviewed. Then, for 1300 MWe PWR nuclear power plant, temperature and neutron flux measurement are described. Mobile fission chambers, with their large measuring range and accurate location allow a good knowledge of the core. Other incore measures are possible because of flux detector thimble tubes inserted in the reactor core [fr

  3. Combined analysis of neutron and photon flux measurements for the Jules Horowitz reactor core mapping

    Energy Technology Data Exchange (ETDEWEB)

    Fourmentel, D.; Villard, J. F.; Lyoussi, A. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Reynard-Carette, C. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France); Bignan, G.; Chauvin, J. P.; Gonnier, C.; Guimbal, P.; Malo, J. Y. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Carette, M.; Janulyte, A.; Merroun, O.; Brun, J.; Zerega, Y.; Andre, J. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France)

    2011-07-01

    We study the combined analysis of nuclear measurements to improve the knowledge of the irradiation conditions in the experimental locations of the future Jules Horowitz Reactor (JHR). The goal of the present work is to measure more accurately neutron flux, photon flux and nuclear heating in the reactor. In a Material Testing Reactor (MTR), nuclear heating is a crucial parameter to design the experimental devices to be irradiated in harsh nuclear conditions. This parameter drives the temperature of the devices and of the samples. The numerical codes can predict this parameter but in-situ measurements are necessary to reach the expected accuracy. For this reason, one objective of the IN-CORE program [1] is to study the combined measurements of neutron and photon flux and their cross advanced interpretation. It should be reminded that both neutron and photon sensors are not totally selective as their signals are due to neutron and photon interactions. We intend to measure the neutron flux by three different kinds of sensors (Uranium Fission chamber, Plutonium Fission chamber and Self Powered Neutron Detector), the photon flux by two different sensors (Ionization chamber and Self Powered Gamma Detector) and the nuclear heating by two different ones (Differential calorimeter and Gamma Thermometer). For the same parameter, we expect that the use of different kinds of sensors will allow a better estimation of the aimed parameter by mixing different spectrum responses and different neutron and gamma contributions. An experimental test called CARMEN-1 is scheduled in OSIRIS reactor (CEA Saclay - France) at the end of 2011, with the goal to map irradiation locations in the reactor reflector to get a first validation of the analysis model. This article focuses on the sensor selection for CARMEN-1 experiment and to the way to link neutron and photon flux measurements in view to reduce their uncertainties but also to better assess the neutron and photon contributions to nuclear

  4. Combined analysis of neutron and photon flux measurements for the Jules Horowitz reactor core mapping

    International Nuclear Information System (INIS)

    Fourmentel, D.; Villard, J. F.; Lyoussi, A.; Reynard-Carette, C.; Bignan, G.; Chauvin, J. P.; Gonnier, C.; Guimbal, P.; Malo, J. Y.; Carette, M.; Janulyte, A.; Merroun, O.; Brun, J.; Zerega, Y.; Andre, J.

    2011-01-01

    We study the combined analysis of nuclear measurements to improve the knowledge of the irradiation conditions in the experimental locations of the future Jules Horowitz Reactor (JHR). The goal of the present work is to measure more accurately neutron flux, photon flux and nuclear heating in the reactor. In a Material Testing Reactor (MTR), nuclear heating is a crucial parameter to design the experimental devices to be irradiated in harsh nuclear conditions. This parameter drives the temperature of the devices and of the samples. The numerical codes can predict this parameter but in-situ measurements are necessary to reach the expected accuracy. For this reason, one objective of the IN-CORE program [1] is to study the combined measurements of neutron and photon flux and their cross advanced interpretation. It should be reminded that both neutron and photon sensors are not totally selective as their signals are due to neutron and photon interactions. We intend to measure the neutron flux by three different kinds of sensors (Uranium Fission chamber, Plutonium Fission chamber and Self Powered Neutron Detector), the photon flux by two different sensors (Ionization chamber and Self Powered Gamma Detector) and the nuclear heating by two different ones (Differential calorimeter and Gamma Thermometer). For the same parameter, we expect that the use of different kinds of sensors will allow a better estimation of the aimed parameter by mixing different spectrum responses and different neutron and gamma contributions. An experimental test called CARMEN-1 is scheduled in OSIRIS reactor (CEA Saclay - France) at the end of 2011, with the goal to map irradiation locations in the reactor reflector to get a first validation of the analysis model. This article focuses on the sensor selection for CARMEN-1 experiment and to the way to link neutron and photon flux measurements in view to reduce their uncertainties but also to better assess the neutron and photon contributions to nuclear

  5. An iterative homogenization technique that preserves assembly core exchanges

    International Nuclear Information System (INIS)

    Mondot, Ph.; Sanchez, R.

    2003-01-01

    A new interactive homogenization procedure for reactor core calculations is proposed that requires iterative transport assembly and diffusion core calculations. At each iteration the transport solution of every assembly type is used to produce homogenized cross sections for the core calculation. The converged solution gives assembly fine multigroup transport fluxes that preserve macro-group assembly exchanges in the core. This homogenization avoids the periodic lattice-leakage model approximation and gives detailed assembly transport fluxes without need of an approximated flux reconstruction. Preliminary results are given for a one-dimensional core model. (authors)

  6. KoFlux: Korean Regional Flux Network in AsiaFlux

    Science.gov (United States)

    Kim, J.

    2002-12-01

    participate in this cooperative effort in advancing the flux network and we look forward to prosperous results. Acknowledgement. The financial support for KoFlux researchers were made by the Ministry of Environment (Next Generation Core Environmental Technology Development Program); Korea Science and Engineering Foundation (SRC Program); the Ministry of Science and Technology (NRL Program); and the Ministry of Education (Brain Korea 21 Program).

  7. Proposal of C-core Type Transverse Flux Motor for Ship Propulsion – Increasing Torque Density by Dense Stator Configuration –

    Directory of Open Access Journals (Sweden)

    Y. Yamamoto

    2014-02-01

    Full Text Available Electric ship propulsion system has been drawing attention as a solution for savings in energy and maintenance costs. The system is mainly composed of motor, converter and gearbox and required for high torque at low speed. In this situation, transverse flux motors (TFMs have been proposed to fulfill the low-speed high-torque characteristic due to suitable for short pole pitch and large number of poles to increase torque output. In this trend, we have proposed C-core type motors taking advantage of TFMs’ structure. In this manuscript, a simple design method based on the magnetic-circuit theory and simple modeling of the motor is proposed to search a design parameter for maximizing torque as a pre-process of numerical study. The method takes into consideration the effects of magnetic leakage flux, magnetic saturation and pole-core combination in accordance with the systematic theory. The simple modeling is conducted based on a dense armature structure in previous axial flux motors (AFMs applied to the new motor design. The validity of the method is verified by 3-D finite element analysis (FEA and relative error is at most 20%. The minimalist design is shown to be advantageous for effective use in 3-D FEA. As a detailed design by the FEA, high torque density and low cogging to output ratio can be achieved simultaneously in the proposed machine.

  8. Magnetohydrodynamic stability of spheromak plasma in toroidal flux conserver with rectangular cross section, 2

    International Nuclear Information System (INIS)

    Kaneko, Shobu; Tsutsui, Hiroaki; Miyazaki, Takeshi; Taguchi, Masayoshi.

    1985-08-01

    The magnetohydrodynamic equilibrium states by Hill's vortex model and by the Coulomb-wave-function model are proved to be unstable. New MHD equilibrium configurations are determined by using another model for which dp/dψ = 0 on the magnetic axis. Here p is the pressure and ψ is the flux function. The values of the safety factor on the magnetic axis, q axis , are evaluated for these configurations. The MHD stability of these equilibrium states is investigated by the Mercier criterion. The values of the maximum beta ratio β max are evaluated for this model. The optimized pressure distributions are determined by use of the Mercier criterion and the values of β max are also evaluated for these pressure distributions. The values of β max are shown to be at most 12 %, if the condition q axis < 1 is required. (author)

  9. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR

    International Nuclear Information System (INIS)

    Martinez C, E.

    2011-01-01

    One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-θ and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutron flux combining or summarizing the fluxes calculated by DORT r-θ, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutron flux and fluence, using three different meshes in the directions r, θ and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58 E + 12 n/cm 2 s, at a height H 4 (239.07 cm) and angle 32.236 o in the core shroud and 4.00 E + 09 n/cm 2 s at a height H 4 and angle 35.27 o in the inner wall of the reactor vessel, positions that are consistent to within ±10% over the ones reported in the literature. (Author)

  10. Automatic welding technologies for long-distance pipelines by use of all-position self-shielded flux cored wires

    Directory of Open Access Journals (Sweden)

    Zeng Huilin

    2014-10-01

    Full Text Available In order to realize the automatic welding of pipes in a complex operation environment, an automatic welding system has been developed by use of all-position self-shielded flux cored wires due to their advantages, such as all-position weldability, good detachability, arc's stability, low incomplete fusion, no need for welding protective gas or protection against wind when the wind speed is < 8 m/s. This system consists of a welding carrier, a guide rail, an auto-control system, a welding source, a wire feeder, and so on. Welding experiments with this system were performed on the X-80 pipeline steel to determine proper welding parameters. The welding technique comprises root welding, filling welding and cover welding and their welding parameters were obtained from experimental analysis. On this basis, the mechanical properties tests were carried out on welded joints in this case. Results show that this system can help improve the continuity and stability of the whole welding process and the welded joints' inherent quality, appearance shape, and mechanical performance can all meet the welding criteria for X-80 pipeline steel; with no need for windbreak fences, the overall welding cost will be sharply reduced. Meanwhile, more positive proposals were presented herein for the further research and development of this self-shielded flux core wires.

  11. Influence of preheating on API 5L-X80 pipeline joint welding with self shielded flux-cored wire

    International Nuclear Information System (INIS)

    Cooper, R.; Silva, J. H. F.; Trevisan, R. E.

    2004-01-01

    The present work refers to the characterization of API 5L-X80 pipeline joints welded with self-shielded flux cored wire. This process was evaluated under preheating conditions, with an uniform and steady heat input. All joints were welded in flat position (1G), with the pipe turning and the torch still. Tube dimensions were 762 mm in external diameter and 16 mm in thickness. Welds were applied on single V-groove, with six weld beads, along with three levels of preheating temperatures (room temperature, 100 degree centigree, 160 degree centigree). These temperatures were maintained as inter pass temperature. The filler metal E71T8-K6 with mechanical properties different from parent metal was used in under matched conditions. The weld characterization is presented according to the mechanical test results of tensile strength, hardness and impact test. The mechanical tests were conducted according to API 1104, AWS and ASTM standards. API 1104 and API 51 were used as screening criteria. According to the results obtained, it was possible to remark that it is appropriate to weld API 5L-X80 steel ducts with Self-shielded Flux Cored wires, in conformance to the API standards and no preheat temperature is necessary. (Author) 22 refs

  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. Survey of the thermal and fast neutron flux distribution in the core of IPR-R1 reactor

    International Nuclear Information System (INIS)

    Guimaraes, R.R.R.; Santoro, C.A.B.

    1984-01-01

    A methodology to obtain the neutron flux distribution inside the core is presented, aiming to analize the project of reactor increasing power. The technique of measures by activation with irradiation of steel eletrodes of 700 mm of lenght, put in acrylic rods was used. In the detection process and in the counting of activation product, a Ge(Li) detector with high resolution and a scanning mechanical system, constructed and projected in CDTN (Nuclear Technology Development Center) were used. (E.G.) [pt

  14. Survey of the thermal and fast neutron flux distribution in the core of IPR-R1 reactor

    International Nuclear Information System (INIS)

    Guimaraes, R.R.R.

    1985-01-01

    A methodology to obtain the neutron flux distribution inside the core of a reactor is presented, aiming to analyze specifications for increasing reactor power. The activation measurement technique with irradiation of steel eletrodes of 700 mm of lenght, put in acrylic rods was used. In the detection process and in the counting of activation product, a Ge (Li) detector with high resolution and a scanning mechanical system, constructed and projected in CDTN (Nuclear Technology Development Center) were used. (E.G.) [pt

  15. Detent Force Reduction of a C-Core Linear Flux-Switching Permanent Magnet Machine with Multiple Additional Teeth

    Directory of Open Access Journals (Sweden)

    Yi Du

    2017-03-01

    Full Text Available C-core linear flux-switching permanent magnet (PM machines (LFSPMs are attracting more and more attention due to their advantages of simplicity and robustness of the secondary side, high power density and high torque density, in which both PMs and armature windings are housed in the primary side. The primary salient tooth wound with a concentrated winding consists of C-shaped iron core segments between which PMs are sandwiched and the magnetization directions of these PMs are adjacent and alternant in the horizontal direction. On the other hand, the secondary side is composed of a simple iron core with salient teeth so that it is very suitable for long stroke applications. However, the detent force of the C-core LFSPM machine is relatively high and the magnetic circuit is unbalanced due to the end effect. Thus, a new multiple additional tooth which consists of an active and a traditional passive additional tooth, is employed at each end side of the primary in this paper, so that the asymmetry due to end effect can be depressed and the detent force can be reduced by adjusting the passive additional tooth position. By using the finite element method, the characteristics and performances of the proposed machine are analyzed and verified.

  16. Three dimensional simulation study of spheromak injection into magnetized plasmas

    International Nuclear Information System (INIS)

    Suzuki, Y.; Watanabe, T.H.; Sato, T.; Hayashi, T.

    2000-01-01

    The three dimensional dynamics of a spheromak-like compact toroid (SCT) plasmoid, which is injected into a magnetized target plasma region, is investigated by using MHD numerical simulations. It is found that the process of SCT penetration into this region is much more complicated than that which has been analysed so far by using a conducting sphere (CS) model. The injected SCT suffers from a tilting instability, which grows with a similar timescale to that of the SCT penetration. The instability is accompanied by magnetic reconnection between the SCT magnetic field and the target magnetic field, which disrupts the magnetic configuration of the SCT. Magnetic reconnection plays a role in supplying the high density plasma, initially confined in the SCT magnetic field, to the target region. The penetration depth of the SCT high density plasma is also examined. It is shown to be shorter than that estimated from the CS model. The SCT high density plasma is decelerated mainly by the Lorentz force of the target magnetic field, which includes not only the magnetic pressure force but also the magnetic tension force. Furthermore, by comparing the SCT plasmoid injection with the bare plasmoid injection, magnetic reconnection is considered to relax the magnetic tension force, i.e. the deceleration of the SCT plasmoid. (author)

  17. RAtional Mapping (RAM) of in-core data

    International Nuclear Information System (INIS)

    Bonalumi, R.A.; Kherani, N.P.

    1983-01-01

    The paper describes and demonstrates a unique processing of in-core flux detector data, such that the detailed in-core power distribution can be derived with great accuracy by combining a specially 'smoothed-out' set of in-core data with neutron diffusion theory. RAM is designed in such a way that erratic detector signals are recognized very efficiently and can be eliminated from the experimental data set: this is achieved by modal expansion of the difference between theoretical fluxes and experimental fluxes at the detector sites. Sensitivity studies have shown that RAM is quite stable, does not absorb the 'wild' detector errors in the mapping procedure and results in mapped fluxes with errors about three times smaller than would be obtained by direct interpolation of detector readings

  18. Core mechanics and configuration behavior of advanced LMFBR core restraint concepts

    International Nuclear Information System (INIS)

    Fox, J.N.; Wei, B.C.

    1978-02-01

    Core restraint systems in LMFBRs maintain control of core mechanics and configuration behavior. Core restraint design is complex due to the close spacing between adjacent components, flux and temperature gradients, and irradiation-induced material property effects. Since the core assemblies interact with each other and transmit loads directly to the core restraint structural members, the core assemblies themselves are an integral part of the core restraint system. This paper presents an assessment of several advanced core restraint system and core assembly concepts relative to the expected performance of currently accepted designs. A recommended order for the development of the advanced concepts is also presented

  19. Neutron flux parameters for k{sub 0}-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

    Energy Technology Data Exchange (ETDEWEB)

    Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Sarmani, S. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Masood, Z. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia)

    2011-02-15

    The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutron flux parameters as thermal to epithermal neutron flux ratio (f), epithermal neutron flux shape factor ({alpha}), thermal neutron flux ({phi}{sub th}) and epithermal neutron flux ({phi}{sub epi}) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k{sub 0}-neutron activation analysis (k{sub 0}-NAA).The f and {alpha} were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of f and {alpha} ranged from 33.49 to 47.33 and -0.07 to -0.14, respectively. The {phi}{sub th} and the {phi}{sub epi} were measured as 2.03 x 10{sup 12} (cm{sup -2} s{sup -1}) and 6.05 x 10{sup 10} (cm{sup -2} s{sup -1}) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

  20. Impact of shelf life on measured prompt fraction of spare Inconel in-core flux detectors

    Energy Technology Data Exchange (ETDEWEB)

    Mohindra, VK; Sadeghi, S. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Crouse, B. [Darlington Nuclear Generating Station, Bowmanville, Ontario (Canada)

    2008-07-01

    Prompt fraction measurements associated with spare self-powered Inconel In-Core Flux Detectors (ICFDs) carried out a few years after installation on Shut Down System number 1 (SDS1) and Reactor Regulating System (RRS) at Darlington Nuclear Generating Station (DNGS), were found to be lower than those of the original detectors. These detectors, spares and originals, were manufactured in the late 80s, however, the former were kept at manufacturer's warehouse and latter were installed in the reactor core within a few years after manufacturing. Although the prompt fractions of the spare detectors were relatively low, the electronic/electrical behavior of the spare detectors was intact. The first batch of the original detectors performed as per the design requirements. Therefore, it is suspected that during shelf life, spare Inconel in-core flux detectors underwent changes that lowered their measured values of prompt fraction, which were taken within a few years after installation in the reactor. Detailed study of detectors' material composition and impurity concentrations revealed no association with the lower prompt fraction measurements. The evaluation of the limited data of the original and spare Inconel ICFDs installed at Darlington showed: 1. The reduction in prompt fraction was roughly proportional to the shelf life of the detectors; and 2. The rate of reduction in prompt fraction during storage was about double the rate of reduction during operation in the reactor. Above observations were based on the data provided by DNGS for a few detectors. The purpose of this paper is two fold, firstly to present the results of the complete study carried out to investigate the cause of relatively low prompt fractions measured on spare SDS1 and RRS Inconel ICFDs at DNGS, and secondly to generate interest/awareness within other CANDU utilities to add to the database of prompt fractions of spare Inconel ICFDs measured after installation. The data will help to improve

  1. Magnetic flux concentration methods for magnetic energy harvesting module

    Directory of Open Access Journals (Sweden)

    Wakiwaka Hiroyuki

    2013-01-01

    Full Text Available This paper presents magnetic flux concentration methods for magnetic energy harvesting module. The purpose of this study is to harvest 1 mW energy with a Brooks coil 2 cm in diameter from environmental magnetic field at 60 Hz. Because the harvesting power is proportional to the square of the magnetic flux density, we consider the use of a magnetic flux concentration coil and a magnetic core. The magnetic flux concentration coil consists of an air­core Brooks coil and a resonant capacitor. When a uniform magnetic field crossed the coil, the magnetic flux distribution around the coil was changed. It is found that the magnetic field in an area is concentrated larger than 20 times compared with the uniform magnetic field. Compared with the air­core coil, our designed magnetic core makes the harvested energy ten­fold. According to ICNIRP2010 guideline, the acceptable level of magnetic field is 0.2 mT in the frequency range between 25 Hz and 400 Hz. Without the two magnetic flux concentration methods, the corresponding energy is limited to 1 µW. In contrast, our experimental results successfully demonstrate energy harvesting of 1 mW from a magnetic field of 0.03 mT at 60 Hz.

  2. Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

    International Nuclear Information System (INIS)

    Heard, F.J.

    1997-01-01

    The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity

  3. Using MCNP for in-core instrument calibration in CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, D.C. [Point Lepreau Generating Station, NB Power, Lepreau, New Brunswick (Canada); Anghel, V.N.P.; Sur, B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2002-07-01

    The calibration of in-core instruments is important for safe and economical CANDU operation. However, in-core detectors are not normally suited to bench calibration procedures. This paper describes the use and validation of detailed neutron transport calculations for the purpose of calibrating the response of in-core neutron flux detectors. The Monte-Carlo transport code, MCNP, was used to model the thermal neutron flux distribution in the region around self-powered in-core flux detectors (ICFDs), and in the vicinity of the calandria edge. The ICFD model was used to evaluate the reduction in signal of a given detector (the 'detector shading factor') due to neutron absorption in surrounding materials, detectors, and lead-cables. The calandria edge model was used to infer the accuracy of the calandria edge position from flux scans performed by AECL's traveling flux detector (TFD) system. The MCNP results were checked against experimental results on ICFDs, and also against shading factors computed by other means. The use of improved in-core detector calibration factors obtained by this new methodology will improve the accuracy of spatial flux control performance in CANDU-6 reactors. The accurate determination of TFD based calandria edge position is useful in the quantitative measurement of changes in in-core component dimensions and position due to aging, such as pressure tube sag. (author)

  4. Calibration of the nuclear power channels for the cylindrical configuration of the IPEN/MB-01 reactor obtained from the measurements of the spatial neutron flux distribution in the reactor core through the irradiation of gold foils

    Energy Technology Data Exchange (ETDEWEB)

    Bitelli, Ulysses d' Utra; Silva, Alexandre F. Povoa da; Mura, Luiz Ernesto Credidio; Aredes, Vitor Ottoni Garcia; Santos, Diogo Feliciano dos, E-mail: ubitelli@ipen.br, E-mail: alexpovoa@yahoo.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The activation foil is one of the most used techniques to obtain and compare nuclear parameters from the nuclear data libraries, given by a gamma spectrometry system. Through the measurements of activity induced in the foils, it is possible to determine the neutron flux profile exactly where it has been irradiated. The power level operation of the reactor is a parameter directly proportional to the average neutron flux in the core. The objective of this work is to obtain, for a cylindrical configuration, the power generation through a spatial thermal neutron flux distribution in the core of IPEN/MB-01 Reactor, by irradiating gold foils positioned symmetrically into the core. They are put in a Lucite plate which will not interfere in the analysis of the neutron flux, because of its low microscopic absorption cross section for the analyzed neutrons. The foils are irradiated with and without cadmium covered small plates, to obtain the thermal and epithermal neutron flux, through specific equations. The correlation between the average power neutron flux, as a result of the foil's irradiation, and the average power digital neutron flux of the nuclear power channels, allows the calibration of the nuclear channels of the reactor. This same correlation was done in 2008 with the reactor in a rectangular configuration, which resulted in a specific calibration of the power level operation. This calibration cannot be used in the cylindrical configuration, because the nuclear parameters could change, which may lead to a different neutron profile. Furthermore, the precise knowledge of the power neutron flux in the core also validates the mathematics used to calculate the power neutron flux. (author)

  5. Nonlinear Dynamic Model of PMBLDC Motor Considering Core Losses

    DEFF Research Database (Denmark)

    Fasil, Muhammed; Mijatovic, Nenad; Jensen, Bogi Bech

    2017-01-01

    The phase variable model is used commonly when simulating a motor drive system with a three-phase permanent magnet brushless DC (PMBLDC) motor. The phase variable model neglects core losses and this affects its accuracy when modelling fractional-slot machines. The inaccuracy of phase variable mod...... on the detailed analysis of the flux path and the variation of flux in different components of the machine. A prototype of fractional slot axial flux PMBLDC in-wheel motor is used to assess the proposed nonlinear dynamic model....... of fractional-slot machines can be attributed to considerable armature flux harmonics, which causes an increased core loss. This study proposes a nonlinear phase variable model of PMBLDC motor that considers the core losses induced in the stator and the rotor. The core loss model is developed based...

  6. Automated reactivity anomaly surveillance in the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Knutson, B.J.; Harris, R.A.; Honeyman, D.J.; Shook, A.T.; Krohn, C.N.

    1985-01-01

    The automated technique for monitoring core reactivity during power operation used at the Fast Flux Test Facility (FFTF) is described. This technique relies on comparing predicted to measured rod positions to detect any anomalous (or unpredicted) core reactivity changes. It is implemented on the Plant Data System (PDS) computer and, thus, provides rapid indication of any abnormal core conditions. The prediction algorithms use thermal-hydraulic, control rod position and neutron flux sensor information to predict the core reactivity state

  7. PODESY program for flux mapping of CNA II reactor:

    International Nuclear Information System (INIS)

    Ribeiro Guevara, Sergio

    1988-01-01

    The PODESY program, developed by KWU, calculates the spatial flux distribution of CNA II reactor through a three-dimensional expansion of 90 incore detector measurements. The calculation is made in three steps: a) short-term calculation which considers the control rod positions and it has to be done each time the flux mapping is calculated; b) medium-term calculation which includes local burn-up dependent calculation made by diffusion methods in macro-cell configurations (seven channels in hexagonal distribution), and c) long-term calculation, or macroscopic flux determination, that is a fitting and expansion of measured fluxes, previously corrected by local effects, using the eigen functions of the modified diffusion equation. The paper outlines development of step (c) of the calculation. The incore detectors have been located in the central zone of the core. In order to obtain low errors in the expansion procedure it is necessary to include additional points, whose flux values are assumed to be equivalent to detector measurements. These flux values are calculated with detector measurements and a spatial flux distribution calculated by a PUMA code. This PUMA calculation employs a smooth burn-up distribution (local burn-up variations are considered in step (b) of the whole calculation) representing the state of core evolution at the calculation time. The core evolution referred to ends when the equilibrium core condition is reached. Additionally, a calculation method to be employed in the plant in case of incore detector failures, is proposed. (Author) [es

  8. 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)

  9. Theoretical aspects of the use of pulsed reflectometry in a spheromak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, B. J., LLNL

    1998-06-11

    Pulsed reflectometry using both ordinary (O) and extraordinary (X) modes has the potential of providing time and space-resolved measurements of the electron density, the magnitude of the magnetic field, and the magnetic shear as a function of radius. Such a diagnostic also yields the current profile from the curl of the magnetic field. This research addresses theoretical issues associated with the use of reflectometry in the SSPX spheromak experiment at the Lawrence Livermore National Laboratory. We have extended a reflectometry simulation model to accommodate O and X-mode mixed polarization and linear mode conversion between the two polarizations. A Wentzel-Kramers-Brillouin-Jeffreys (WKBJ) formula for linear mode conversion agrees reasonably well with direct numerical solutions of the wave equation, and we have reconstructed the magnetic pitch-angle profile by matching the results of the WKBJ formula with the mode conversion data observed in simulations using a least-squares determination of coefficients in trial functions for the profile. The reflectometry data also yield information on fluctuations. Instrumental issues, e.g., the effects of microwave mixers and filters on model reflectometry pulses, have been examined to optimize the performance of the reflectometry diagnostics.

  10. First In-Core Simultaneous Measurements of Nuclear Heating and Thermal Neutron Flux Obtained With the Innovative Mobile Calorimeter CALMOS Inside the OSIRIS Reactor

    Science.gov (United States)

    Carcreff, Hubert; Salmon, Laurent; Bubendorff, Jacques; Lepeltier, Valérie

    2016-10-01

    Nuclear heating inside a MTR reactor has to be known in order to design and run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. Calorimeter working modes, measurement procedures, main modeling and experimental results and expected advantages of this new technique have been already presented in previous papers. However, these first in-core measurements were not performed beyond 6 W · g-1, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutron flux obtained by the CALMOS device at 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutron flux up to 2.7 1014 n · cm-2 · s-1 and nuclear heating up to 12 W · g-1. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a comparison is made between results obtained by the probe calibration coefficient and the zero methods. Thermal neutron flux evaluation from SPND signal processing required a specific TRIPOLI-4 Monte Carlo calculation which has been performed with the precise CALMOS cell geometry. In addition, the Finite Element model for temperatures map prediction inside the calorimetric cell has been upgraded with recent experimental data obtained up to 12 W · g-1. Finally, the experience feedback led us to improvement perspectives. A second device is

  11. Use of sup(233)U for high flux reactors

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Liem, P.H.

    1991-01-01

    The feasibility design study on the graphite moderated gas cooled reactor as a high flux reactor has been performed. The core of the reactor is equipped with two graphite reflectors, i.e., the inner reflector and the outer reflector. The highest value of the thermal neutron flux and moderately high thermal neutron flux are expected to be achieved in the inner reflector region and in the outer reflector region respectively. This reactor has many merits comparing to the conventional high flux reactors. It has the inherent safety features associated with the modular high temperature reactors. Since the core is composed with pebble bed, the on-power refueling can be performed and the experiment time can be chosen as long as necessary. Since the thermal-to-fast flux ratio is large, the background neutron level is low and material damage induced by fast neutrons are small. The calculation was performed using a four groups diffusion approximation in a one-dimensional spherical geometry and a two-dimensional cylindrical geometry. By choosing the optimal values of the core-reflector geometrical parameters and moderator-to-fuel atomic density, high thermal neutron flux can be obtained. Because of the thermal neutron flux can be obtained. Because of the thermal design constraint, however, this design will produce a relatively large core volume (about 10 7 cc) and consequently a higher reactor power (100 MWth). Preliminary calculational results show that with an average power density of only 10 W/cc, maximum thermal neutron flux of 10 15 cm -2 s -1 can be achieved in the inner reflector. The eta value of 233 U is larger than 235 U. By introducing 233 U as the fissile material for this reactor, the thermal neutron flux level can be increased by about 15%. (author). 3 refs., 2 figs., 4 tabs

  12. Reactor core flow rate control system

    International Nuclear Information System (INIS)

    Sakuma, Hitoshi; Tanikawa, Naoshi; Takahashi, Toshiyuki; Miyakawa, Tetsuya.

    1996-01-01

    When an internal pump is started by a variable frequency power source device, if magnetic fields of an AC generator are introduced after the rated speed is reached, neutron flux high scram occurs by abrupt increase of a reactor core flow rate. Then, in the present invention, magnetic fields for the AC generator are introduced at a speed previously set at which the fluctuation range of the reactor core flow rate (neutron flux) by the start up of the internal pump is within an allowable value. Since increase of the speed of the internal pump upon its start up is suppressed to determine the change of the reactor core flow rate within an allowable range, increase of neutron fluxes is suppressed to enable stable start up. Then, since transition boiling of fuels caused by abrupt decrease of the reactor core flow rate upon occurrence of abnormality in an external electric power system is prevented, and the magnetic fields for the AC generator are introduced in such a manner to put the speed increase fluctuation range of the internal pump upon start up within an allowable value, neutron flux high scram is not caused to enable stable start-up. (N.H.)

  13. Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of neutron flux distribution and its reactivity ratio

    Energy Technology Data Exchange (ETDEWEB)

    Bitelli, Ulysses d' Utra; Aredes, Vitor O.G.; Mura, Luiz E.C.; Santos, Diogo F. dos; Silva, Alexandre P. da, E-mail: ubitelli@ipen.br, E-mail: vitoraredes@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    When compared to a rectangular parallelepiped configuration the cylindrical configuration of a nuclear reactor core has a better neutron economy because in this configuration the probability of the neutron leakage is smaller, causing an increase in overall reactivity in the system to the same amount of fuel used. In this work we obtained a critical cylindrical configuration with the control rods 89.50% withdraw from the active region of the IPEN/MB-01 core. This is the cylindrical configuration minimum possible excess of reactivity. Thus we obtained a cylindrical configuration with a diameter of only 28 fuel rods with lowest possible excess of reactivity. For this purpose, 112 peripheral fuel rods are removed from standard reactor core (rectangular parallelepiped of 28x28 fuel rods). In this configuration the excesses of reactivity is approximated 279 pcm. From there, we characterize the neutron field by measuring the spatial distribution of the thermal and epithermal neutron flux for the reactor operating power of 83 watts measured by neutron noise analysis technique and 92.08± 0.07 watts measured by activation technique [10]. The values of thermal and epithermal neutron flux in different directions, axial, radial north-south and radial east-west, are obtained in the asymptotic region of the reactor core, away from the disturbances caused by the reflector and control bar, by irradiating thin gold foils infinitely diluted (1% Au - 99% Al) with and without (bare) cadmium cover. In addition to the distribution of neutron flux, the moderator temperature coefficient, the void coefficient, calibration of the control rods were measured. (author)

  14. Unstructured Computational Aerodynamics on Many Integrated Core Architecture

    KAUST Repository

    Al Farhan, Mohammed A.

    2016-06-08

    Shared memory parallelization of the flux kernel of PETSc-FUN3D, an unstructured tetrahedral mesh Euler flow code previously studied for distributed memory and multi-core shared memory, is evaluated on up to 61 cores per node and up to 4 threads per core. We explore several thread-level optimizations to improve flux kernel performance on the state-of-the-art many integrated core (MIC) Intel processor Xeon Phi “Knights Corner,” with a focus on strong thread scaling. While the linear algebraic kernel is bottlenecked by memory bandwidth for even modest numbers of cores sharing a common memory, the flux kernel, which arises in the control volume discretization of the conservation law residuals and in the formation of the preconditioner for the Jacobian by finite-differencing the conservation law residuals, is compute-intensive and is known to exploit effectively contemporary multi-core hardware. We extend study of the performance of the flux kernel to the Xeon Phi in three thread affinity modes, namely scatter, compact, and balanced, in both offload and native mode, with and without various code optimizations to improve alignment and reduce cache coherency penalties. Relative to baseline “out-of-the-box” optimized compilation, code restructuring optimizations provide about 3.8x speedup using the offload mode and about 5x speedup using the native mode. Even with these gains for the flux kernel, with respect to execution time the MIC simply achieves par with optimized compilation on a contemporary multi-core Intel CPU, the 16-core Sandy Bridge E5 2670. Nevertheless, the optimizations employed to reduce the data motion and cache coherency protocol penalties of the MIC are expected to be of value for CFD and many other unstructured applications as many-core architecture evolves. We explore large-scale distributed-shared memory performance on the Cray XC40 supercomputer, to demonstrate that optimizations employed on Phi hybridize to this context, where each of

  15. Unstructured Computational Aerodynamics on Many Integrated Core Architecture

    KAUST Repository

    Al Farhan, Mohammed A.; Kaushik, Dinesh K.; Keyes, David E.

    2016-01-01

    Shared memory parallelization of the flux kernel of PETSc-FUN3D, an unstructured tetrahedral mesh Euler flow code previously studied for distributed memory and multi-core shared memory, is evaluated on up to 61 cores per node and up to 4 threads per core. We explore several thread-level optimizations to improve flux kernel performance on the state-of-the-art many integrated core (MIC) Intel processor Xeon Phi “Knights Corner,” with a focus on strong thread scaling. While the linear algebraic kernel is bottlenecked by memory bandwidth for even modest numbers of cores sharing a common memory, the flux kernel, which arises in the control volume discretization of the conservation law residuals and in the formation of the preconditioner for the Jacobian by finite-differencing the conservation law residuals, is compute-intensive and is known to exploit effectively contemporary multi-core hardware. We extend study of the performance of the flux kernel to the Xeon Phi in three thread affinity modes, namely scatter, compact, and balanced, in both offload and native mode, with and without various code optimizations to improve alignment and reduce cache coherency penalties. Relative to baseline “out-of-the-box” optimized compilation, code restructuring optimizations provide about 3.8x speedup using the offload mode and about 5x speedup using the native mode. Even with these gains for the flux kernel, with respect to execution time the MIC simply achieves par with optimized compilation on a contemporary multi-core Intel CPU, the 16-core Sandy Bridge E5 2670. Nevertheless, the optimizations employed to reduce the data motion and cache coherency protocol penalties of the MIC are expected to be of value for CFD and many other unstructured applications as many-core architecture evolves. We explore large-scale distributed-shared memory performance on the Cray XC40 supercomputer, to demonstrate that optimizations employed on Phi hybridize to this context, where each of

  16. Critical heat flux evaluation

    International Nuclear Information System (INIS)

    Banner, D.

    1995-01-01

    Critical heat flux (CHF) is of importance for nuclear safety and represents the major limiting factors for reactor cores. Critical heat flux is caused by a sharp reduction in the heat transfer coefficient located at the outer surface of fuel rods. Safety requires that this phenomenon also called the boiling crisis should be precluded under nominal or incidental conditions (Class I and II events). CHF evaluation in reactor cores is basically a two-step approach. Fuel assemblies are first tested in experimental loops in order to determine CHF limits under various flow conditions. Then, core thermal-hydraulic calculations are performed for safety evaluation. The paper will go into more details about the boiling crisis in order to pinpoint complexity and lack of fundamental understanding in many areas. Experimental test sections needed to collect data over wide thermal-hydraulic and geometric ranges are described CHF safety margin evaluation in reactors cores is discussed by presenting how uncertainties are mentioned. From basic considerations to current concerns, the following topics are discussed; knowledge of the boiling crisis, CHF predictors, and advances thermal-hydraulic codes. (authors). 15 refs., 4 figs

  17. Parasitic momentum flux in the tokamak core

    Science.gov (United States)

    Stoltzfus-Dueck, T.

    2017-10-01

    Tokamak plasmas rotate spontaneously without applied torque. This intrinsic rotation is important for future low-torque devices such as ITER, since rotation stabilizes certain instabilities. In the mid-radius `gradient region,' which reaches from the sawtooth inversion radius out to the pedestal top, intrinsic rotation profiles may be either flat or hollow, and can transition suddenly between these two states, an unexplained phenomenon referred to as rotation reversal. Theoretical efforts to explain the mid-radius rotation shear have largely focused on quasilinear models, in which the phase relationships of some selected instability result in a nondiffusive momentum flux (``residual stress''). In contrast, the present work demonstrates the existence of a robust, fully nonlinear symmetry-breaking momentum flux that follows from the free-energy flow in phase space and does not depend on any assumed linear eigenmode structure. The physical origin is an often-neglected portion of the radial ExB drift, which is shown to drive a symmetry-breaking outward flux of co-current momentum whenever free energy is transferred from the electrostatic potential to ion parallel flows. The fully nonlinear derivation relies only on conservation properties and symmetry, thus retaining the important contribution of damped modes. The resulting rotation peaking is counter-current and scales as temperature over plasma current. As first demonstrated by Landau, this free-energy transfer (thus also the corresponding residual stress) becomes inactive when frequencies are much higher than the ion transit frequency, which allows sudden transitions between hollow and flat profiles. Simple estimates suggest that this mechanism may be consistent with experimental observations. This work was funded in part by the Max-Planck/Princeton Center for Plasma Physics and in part by the U.S. Dept. of Energy, Office of Science, Contract No. DE-AC02-09CH11466.

  18. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    Energy Technology Data Exchange (ETDEWEB)

    Travis, Adam R [ORNL

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

  19. Oyster Creek fuel thermal margin during core thermal-hydraulic oscillations

    International Nuclear Information System (INIS)

    Dougher, J.D.

    1990-01-01

    The Oyster Creek nuclear facility, a boiling water reactor (BWR)-2 plant type, has never experienced core thermal-hydraulic instability. Power oscillations, however, have been observed in other BWR cores both domestically and internationally. Two modes of oscillations have been observed, core wide and regional half-core. During core wide oscillations, the neutron flux in the core oscillates in the radial fundamental mode. During regional half-core oscillations, higher order harmonics in the radial plane result in out-of-phase oscillations with the neutron flux in one half of the core oscillating 180 deg out-of-phase with the neutron flux in the other half of the core. General Design Criteria 12 requires either prevention or detection and suppression of power oscillations which could result in violations of fuel design limits. Analyses performed by General Electric have demonstrated that for large-magnitude oscillations the potential exists for violation of the safety limit minimum critical power ratio (MCPR). However, for plants with a flow-biased neutron flux scram automatic mitigation of oscillations may be provided at an oscillation magnitude below that at which the safety limit is challenged. Plant-specific analysis for Oyster Creek demonstrates that the existing average power range monitor (APRM) system will sense and suppress power oscillations prior to violation of any safety limits

  20. 210 Pb fluxes in sediment layers sampled from Northern Patagonia lakes

    International Nuclear Information System (INIS)

    Ribeiro Guevara, S.; Sanchez, R.; Arribere, M.; Rizzo, A.

    2003-01-01

    Unsupported 210 Pb fluxes were determined from sediment core inventories in lakes located in Northern Patagonia, Argentina. Total 210 Pb, 226 Ra, associated with supported 210 Pb, and 137 Cs specific activity profiles were measured by gamma-ray spectrometry. Unsupported 210 Pb fluxes showed very low values when compared to other regions, with a 12 fold variation, ranging from 4 to 48 Bq m -2 x y -1 . The linear correlation observed between the 210 Pb fluxes and 137 Cs cumulative fluxes in sediment cores sampled from water bodies within a zone with similar precipitation demonstrated that both radioisotopes behave in the same manner in these systems concerning the processes occurred from fallout to sediment deposition, and that there are no appreciable local or regional sources of unsupported 210 Pb. Positive correlation of 210 Pb fluxes with organic matter contents of the uppermost sediment core layers was also observed. (author)

  1. Fuel requirements for experimental devices in MTR reactors. A perturbation model for reactor core analysis

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, A.

    1991-01-01

    Irradiation in neutron absorbing devices, requiring high fast neutron fluxes in the core or high thermal fluxes in the reflector and flux traps, lead to higher density fuel and larger core dimensions. A perturbation model of the reactor core helps to estimate the fuel requirements. (orig.)

  2. A new thermal-hydraulic core module based on the drift-flux model for the DSNP

    International Nuclear Information System (INIS)

    Silverman, I.; Shapira, M.; Saphier, D.; Elias, E.

    1996-01-01

    As a part of expanding the capabilities of the reactor calculations group at Soreq - NRC a new core fuel channel module is under development. The module solves the energy equations inside the fuel rod and mass, momentum and energy equations in the coolant channel. The module uses an approximation to the drift-flux model for the solution of the coolant conditions. This module is a part of DSNP library of modules and is used in the transient simulation of nuclear power plants. Several test cases were executed simulating the AP600 PWR. Comparison of the channel model with COBRA-4I and RELAP-5 calculations have shown good agreement. It was found that the previous homogeneous equilibrium model produced adequate results for power plant simulation until boiling conditions appear in a fuel channel (authors)

  3. A new thermal-hydraulic core module based on the drift-flux model for the DSNP

    Energy Technology Data Exchange (ETDEWEB)

    Silverman, I; Shapira, M; Saphier, D [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center; Elias, E [Technion-Israel Inst. of Tech., Haifa (Israel). Dept. of Mechanical Engineering

    1996-12-01

    As a part of expanding the capabilities of the reactor calculations group at Soreq - NRC a new core fuel channel module is under development. The module solves the energy equations inside the fuel rod and mass, momentum and energy equations in the coolant channel. The module uses an approximation to the drift-flux model for the solution of the coolant conditions. This module is a part of DSNP library of modules and is used in the transient simulation of nuclear power plants. Several test cases were executed simulating the AP600 PWR. Comparison of the channel model with COBRA-4I and RELAP-5 calculations have shown good agreement. It was found that the previous homogeneous equilibrium model produced adequate results for power plant simulation until boiling conditions appear in a fuel channel (authors).

  4. Neutronic analysis of a reference LEU core for Pakistan research reactor using oxide fuel

    International Nuclear Information System (INIS)

    Akhtar, K.M.; Qazi, M.K.; Bokhari, I.H.; Khan, L.A.; Pervez, S.

    1988-07-01

    Neutronic analysis of a 10 MW reference core for PARR, having 28 fresh LEU fuel elements arranged in a 6x5 configuration has been carried out using standard computer codes WIMS-D, EXTERMINATOR-II, and CITATION. Total nuclear power peaking of 3.2 has bee found to occur in the fuel plate adjacent to the water filled central flux trap at the depth of 43.8 cm from the top of the active core. Replacement of water in central flux trap with an aluminum block, having a 50 mm diameter water filled irradiation channel changes the flux profiles in fuel, core side flux trap and reflector. The thermal flux in the central flux trap decreases by about 53%. Therefore some of the fuel elements will have to be removed and the new configuration has to be analysed to determine the first operating core. However, after achieving some burn-up and confirmation from thermal hydraulic analysis, the core configuration analysed, will be the final working core. (orig./A.B.)

  5. 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.

  6. Fast Flux Watch: A mechanism for online detection of fast flux networks

    Directory of Open Access Journals (Sweden)

    Basheer N. Al-Duwairi

    2014-07-01

    Full Text Available Fast flux networks represent a special type of botnets that are used to provide highly available web services to a backend server, which usually hosts malicious content. Detection of fast flux networks continues to be a challenging issue because of the similar behavior between these networks and other legitimate infrastructures, such as CDNs and server farms. This paper proposes Fast Flux Watch (FF-Watch, a mechanism for online detection of fast flux agents. FF-Watch is envisioned to exist as a software agent at leaf routers that connect stub networks to the Internet. The core mechanism of FF-Watch is based on the inherent feature of fast flux networks: flux agents within stub networks take the role of relaying client requests to point-of-sale websites of spam campaigns. The main idea of FF-Watch is to correlate incoming TCP connection requests to flux agents within a stub network with outgoing TCP connection requests from the same agents to the point-of-sale website. Theoretical and traffic trace driven analysis shows that the proposed mechanism can be utilized to efficiently detect fast flux agents within a stub network.

  7. Soft magnetic characteristics of laminated magnetic block cores assembled with a high Bs nanocrystalline alloy

    Directory of Open Access Journals (Sweden)

    Atsushi Yao

    2018-05-01

    Full Text Available This paper focuses on an evaluation of core losses in laminated magnetic block cores assembled with a high Bs nanocrystalline alloy in high magnetic flux density region. To discuss the soft magnetic properties of the high Bs block cores, the comparison with amorphous (SA1 block cores is also performed. In the high Bs block core, both low core losses and high saturation flux densities Bs are satisfied in the low frequency region. Furthermore, in the laminated block core made of the high Bs alloy, the rate of increase of iron losses as a function of the magnetic flux density remains small up to around 1.6 T, which cannot be realized in conventional laminated block cores based on amorphous alloy. The block core made of the high Bs alloy exhibits comparable core loss with that of amorphous alloy core in the high-frequency region. Thus, it is expected that this laminated high Bs block core can achieve low core losses and high saturation flux densities in the high-frequency region.

  8. Analysis of the Ford Nuclear Reactor LEU core

    Energy Technology Data Exchange (ETDEWEB)

    Rathkopf, J A; Drumm, C R; Martin, W R; Lee, J C [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

    1983-09-01

    This paper has summarized the current status of the effort to analyze the FNR HEU/LEU cores and to compare the calculated results with measurements. In general, calculated predictions of experimental results are quite good, especially for global parameters such as reactivity, as seen in the single HEU/LEU element substitution experiment and the LEU full core critical loading. Shim rod worths are predicted well for two of the rods but too high for a third rod possibly due to inaccurate thermal flux distribution calculation. The calculated thermal flux maps show excellent agreement with experiment throughout the FNR core. In the heavy water tank, however, experimental values for the thermal flux obtained by different methods are inconsistent among themselves as well as with the calculated finding. Work is under.way to use our computational tools to correct the discrepancies between the various measurement techniques and to improve the computational results for flux distribution and the rod worth experiment. Although uncertainties exist in our analysis, as evidenced by the discrepancies mentioned above, we consider our present calculational package to be a useful, reasonably accurate, and efficient system for performing analyses of MTR LEU/HEU core configurations.

  9. Design and analysis of a 3D-flux flux-switching permanent magnet machine with SMC cores and ferrite magnets

    Directory of Open Access Journals (Sweden)

    Chengcheng Liu

    2017-05-01

    Full Text Available Since permanent magnets (PM are stacked between the adjacent stator teeth and there are no windings or PMs on the rotor, flux-switching permanent magnet machine (FSPMM owns the merits of good flux concentrating and robust rotor structure. Compared with the traditional PM machines, FSPMM can provide higher torque density and better thermal dissipation ability. Combined with the soft magnetic composite (SMC material and ferrite magnets, this paper proposes a new 3D-flux FSPMM (3DFFSPMM. The topology and operation principle are introduced. It can be found that the designed new 3DFFSPMM has many merits over than the traditional FSPMM for it can utilize the advantages of SMC material. Moreover, the PM flux of this new motor can be regulated by using the mechanical method. 3D finite element method (FEM is used to calculate the magnetic field and parameters of the motor, such as flux density, inductance, PM flux linkage and efficiency map. The demagnetization analysis of the ferrite magnet is also addressed to ensure the safety operation of the proposed motor.

  10. 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.)

  11. Weld metal microstructures of hardfacing deposits produced by self-shielded flux-cored arc welding

    International Nuclear Information System (INIS)

    Dumovic, M.; Monaghan, B.J.; Li, H.; Norrish, J.; Dunne, D.P.

    2015-01-01

    The molten pool weld produced during self-shielded flux-cored arc welding (SSFCAW) is protected from gas porosity arising from oxygen and nitrogen by reaction ('killing') of these gases by aluminium. However, residual Al can result in mixed micro-structures of δ-ferrite, martensite and bainite in hardfacing weld metals produced by SSFCAW and therefore, microstructural control can be an issue for hardfacing weld repair. The effect of the residual Al content on weld metal micro-structure has been examined using thermodynamic modeling and dilatometric analysis. It is concluded that the typical Al content of about 1 wt% promotes δ-ferrite formation at the expense of austenite and its martensitic/bainitic product phase(s), thereby compromising the wear resistance of the hardfacing deposit. This paper also demonstrates how the development of a Schaeffler-type diagram for predicting the weld metal micro-structure can provide guidance on weld filler metal design to produce the optimum microstructure for industrial hardfacing applications.

  12. Calibration of the nuclear power channels of the IPEN/MB-01 reactor obtained from the measurements of the spatial thermal neutron flux distribution in the reactor core through the irradiation of infinitely diluted gold foils

    International Nuclear Information System (INIS)

    Goncalves, Lucas Batista

    2008-01-01

    Several nuclear parameters are obtained through the gamma spectrometry of targets irradiated in a research reactor core and this is the case of the activation foils which make possible, through the measurements of the activity induced, to determine the neutron flux in the place where they had been irradiated. The power level operation of the reactor is a parameter directly proportional to the average neutron flux in the core. This work aims to get the power operation of the reactor through of spatial neutron flux distribution in the core of IPEN/MB-01 reactor by the irradiation of infinitely diluted gold foils and prudently located in its interior. These foils were made in the form of metallic alloy in concentration levels such that the phenomena of flux disturbance, as the self-shielding factors to neutrons become worthless. These activation foils has only 1% of dispersed gold atoms in an aluminium matrix content of 99% of this element. The irradiations of foils have been carried through with and without cadmium plate. The total correlation between the average thermal neutron flux obtained by irradiation of infinitely diluted activation foils and the average digital value of current of the nuclear power channels 5 and 6 (non-compensated ionization chambers - CINC), allow the calibration of the nuclear channels of the IPEN/MB-01 reactor. (author)

  13. First in-core simultaneous measurements of nuclear heating and thermal neutron flux obtained with the innovative mobile calorimeter CALMOS inside the OSIRIS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lepeltier, Valerie; Bubendorff, Jacques; Carcreff, Hubert [Nuclear studies and reactor irradiation Service, CEA Saclay 91191 Gif sur Yvette (France); Salmon, Laurent [Thermalhydraulics and Fluid Mechanics Section, CEA Saclay 91191 Gif sur Yvette, (France)

    2015-07-01

    Nuclear heating inside a MTR reactor has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. This development required preliminary modelling and irradiation of mock-ups of the calorimetric probe in the ex-core area, where nuclear heating rate does not exceed 2 W.g{sup -1}. The calorimeter working modes, the different measurement procedures allowed with such a new probe, the main modeling and experimental results and expected advantages of this new technique have been already presented. However, these first in-core measurements were not performed beyond 6 W.g{sup -1}, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutron flux obtained by the CALMOS device at the 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutron flux up to 2.7 10{sup 14} n.cm{sup -2}.s{sup -1} and nuclear heating up to 12 W.g{sup -1}. A comprehensive measurement campaign carried out from 2013 to 2015 inside all accessible irradiation locations of the core, allowed to qualify definitively this new device, not only in terms of measurement ability but also in terms of reliability. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a systematic comparison is made between results obtained by

  14. Study on the method of determining the sub-criticality of a reactor via the measurement of core neutron flux spatial distribution

    International Nuclear Information System (INIS)

    Ma Aifeng; Jiang Xiaofeng; Zhang Shaohong

    2007-01-01

    A new methodology based on rigorous reactor physics theory astead of the point reactor assumption was proposed to determine or monitor the sub-criticality ora reactor, especially the sub-critical reactor of ADS, via the measurement of in-core flux spatial distribution. Preliminary numerical studies on the 1st ADS sub-critical experimental facilities-Venus No.1 in China have demonstrated the feasibility of this new method. Related discussions pointed out the potential applications of the method. (authors)

  15. 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

  16. Absolute measurement of neutron fluxes inside the reactor core

    International Nuclear Information System (INIS)

    Ajdacic, S. V.

    1964-10-01

    The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li 6 -semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li 6 -semiconductor spectrometer with plane geometry is given. A new type of Li 6 -semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li 6 -spectrometer made (author)

  17. Flux Pinning in Superconductors

    CERN Document Server

    Matsushita, Teruo

    2007-01-01

    The book covers the flux pinning mechanisms and properties and the electromagnetic phenomena caused by the flux pinning common for metallic, high-Tc and MgB2 superconductors. The condensation energy interaction known for normal precipitates or grain boundaries and the kinetic energy interaction proposed for artificial Nb pins in Nb-Ti, etc., are introduced for the pinning mechanism. Summation theories to derive the critical current density are discussed in detail. Irreversible magnetization and AC loss caused by the flux pinning are also discussed. The loss originally stems from the ohmic dissipation of normal electrons in the normal core driven by the electric field induced by the flux motion. The readers will learn why the resultant loss is of hysteresis type in spite of such mechanism. The influence of the flux pinning on the vortex phase diagram in high Tc superconductors is discussed, and the dependencies of the irreversibility field are also described on other quantities such as anisotropy of supercondu...

  18. Surface flux density distribution characteristics of bulk high-Tc superconductor in external magnetic field

    International Nuclear Information System (INIS)

    Torii, S.; Yuasa, K.

    2004-01-01

    Various magnetic levitation systems using oxide superconductors are developed as strong pinning forces are obtained in melt-processed bulk. However, the trapped flux of superconductor is moved by flux creep and fluctuating magnetic field. Therefore, to examine the internal condition of superconductor, the authors measure the dynamic surface flux density distribution of YBCO bulk. Flux density measurement system has a structure with the air-core coil and the Hall sensors. Ten Hall sensors are arranged in series. The YBCO bulk, which has 25 mm diameter and 13 mm thickness, is field cooled by liquid nitrogen. After that, magnetic field is changed by the air-core coil. This paper describes about the measured results of flux density distribution of YBCO bulk in the various frequencies of air-core coils currents

  19. Surface flux density distribution characteristics of bulk high- Tc superconductor in external magnetic field

    Science.gov (United States)

    Torii, S.; Yuasa, K.

    2004-10-01

    Various magnetic levitation systems using oxide superconductors are developed as strong pinning forces are obtained in melt-processed bulk. However, the trapped flux of superconductor is moved by flux creep and fluctuating magnetic field. Therefore, to examine the internal condition of superconductor, the authors measure the dynamic surface flux density distribution of YBCO bulk. Flux density measurement system has a structure with the air-core coil and the Hall sensors. Ten Hall sensors are arranged in series. The YBCO bulk, which has 25 mm diameter and 13 mm thickness, is field cooled by liquid nitrogen. After that, magnetic field is changed by the air-core coil. This paper describes about the measured results of flux density distribution of YBCO bulk in the various frequencies of air-core coils currents.

  20. Simulation experiments concerning core meltdown

    International Nuclear Information System (INIS)

    Werle, H.

    1979-01-01

    A gas stream causes a remarkable increase in the interfacial heat flux (by a factor of 8 for v = 0.63 cm/s, v = gas volume flux/horizontal area). The most important characteristics of the system investigated (silicon oil/wood metal) are relatively similar to those of a core melt, Therefore a remarkable increase of the interfacial heat transfer by the gas release may be expected also for a core melt, compared with earlier investigations at the system silicon oil/water the influence of a gas stream is nevertheless remarkably lower for silicon oil/wood metal. This shows that the density ratio plays an important role. (orig./RW) [de

  1. Observation of magnetic diffusion in the Earth's outer core from Magsat, Orsted, and CHAMP data

    DEFF Research Database (Denmark)

    Chulliat, A.; Olsen, Nils

    2010-01-01

    The frozen flux assumption consists in neglecting magnetic diffusion in the core. It has been widely used to compute core flows from geomagnetic observations. Here we investigate the validity of this assumption over the time interval 1980-2005, using high-precision magnetic data from the Magsat......, Orsted, and CHAMP satellites. A detectable change of magnetic fluxes through patches delimited by curves of zero radial magnetic field at the core-mantle boundary is associated with a failure of the frozen flux assumption. For each epoch (1980 and 2005), we calculate spatially regularized models...... of the core field which we use to investigate the change of reversed magnetic flux at the core surface. The largest and most robust change of reversed flux is observed for two patches: one located under St. Helena Island (near 20 degrees S, 15 degrees E); the other, much larger, is located under the South...

  2. Does historical wildfire activity alter metal fluxes to northern lakes?

    Science.gov (United States)

    Pelletier, N.; Chetelat, J.; Vermaire, J. C.; Palmer, M.; Black, J.; Pellisey, J.; Tracz, B.; van der Wielen, S.

    2017-12-01

    Current drought conditions in northwestern Canada are conducive to more frequent and severe wildfires that may mobilize mercury and other metals accumulated in soil and biomass. There is evidence that wildfires can remobilize and transport mercury within and outside catchments by atmospheric volatilization, particulate emissions and catchment soil erosion. However, the effect of fires on mercury fluxes to nearby lake sediments remains unclear. In this study, we use a combination of 10 dated lake sediment cores and four nearby ombrotrophic peatland cores to investigate the effects of wildfires on mercury fluxes to lake sediments. Lakes varying in catchment size and distance from recent fire events were sampled. Mercury concentrations in the environmental archives were measured, and macroscopic charcoal particles (>100 um) were counted at high resolution in the sediments to observe the co-variation of the local fire history and mercury fluxes. Mercury flux recorded in ombrotrophic peat cores provided an estimate of the historical atmospheric mercury flux from local and regional atmospheric deposition. The mercury flux recorded in lake sediments corresponds to the sum of direct atmospheric deposition and catchment transport. In combination, these archives will allow for the partitioning of mercury loading attributable to catchment transport from direct atmospheric deposition. After correcting the fluxes for particle focusing and terragenic elements input, flux from different lakes will be compared based on their catchment size and their temporal and spatial proximity known fire events. Altogether, our preliminary results using these paleolimnological methods will provide new insights on mercury transport processes that are predicted to become more important under a changing climate.

  3. Constraining Genome-Scale Models to Represent the Bow Tie Structure of Metabolism for 13C Metabolic Flux Analysis

    Directory of Open Access Journals (Sweden)

    Tyler W. H. Backman

    2018-01-01

    Full Text Available Determination of internal metabolic fluxes is crucial for fundamental and applied biology because they map how carbon and electrons flow through metabolism to enable cell function. 13 C Metabolic Flux Analysis ( 13 C MFA and Two-Scale 13 C Metabolic Flux Analysis (2S- 13 C MFA are two techniques used to determine such fluxes. Both operate on the simplifying approximation that metabolic flux from peripheral metabolism into central “core” carbon metabolism is minimal, and can be omitted when modeling isotopic labeling in core metabolism. The validity of this “two-scale” or “bow tie” approximation is supported both by the ability to accurately model experimental isotopic labeling data, and by experimentally verified metabolic engineering predictions using these methods. However, the boundaries of core metabolism that satisfy this approximation can vary across species, and across cell culture conditions. Here, we present a set of algorithms that (1 systematically calculate flux bounds for any specified “core” of a genome-scale model so as to satisfy the bow tie approximation and (2 automatically identify an updated set of core reactions that can satisfy this approximation more efficiently. First, we leverage linear programming to simultaneously identify the lowest fluxes from peripheral metabolism into core metabolism compatible with the observed growth rate and extracellular metabolite exchange fluxes. Second, we use Simulated Annealing to identify an updated set of core reactions that allow for a minimum of fluxes into core metabolism to satisfy these experimental constraints. Together, these methods accelerate and automate the identification of a biologically reasonable set of core reactions for use with 13 C MFA or 2S- 13 C MFA, as well as provide for a substantially lower set of flux bounds for fluxes into the core as compared with previous methods. We provide an open source Python implementation of these algorithms at https://github.com/JBEI/limitfluxtocore.

  4. Absolute measurement of neutron fluxes inside the reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Ajdacic, S V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1964-10-15

    The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li{sup 6}-semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li{sup 6}-semiconductor spectrometer with plane geometry is given. A new type of Li{sup 6}-semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li{sup 6}-spectrometer made (author)

  5. Core fluctuations and current profile dynamics in the MST reversed-field pinch

    International Nuclear Information System (INIS)

    Brower, D.L.; Ding, W.X.; Lei, J.

    2003-01-01

    First measurements of the current density profile, magnetic field fluctuations and electrostatic (e.s.) particle flux in the core of a high-temperature reversed-field pinch (RFP) are presented. We report three new results: (1) The current density peaks during the slow ramp phase of the sawtooth cycle and flattens promptly at the crash. Profile flattening can be linked to magnetic relaxation and the dynamo which is predicted to drive anti-parallel current in the core. Measured core magnetic fluctuations are observed to increases four-fold at the crash. Between sawtooth crashes, measurements indicate the particle flux driven by e.s. fluctuations is too small to account for the total radial particle flux. (2) Core magnetic fluctuations are observed to decrease at least twofold in plasmas where energy confinement time improves ten-fold. In this case, the radial particle flux is also reduced, suggesting core e.s. fluctuation-induced transport may play role in confinement. (3) The parallel current density increases in the outer region of the plasma during high confinement, as expected, due to the applied edge parallel electric field. However, the core current density also increases due to dynamo reduction and the emergence of runaway electrons. (author)

  6. Device for measuring neutron-flux distribution density

    International Nuclear Information System (INIS)

    Rozenbljum, N.D.; Mitelman, M.G.; Kononovich, A.A.; Kirsanov, V.S.; Zagadkin, V.A.

    1977-01-01

    An arrangement is described for measuring the distribution of neutron flux density over the height of a nuclear reactor core and which may be used for monitoring energy release or for detecting deviations of neutron flux from an optimal level so that subsequent balance can be achieved. It avoids mutual interference of detectors. Full constructional details are given. (UK)

  7. DETECTING GRAVITY MODES IN THE SOLAR {sup 8} B NEUTRINO FLUX

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Ilídio [Centro Multidisciplinar de Astrofísica, Instituto Superior Técnico, Universidade de Lisboa, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Turck-Chièze, 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)

    2014-09-10

    The detection of gravity modes produced in the solar radiative zone has been a challenge in modern astrophysics for more than 30 yr and their amplitude in the core is not yet determined. In this Letter, we develop a new strategy to look for standing gravity modes through solar neutrino fluxes. We note that due to a resonance effect, the gravity modes of low degree and low order have the largest impact on the {sup 8} B neutrino flux. The strongest effect is expected to occur for the dipole mode with radial order 2, corresponding to periods of about 1.5 hr. These standing gravity waves produce temperature fluctuations that are amplified by a factor of 170 in the boron neutrino flux for the corresponding period, in consonance with the gravity modes. From current neutrino observations, we determine that the maximum temperature variation due to the gravity modes in the Sun's core is smaller than 5.8 × 10{sup –4}. This study clearly shows that due to their high sensitivity to the temperature, the {sup 8} B neutrino flux time series is an excellent tool to determine the properties of gravity modes in the solar core. Moreover, if gravity mode footprints are discovered in the {sup 8} B neutrino flux, this opens a new line of research to probe the physics of the solar core as non-standing gravity waves of higher periods cannot be directly detected by helioseismology but could leave their signature on boron neutrino or on other neutrino fluxes.

  8. 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)

  9. Operational characteristics of hybrid-type SFCL with closed and open cores

    International Nuclear Information System (INIS)

    Cho, Y.S.; Lee, N.Y.; Choi, H.S.; Chung, D.C.; Lim, S.H.

    2007-01-01

    We investigated the operational characteristics of the hybrid-type superconducting fault current limiter (SFCL) with the closed and the open cores, which induced the variation of the magnetic flux between the primary and the secondary windings. The experimental set-up of the hybrid-type SFCL with the closed and the open cores were prepared and the experimental analyses for the current limiting characteristics were performed. The peak value of the fault current in the hybrid-type SFCL with the open core was higher than that of the closed core at the first cycle after fault occurrence. However, in the case of the hybrid-type SFCL with the open core, the limiting current level after fault occurrence was decreased less than that of the hybrid-type SFCL with the closed core, because the magnetic leakage reluctance of the open core was higher than that of the closed core. The quench time (T q ) and the arrival time (T a ) for the peak voltage (V SC ) in the hybrid-type SFCL with the closed core were faster than that of the hybrid-type SFCL with the open core due to the increase of the mutual flux. We verified that the consumption power in the hybrid-type SFCL with the open core was larger owing to the increase of leakage flux by the reduction of mutual inductance between primary and secondary windings

  10. FFTF [Fast Flux Test Facility] management

    International Nuclear Information System (INIS)

    Bennett, C.L.

    1986-11-01

    Fuel Management at the Fast Flux Test Facility (FFTF) involves more than just the usual ex-core and in-core management of standard fuel and non-fuel components between storage locations and within the core since it is primarily an irradiation test facility. This mission involves testing an ever increasing variety of fueled and non-fueled experiments, each having unique requirements on the reactor core as well as having its own individual impact on the reload design. This paper describes the fuel management process used by the Westinghouse Hanford Company Core Engineering group that has led to the successful reload design of nine operating cycles and the irradiation of over 120 tests

  11. A platinum in-core flux detector

    International Nuclear Information System (INIS)

    Shields, R.B.

    1976-01-01

    The performance is described of a platinum emitter self-powered detector having the following parameters: emitter diameter 0.51 mm, Inconel 600 collector of 1.5 mm outer diameter and 0.25 mm wall thickness, compacted powder MgO insulant, thermal neutron flux 10 14 n.cm -2 .s -1 and gamma radiation dose rate 1.2 x 10 8 rad.h -1 . The advantage of the detector is its sensitivity to both neutrons and gamma radiation. A comparison is made with other types of detectors using Ce, Ta, Os, Rh, V, Co, Zr as emitters, especially in relation to the emitter response time to neutrons or gammas, the output signal amplitude, sensitivity, and the emitter half-life. Extensive tests of the detectors proceeded for two years on the NRU and CANDU-BLW reactors in Gentilly, Canada. (J.B.)

  12. Concerns about the dynamic responses of in-core flux detectors

    Energy Technology Data Exchange (ETDEWEB)

    Cuttler, J.M., E-mail: jerrycuttler@rogers.com [Cuttler & Associates Inc., Mississauga, Ontario (Canada); Gill, H.; Scrannage, R.; Paquette, P., E-mail: jerrycuttler@rogers.com [Bruce Power, Tiverton, Ontario (Canada)

    2012-07-01

    CANDUs are determining the dynamic responses of flux detectors by a method open to question. It ignores relative changes in local flux conditions, which are significant during trips. Calculated prompt fractions (PFs) are widespread. The SIR detector development calculated the PF change with irradiation on a physical basis. Measurements were made over many years. The current results do not agree with the 1996 predictions. Some values are below the safety analysis limit. This has resulted in detector replacement, imposition of CPPF penalties on trip margins, additional safety analyses and other actions. This paper shows that such measurements are not required. (author)

  13. Sub-core permeability and relative permeability characterization with Positron Emission Tomography

    Science.gov (United States)

    Zahasky, C.; Benson, S. M.

    2017-12-01

    This study utilizes preclinical micro-Positron Emission Tomography (PET) to image and quantify the transport behavior of pulses of a conservative aqueous radiotracer injected during single and multiphase flow experiments in a Berea sandstone core with axial parallel bedding heterogeneity. The core is discretized into streamtubes, and using the micro-PET data, expressions are derived from spatial moment analysis for calculating sub-core scale tracer flux and pore water velocity. Using the flux and velocity data, it is then possible to calculate porosity and saturation from volumetric flux balance, and calculate permeability and water relative permeability from Darcy's law. Full 3D simulations are then constructed based on this core characterization. Simulation results are compared with experimental results in order to test the assumptions of the simple streamtube model. Errors and limitations of this analysis will be discussed. These new methods of imaging and sub-core permeability and relative permeability measurements enable experimental quantification of transport behavior across scales.

  14. Reluctance motor employing superconducting magnetic flux switches

    International Nuclear Information System (INIS)

    Spyker, R.L.; Ruckstadter, E.J.

    1992-01-01

    This paper reports that superconducting flux switches controlling the magnetic flux in the poles of a motor will enable the implementation of a reluctance motor using one central single phase winding. A superconducting flux switch consists of a ring of superconducting material surrounding a ferromagnetic pole of the motor. When in the superconducting state the switch will block all magnetic flux attempting to flow in the ferromagnetic core. When switched to the normal state the superconducting switch will allow the magnetic flux to flow freely in that pole. By using one high turns-count coil as a flux generator, and selectively channeling flux among the various poles using the superconducting flux switch, 3-phase operation can be emulated with a single-hase central AC source. The motor will also operate when the flux generating coil is driven by a DC current, provided the magnetic flux switches see a continuously varying magnetic flux. Rotor rotation provides this varying flux due to the change in stator pole inductance it produces

  15. EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

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

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Wang, Chenglong; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

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

  17. Transient neutrons flux behaviour in a spherical reactor core

    International Nuclear Information System (INIS)

    Souza, A.W.A. de.

    1978-11-01

    This work studies the transient neutron flux in a fast reactor of spherical geometry. The burning of U 235 nuclei is equated and two kinds of reflector were studied. The numeric solutions are then compared with the results for those reflectors. (author) [pt

  18. 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.)

  19. A neutronic feasibility study for LEU conversion of the High Flux Beam Reactor (HFBR)

    International Nuclear Information System (INIS)

    Pond, R.B.; Hanan, N.A.; Matos, J.E.

    1997-01-01

    A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm 3 in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept. (author)

  20. Surface flux density distribution characteristics of bulk high-T{sub c} superconductor in external magnetic field

    Energy Technology Data Exchange (ETDEWEB)

    Torii, S.; Yuasa, K

    2004-10-01

    Various magnetic levitation systems using oxide superconductors are developed as strong pinning forces are obtained in melt-processed bulk. However, the trapped flux of superconductor is moved by flux creep and fluctuating magnetic field. Therefore, to examine the internal condition of superconductor, the authors measure the dynamic surface flux density distribution of YBCO bulk. Flux density measurement system has a structure with the air-core coil and the Hall sensors. Ten Hall sensors are arranged in series. The YBCO bulk, which has 25 mm diameter and 13 mm thickness, is field cooled by liquid nitrogen. After that, magnetic field is changed by the air-core coil. This paper describes about the measured results of flux density distribution of YBCO bulk in the various frequencies of air-core coils currents.

  1. Analysis of neutron flux increase in the horizontal experimental channels of Ra reactor - masters thesis

    International Nuclear Information System (INIS)

    Strugar, P.

    1964-12-01

    Calculation and experimental results shown in this paper show that higher thermal neutron flux is obtained in the reactor core with central horizontal reflector at the same power level. The flux is increased when the moderation capability of the core is decreased. Apart from increase of the thermal component of the neutron flux in the experimental channels, the central reflector causes decrease of the epithermal neutron flux and gamma radiation intensity. This is very useful for studying (n, γ) reaction, neutron diffraction, etc. [sr

  2. Formation and sustainment of field reversed configuration (FRC) plasmas by spheromak merging and neutral beam injection

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, Masaaki [Princeton Plasma Physics Laboratory, Princeton University Princeton, New Jersey USA (United States)

    2016-03-25

    This paper briefly reviews a compact toroid reactor concept that addresses critical issues for forming, stabilizing and sustaining a field reversed configuration (FRC) with the use of plasma merging, plasma shaping, conducting shells, neutral beam injection (NBI). In this concept, an FRC plasma is generated by the merging of counter-helicity spheromaks produced by inductive discharges and sustained by the use of neutral beam injection (NBI). Plasma shaping, conducting shells, and the NBI would provide stabilization to global MHD modes. Although a specific FRC reactor design is outside the scope of the present paper, an example of a promising FRC reactor program is summarized based on the previously developed SPIRIT (Self-organized Plasmas by Induction, Reconnection and Injection Techniques) concept in order to connect this concept to the recently achieved the High Performance FRC plasmas obtained by Tri Alpha Energy [Binderbauer et al, Phys. Plasmas 22,056110, (2015)]. This paper includes a brief summary of the previous concept paper by M. Yamada et al, Plasma Fusion Res. 2, 004 (2007) and the recent experimental results from MRX.

  3. A new self-powered flux detector

    International Nuclear Information System (INIS)

    Allan, C.J.

    1979-11-01

    It has been found that an Inconel-Inconel coaxial cable can be used as a fast-responding, neutron, self-powered flux detector if the core wire is sufficiently large. Test results obtained with such a detector, having a core wire approximately 1.5 mm in diameter, are presented. Other materials suitable for use as an emitter material, in such a relatively large diameter detector, also are included. (auth)

  4. 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.

  5. 1500-year Record of trans-Pacific Dust Flux collected from the Denali Ice Core, Mt. Hunter, Alaska

    Science.gov (United States)

    Saylor, P. L.; Osterberg, E. C.; Koffman, B. G.; Winski, D.; Ferris, D. G.; Kreutz, K. J.; Wake, C. P.; Handley, M.; Campbell, S. W.

    2016-12-01

    Mineral dust aerosols are a critical component of the climate system through their influence on atmospheric radiative forcing, ocean productivity, and surface albedo. Dust aerosols derived from Asian deserts are known to reach as far as Europe through efficient transport in the upper tropospheric westerlies. While centennially-to-millennially resolved Asian dust records exist over the late Holocene from North Pacific marine sediment cores and Asian loess deposits, a high-resolution (sub-annual to decadal) record of trans-Pacific dust flux will significantly improve our understanding of North Pacific dust-climate interactions and provide paleoclimatological context for 20th century dust activity. Here we present an annually resolved 1500-year record of trans-Pacific dust transport based on chemical and physical dust measurements in parallel Alaskan ice cores (208 m to bedrock) collected from the summit plateau of Mt. Hunter in Denali National Park. The cores were sampled at high resolution using a continuous melter system with discrete analyses for major ions (Dionex ion chromatograph), trace elements (Element2 inductively coupled plasma mass spectrometer), and stable water isotope ratios (Picarro laser ringdown spectroscopy), and continuous flow analysis for dust concentration and size distribution (Klotz Abakus). We compare the ice core dust record to instrumental aerosol stations, satellite observations, and dust model data from the instrumental period, and evaluate climatic controls on dust emission and trans-Pacific transport using climate reanalysis data, to inform dust-climate relationships over the past 1500 years. Physical particulate and chemical data demonstrate remarkable fidelity at sub-annual resolution, with both displaying a strong springtime peak consistent with periods of high dust activity over Asian desert source regions. Preliminary results suggest volumetric mode typically ranges from 4.5 - 6.5 um, with a mean value of 5.5 um. Preliminary

  6. 3D computer visualization and animation of CANDU reactor core

    International Nuclear Information System (INIS)

    Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

    1999-01-01

    Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

  7. Research on plasma core reactors

    International Nuclear Information System (INIS)

    Jarvis, G.A.; Barton, D.M.; Helmick, H.H.; Bernard, W.; White, R.H.

    1977-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with 1-m-diam by 1-m-long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF 6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000-cm 3 aluminum canister in the central region was fueled with UF 6 gas and fission density distributions determined. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation

  8. Improvements in EBR-2 core depletion calculations

    International Nuclear Information System (INIS)

    Finck, P.J.; Hill, R.N.; Sakamoto, S.

    1991-01-01

    The need for accurate core depletion calculations in Experimental Breeder Reactor No. 2 (EBR-2) is discussed. Because of the unique physics characteristics of EBR-2, it is difficult to obtain accurate and computationally efficient multigroup flux predictions. This paper describes the effect of various conventional and higher order schemes for group constant generation and for flux computations; results indicate that higher-order methods are required, particularly in the outer regions (i.e. the radial blanket). A methodology based on Nodal Equivalence Theory (N.E.T.) is developed which allows retention of the accuracy of a higher order solution with the computational efficiency of a few group nodal diffusion solution. The application of this methodology to three-dimensional EBR-2 flux predictions is demonstrated; this improved methodology allows accurate core depletion calculations at reasonable cost. 13 refs., 4 figs., 3 tabs

  9. PAH fluxes in the Laja Lake of south central Chile Andes over the last 50 years: Evidence from a dated sediment core

    International Nuclear Information System (INIS)

    Quiroz, Roberto; Popp, Peter; Urrutia, Roberto; Bauer, Coretta; Araneda, Alberto; Treutler, Hanns-Christian; Barra, Ricardo

    2005-01-01

    This paper reports the occurrence of polyaromatic hydrocarbons (PAHs) deposition inferred from a sediment core of an Andean lake in south central Chile. Sediments were carefully collected from one of the deepest section of the lake and sliced every 1 cm. The samples were analyzed for PAHs, 137 Cs, 210 Pb, organic carbon and grain-size. The stratigraphic chronology and the sedimentation rates were estimated using the sedimentary signature left by the 137 Cs and 210 Pb fallout as temporal markers. PAHs were quantified by HPLC-fluorescence detection (HPLC-Fluorescence). 15 priority EPA PAHs were analyzed in this study. Based on these results, PAH deposition over the last 50 years was estimated (a period characterized by an important intervention in the area). PAH concentration ranged from 226 to 620 ng g -1 d.w. The highest concentrations of PAHs were found in the core's bottom. The PAH profile is dominated by the presence of perylene indicating a natural source of PAH. In addition, two clear PAH deposition periods could be determined: the most recent with two-four rings PAHs, the older one with five-seven rings predomination. Determined fluxes where 71 to 972 μg m -2 year -1 , dominated by perylene deposition. PAH levels and fluxes are lower compared to the levels found in sediments from remote lakes in Europe and North America. It is concluded that the main source of PAHs into the Laja Lake sediments are of natural origin

  10. The thermal evolution of Mercury's Fe-Si core

    Science.gov (United States)

    Knibbe, Jurriën Sebastiaan; van Westrenen, Wim

    2018-01-01

    We have studied the thermal and magnetic field evolution of planet Mercury with a core of Fe-Si alloy to assess whether an Fe-Si core matches its present-day partially molten state, Mercury's magnetic field strength, and the observed ancient crustal magnetization. The main advantages of an Fe-Si core, opposed to a previously assumed Fe-S core, are that a Si-bearing core is consistent with the highly reduced nature of Mercury and that no compositional convection is generated upon core solidification, in agreement with magnetic field indications of a stable layer at the top of Mercury's core. This study also present the first implementation of a conductive temperature profile in the core where heat fluxes are sub-adiabatic in a global thermal evolution model. We show that heat migrates from the deep core to the outer part of the core as soon as heat fluxes at the outer core become sub-adiabatic. As a result, the deep core cools throughout Mercury's evolution independent of the temperature evolution at the core-mantle boundary, causing an early start of inner core solidification and magnetic field generation. The conductive layer at the outer core suppresses the rate of core growth after temperature differences between the deep and shallow core are relaxed, such that a magnetic field can be generated until the present. Also, the outer core and mantle operate at higher temperatures than previously thought, which prolongs mantle melting and mantle convection. The results indicate that S is not a necessary ingredient of Mercury's core, bringing bulk compositional models of Mercury more in line with reduced meteorite analogues.

  11. Determination flux in the Reactor JEN-1

    International Nuclear Information System (INIS)

    Manas Diaz, L.; Montes Ponce de leon, J.

    1960-01-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs

  12. NIMROD simulations and physics assessment of possible designs for a next generation Steady Inductive Helicity Injection HIT device

    Science.gov (United States)

    Penna, James; Morgan, Kyle; Grubb, Isaac; Jarboe, Thomas

    2017-10-01

    The Helicity Injected Torus - Steady Inductive 3 (HIT-SI3) experiment forms and maintains spheromaks via Steady Inductive Helicity Injection (SIHI) using discrete injectors that inject magnetic helicity via a non-axisymmetric perturbation and drive toroidally symmetric current. Newer designs for larger SIHI-driven spheromaks incorporate a set of injectors connected to a single external manifold to allow more freedom for the toroidal structure of the applied perturbation. Simulations have been carried out using the NIMROD code to assess the effectiveness of various imposed mode structures and injector schema in driving current via Imposed Dynamo Current Drive (IDCD). The results are presented here for varying flux conserver shapes on a device approximately 1.5 times larger than the current HIT-SI3 experiment. The imposed mode structures and spectra of simulated spheromaks are analyzed in order to examine magnetic structure and stability and determine an optimal regime for IDCD sustainment in a large device. The development of scaling laws for manifold operation is also presented, and simulation results are analyzed and assessed as part of the development path for the large scale device.

  13. Contribution to the qualification of calculation methods of reactivity and of flux and power distributions in nuclear pressurized water reactor cores

    International Nuclear Information System (INIS)

    Abit, K.

    1984-01-01

    The last stage of the creation computer methods and calculations consists of verifying the running and qualifying the results obtained. The work of the present thesis consisted of improving a coupling method between radial and axial phenomena in a PWR core, refering to three-dimensional calculations, while ensuring a perfect coherence between the programmed physical models. The calculation results have been compared to measurements of reactivity and of flux distributions realized during start-up tests. Thus, the methods have been applied to the calculation of the evolution of a burnable poison (gadolinium) in view of operation in long campaign. 13 refs [fr

  14. The Solar-flux Third Granulation Signature

    Science.gov (United States)

    Gray, David F.; Oostra, Benjamin

    2018-01-01

    The velocity shifts of spectral lines as a function of line strength, so-called the third signature of granulation, are investigated using three published solar-flux atlases. We use flux atlases because we wish to treat the Sun as a star, against which stellar observations can be compared and judged. The atlases are critiqued and compared to the lower-resolution observations taken with the Elginfield stellar spectrograph. Third-signature plots are constructed for the 6020–6340 Å region. No dependence on excitation potential or wavelength is found over this wavelength span. The shape of the plots from the three solar atlases is essentially the same, with rms line-core velocity differences of 30–35 m s‑1. High-resolution atlas data are degraded to the level of the Elginfield spectrograph and compared to direct observations taken with that spectrograph. The line-core velocities show good agreement, with rms differences of 38 m s‑1. A new standard curve is derived and compared with the previously published one. Only small differences in shape are found, but a significant (+97 m s‑1) change in the zero point is indicated. The bisector of the Fe I 6253 line is mapped onto the third-signature plots and flux deficits are derived, which measure the granule/lane flux imbalance. The lower spectral resolution lowers the flux deficit area slightly and moves the peak of the deficit 0.3–0.5 km s‑1 toward higher velocities. These differences, while significant, are not large compared to measurement errors for stellar data.

  15. Evaluation of in-core neutron flux and temperature field measurements during the second period of power commissioning of the KS-150 reactor

    International Nuclear Information System (INIS)

    Rana, S.B.; Pecho, J.

    1975-01-01

    The in-core flux mapping system in the KS-150 reactor using mapping fuel elements with self-powered detectors is described. Experimental data evaluation using the Fourier analysis and determination of important operation parameters from the detectors and temperature field distribution using thermocouples for measuring coolant outlet temperatures and fuel temperatures are given. The DPZ-1 detectors used, mapping fuel elements and the method of signal registration are described. The results of operation of mapping fuel elements during the 2nd period of the KS-150 reactor commissioning are given. (author)

  16. Evaluation of a novel noninvasive continuous core temperature measurement system with a zero heat flux sensor using a manikin of the human body.

    Science.gov (United States)

    Brandes, Ivo F; Perl, Thorsten; Bauer, Martin; Bräuer, Anselm

    2015-02-01

    Reliable continuous perioperative core temperature measurement is of major importance. The pulmonary artery catheter is currently the gold standard for measuring core temperature but is invasive and expensive. Using a manikin, we evaluated the new, noninvasive SpotOn™ temperature monitoring system (SOT). With a sensor placed on the lateral forehead, SOT uses zero heat flux technology to noninvasively measure core temperature; and because the forehead is devoid of thermoregulatory arteriovenous shunts, a piece of bone cement served as a model of the frontal bone in this study. Bias, limits of agreements, long-term measurement stability, and the lowest measurable temperature of the device were investigated. Bias and limits of agreement of the temperature data of two SOTs and of the thermistor placed on the manikin's surface were calculated. Measurements obtained from SOTs were similar to thermistor values. The bias and limits of agreement lay within a predefined clinically acceptable range. Repeat measurements differed only slightly, and stayed stable for hours. Because of its temperature range, the SOT cannot be used to monitor temperatures below 28°C. In conclusion, the new SOT could provide a reliable, less invasive and cheaper alternative for measuring perioperative core temperature in routine clinical practice. Further clinical trials are needed to evaluate these results.

  17. Development of conceptual nuclear design of 10MWt research reactor core

    International Nuclear Information System (INIS)

    Kim, M. H.; Lim, J. Y.; Win, Naing; Park, J. M.

    2008-03-01

    KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

  18. 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)

  19. The High Flux Beam Reactor at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Shapiro, S.M.

    1994-01-01

    Brookhaven National Laboratory's High Flux Beam Reactor (HFBR) was built because of the need of the scientist to always want 'more'. In the mid-50's the Brookhaven Graphite reactor was churning away producing a number of new results when the current generation of scientists, led by Donald Hughes, realized the need for a high flux reactor and started down the political, scientific and engineering path that led to the BFBR. The effort was joined by a number of engineers and scientists among them, Chemick, Hastings, Kouts, and Hendrie, who came up with the novel design of the HFBR. The two innovative features that have been incorporated in nearly all other research reactors built since are: (i) an under moderated core arrangement which enables the thermal flux to peak outside the core region where beam tubes can be placed, and (ii) beam tubes that are tangential to the core which decrease the fast neutron background without affecting the thermal beam intensity. Construction began in the fall of 1961 and four years later, at a cost of $12 Million, criticality was achieved on Halloween Night, 1965. Thus began 30 years of scientific accomplishments

  20. Evaluation of thermal margin during BWR neutron flux oscillation

    International Nuclear Information System (INIS)

    Takeuchi, Yutaka; Takigawa, Yukio; Chuman, Kazuto; Ebata, Shigeo

    1992-01-01

    Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three-dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio). The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit. A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide -oscillation, because of the sensitivity difference in a flow-to-power gain. A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals. (author)

  1. Flux Density through Guides with Microstructured Twisted Clad DB Medium

    Directory of Open Access Journals (Sweden)

    M. A. Baqir

    2014-01-01

    Full Text Available The paper deals with the study of flux density through a newly proposed twisted clad guide containing DB medium. The inner core and the outer clad sections are usual dielectrics, and the introduced twisted windings at the core-clad interface are treated under DB boundary conditions. The pitch angle of twist is supposed to greatly contribute towards the control over the dispersion characteristics of the guide. The eigenvalue equation for the guiding structure is deduced, and the analytical investigations are made to explore the propagation patterns of flux densities corresponding to the sustained low-order hybrid modes under the situation of varying pitch angles. The emphasis has been put on the effects due to the DB twisted pitch on the propagation of energy flux density through the guide.

  2. Noise analysis based validation of the dynamics of in-core flux detectors and ion chambers used in SDS and RRS systems

    International Nuclear Information System (INIS)

    Gloeckler, O.; Cooke, D.; Tulett, M.V.

    1996-01-01

    The paper concentrates on some of the recent applications of reactor noise analysis in Ontario Hydro's CANDU stations, related to the dynamics of in-core flux detectors (ICFDs) and ion chambers. These applications include (1) detecting anomalies in the dynamics of ICFDs and ion chambers, (2) estimating the effective prompt fractions of ICFDs in power rundown tests and in noise measurement, (3) detecting the mechanical vibration of ICFD instrument tubes induced by moderator flow, (4) detecting the mechanical vibration of fuel channels induced by coolant flow, (5) identifying the cause of excessive signal fluctuations in certain flux detectors, (6) validating the dynamic coupling between liquid zone control signals. Some of these applications are performed on a regular basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the AECB with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the use of noise analysis technology (author)

  3. In-core fuel management practice in HANARO

    International Nuclear Information System (INIS)

    Kim Hark Rho; Lee Choong Sung; Lee Jo Bok

    1997-01-01

    KAERI (KOREA Atomic Energy Research Institute) completed the system performance tests for the HANARO (Hi-flux Advanced Neutron Application Research Reactor) on December 1994. Its initial criticality was achieved on February 8, 1995. A variety of the reactor physics experiments were performed in parallel with configuring the first cycle core and now HANARO is in the third cycle operation. The in-core fuel management in HANARO is performed on the following strategy: 1) the cycle length of the equilibrium core is at least 4 week FPDs, 2) the maximum linear heat generation rate should be within the design limit, 3) the reactor should have shutdown margin of 1% Δk/k at minimum, 4) the available thermal flux should satisfy the users' requirements. This paper presents the fuel management practice in HANARO. Section II briefly describes the design feature of the HANARO and the method of analysis follows in section III and section IV describes In-core fuel management practice and the conclusion is remarked in the final section. (author)

  4. Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of spatial and energetic distribution of neutron flux distribution; Caracterizacao do nucleo cilindrico de menor excesso de reatividade do reator IPEN/MB-01, pela medida da distribuicao espacial e energetica do fluxo de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Aredes, Vitor Ottoni Garcia

    2014-07-01

    In this work was conducted the mapping of the thermal and epithermal neutrons flux and the energy spectrum of the neutrons in the reactor core IPEN/MB-01 for a cylindrical core configuration with minor excess reactivity, which is 28 x 28 fuel rods arranged in north-south and east-west directions. The calibration of control rods for this configuration determined their excess reactivity. The lower excess reactivity in the core decreased neutron flux disturbance caused by the neutron absorbing rods , given that the nuclear reactor was operated with the rods almost completely removed . Was used the 'Activation Analysis Technique' with the thin foil activation detectors ( infinitely diluted and hyper-pure), of different materials that work in different energy ranges, to calculate the saturation activity, used for determining the neutron flux and in the SANDBP code as input for the calculation of the neutrons energy spectrum. To discriminate thermal and epithermal flux , was used the 'Cadmium RatioTechnique' . The activation detectors were distributed in a total of 140 radial and axial positions in the reactor core and 16 irradiation, with bare and covered with cadmium activation foils. A model of this configuration was simulated by MCNP-5 code to determine the cadmium correction factor and comparison of the results obtained experimentally. The cylindrical configuration desired, with 17% less fuel than the standard rectangular configuration (28 x 26 fuel rods), reached criticality with the control rods approximately 90% removed, which decreased considerably the disturbance in neutron flux. Given the highest power density of the 28 x 28 cylindrical core, the neutron flux increased by over 50% in the central regions of the core compared to the values of the 28 x 26 standard rectangular core. (author)

  5. Annually resolved southern hemisphere volcanic history from two Antarctic ice cores

    Science.gov (United States)

    Cole-Dai, Jihong; Mosley-Thompson, Ellen; Thompson, Lonnie G.

    1997-07-01

    The continuous sulfate analysis of two Antarctic ice cores, one from the Antarctic Peninsula region and one from West Antarctica, provides an annually resolved proxy history of southern semisphere volcanism since early in the 15th century. The dating is accurate within ±3 years due to the high rate of snow accumulation at both core sites and the small sample sizes used for analysis. The two sulfate records are consistent with each other. A systematic and objective method of separating outstanding sulfate events from the background sulfate flux is proposed and used to identify all volcanic signals. The resulting volcanic chronology covering 1417-1989 A.D. resolves temporal ambiguities about several recently discovered events. A number of previously unknown, moderate eruptions during late 1600s are uncovered in this chronology. The eruption of Tambora (1815) and the recently discovered eruption of Kuwae (1453) in the tropical South Pacific injected the greatest amount of sulfur dioxide into the southern hemisphere stratosphere during the last half millennium. A technique for comparing the magnitude of volcanic events preserved within different ice cores is developed using normalized sulfate flux. For the same eruptions the variability of the volcanic sulfate flux between the cores is within ±20% of the sulfate flux from the Tambora eruption.

  6. Multilevel power distribution synthesis for a movable flux mapping system

    International Nuclear Information System (INIS)

    Bollacasa, D.; Terney, W.B.; Vincent, G.F.; Dziadosz, D.; Schleicher, T.

    1992-01-01

    A Computer Software package has been developed to support the synthesis of the 3-dimensional power distribution from detector signals from a movable flux mapping system. The power distribution synthesis is based on methodology developed for fixed incore detectors. The full core solution effectively couples all assemblies in the core whether they are instrumented or not. The solution is not subject to approximations for the treatment of assemblies where a measurement cannot be made and provides an accurate representation of axial variations which may be induced by axial blankets, burnable absorber cut back regions and axially zoned flux suppression rods

  7. Method and apparatus for monitoring the axial power distribution within the core of a nuclear reactor, exterior of the reactor

    International Nuclear Information System (INIS)

    Graham, K.F.; Gopal, R.

    1978-01-01

    A method and apparatus for establishing the axial flux distribution of a reactor core from monitored responses obtained exterior of the reactor is described. The monitored responses are obtained from at least three axially spaced flux responsive detectors that are positioned within proximity of the periphery of the reactor core. The detectors provide corresponding electrical outputs representative of the flux monitored. The axial height of the core is figuratively divided at a plurality of space coordinates sufficient to provide reconstruction in point representation of the relative flux shape along the core axis. The relative value of flux at each of the spaced coordinates is then established from a sum of the electrical outputs of the detectors, respectively, algebraically modified by a corresponding preestablished constant

  8. Improved HOR fuel management by flux measurement data feedback

    Energy Technology Data Exchange (ETDEWEB)

    Serov, I.V.; Leege, P.F.A. de; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft University of Technology, Reactor Physics Dep., Interfaculty Reactor Inst., Delft (Netherlands)

    1997-07-01

    Flux distribution in a nuclear reactor can be obtained by utilizing different calculational and experimental methods. The obtained flux distributions are associated with uncertainties and therefore always differ from each other. By combining information from the calculation and experiment using the confluence method, it is possible to obtain a more reliable estimate of the flux distribution than exhibited by the calculation or experiment separately. As a feedback, the fuel burnup distribution, which is used as initial data to the calculation can be improved as well. The confluence method is applied to improvement of the burnup distribution estimates for the HOR research reactor of the Delft University of Technology. An integrated code system CONHOR is developed to match the CITATION results of in-core foil activation rate calculations with in-core experimental data through confluence. The system forms the basis for the advanced fuel management of the reactor. (author)

  9. Improved HOR fuel management by flux measurement data feedback

    International Nuclear Information System (INIS)

    Serov, I.V.; Leege, P.F.A. de; Hoogenboom, J.E.; Gibcus, H.P.M.

    1997-01-01

    Flux distribution in a nuclear reactor can be obtained by utilizing different calculational and experimental methods. The obtained flux distributions are associated with uncertainties and therefore always differ from each other. By combining information from the calculation and experiment using the confluence method, it is possible to obtain a more reliable estimate of the flux distribution than exhibited by the calculation or experiment separately. As a feedback, the fuel burnup distribution, which is used as initial data to the calculation can be improved as well. The confluence method is applied to improvement of the burnup distribution estimates for the HOR research reactor of the Delft University of Technology. An integrated code system CONHOR is developed to match the CITATION results of in-core foil activation rate calculations with in-core experimental data through confluence. The system forms the basis for the advanced fuel management of the reactor. (author)

  10. 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.

  11. Introducing FACETS, the Framework Application for Core-Edge Transport Simulations

    International Nuclear Information System (INIS)

    Cary, John R.; Candy, Jeff; Cohen, Ronald H.; Krasheninnikov, Sergei I.; McCune, Douglas C.; Estep, Donald J.; Larson, Jay W.; Malony, Allen; Worley, Patrick H.; Carlsson, Johann Anders; Hakim, A.H.; Hamill, P.; Kruger, Scott E.; Muzsala, S.; Pletzer, Alexander; Shasharina, Svetlana; Wade-Stein, D.; Wang, N.; McInnes, Lois C.; Wildey, T.; Casper, T.A.; Diachin, Lori A.; Epperly, Thomas; Rognlien, T.D.; Fahey, Mark R.; Kuehn, Jeffery A.; Morris, A.; Shende, Sameer; Feibush, E.; Hammett, Gregory W.; Indireshkumar, K.; Ludescher, C.; Randerson, L.; Stotler, D.; Pigarov, A.; Bonoli, P.; Chang, C.S.; D'Ippolito, D.A.; Colella, Philip; Keyes, David E.; Bramley, R.

    2007-01-01

    The FACETS (Framework Application for Core-Edge Transport Simulations) project began in January 2007 with the goal of providing core to wall transport modeling of a tokamak fusion reactor. This involves coupling previously separate computations for the core, edge, and wall regions. Such a coupling is primarily through connection regions of lower dimensionality. The project has started developing a component-based coupling framework to bring together models for each of these regions. In the first year, the core model will be a 1 dimensional model (1D transport across flux surfaces coupled to a 2D equilibrium) with fixed equilibrium. The initial edge model will be the fluid model, UEDGE, but inclusion of kinetic models is planned for the out years. The project also has an embedded Scientific Application Partnership that is examining embedding a full-scale turbulence model for obtaining the crosssurface fluxes into a core transport code.

  12. Development of Structural Core Components for Breeder Reactors

    International Nuclear Information System (INIS)

    Saibaba, N.

    2013-01-01

    Core structural materials: • The desire is to have only fuel in the core, structural material form 25% of the total core: – To support and to retain the fuel in position; – Provide necessary ducts to make coolant flow through & transfer/remove heat. • For 500 MWe FBR with Oxide fuel (Peak Linear Power 450 W/cm), total fuel pins required in the core are of the order 39277 pins (both inner & outer core Fuel SA); • Considering 217 pins/Fuel SA there are 181 Fuel SA wrapper tubes • These structural materials see hostile core with max temperature and neutron flux

  13. Evolution of Flux Mapping System (FMS) from 540 MWe to 700 MWe Indian PHWR: design perspective

    International Nuclear Information System (INIS)

    Sonavani, Manojkumar; Kelkar, M.G.; Singhvi, P.K.; Roy, S.; Ingle, V.J.

    2013-01-01

    The Flux Mapping System (FMS) of 700 MWe PHWR computes a detailed flux/power distribution of the reactor core using modal synthesis method and is also generate setback on different parameters by monitoring thermal neutron flux at more than 100 points inside the reactor core. These types of setbacks are introduced first time in Indian PHWRs. The paper brings out the Evolution of Flux Mapping System (FMS) from 540 MWe to 700 MWe and the overall design philosophy. The paper emphasizes on comparisons between 540 MWe and 700 MWe design, considerations for architectural design and setbacks for 700 MWe. (author)

  14. Optimization of the Magnetic Field Structure for Sustained Plasma Gun Helicity Injection for Magnetic Turbulence Studies at the Bryn Mawr Plasma Laboratory

    Science.gov (United States)

    Cartagena-Sanchez, C. A.; Schaffner, D. A.; Johnson, H. K.; Fahim, L. E.

    2017-10-01

    A long-pulsed magnetic coaxial plasma gun is being implemented and characterized at the Bryn Mawr Plasma Laboratory (BMPL). A cold cathode discharged between the cylindrical electrodes generates and launches plasma into a 24cm diameter, 2m long chamber. Three separately pulsed magnetic coils are carefully positioned to generate radial magnetic field between the electrodes at the gun edge in order to provide stuffing field. Magnetic helicity is continuously injected into the flux-conserving vacuum chamber in a process akin to sustained slow-formation of spheromaks. The aim of this source, however, is to supply long pulses of turbulent magnetized plasma for measurement rather than for sustained spheromak production. The work shown here details the optimization of the magnetic field structure for this sustained helicity injection.

  15. Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

    International Nuclear Information System (INIS)

    Jammes, Christian; Filliatre, Philippe; Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan

    2015-01-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

  16. Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, Christian; Filliatre, Philippe [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan [Division of Applied Nuclear Physics, Uppsala University, SE-75120 Uppsala, (Sweden)

    2015-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

  17. Flux-limited diffusion coefficients in reactor physics applications

    International Nuclear Information System (INIS)

    Pounders, J.; Rahnema, F.; Szilard, R.

    2007-01-01

    Flux-limited diffusion theory has been successfully applied to problems in radiative transfer and radiation hydrodynamics, but its relevance to reactor physics has not yet been explored. The current investigation compares the performance of a flux-limited diffusion coefficient against the traditionally defined transport cross section. A one-dimensional BWR benchmark problem is examined at both the assembly and full-core level with varying degrees of heterogeneity. (authors)

  18. Thermal performance of Egypt's research reactor core (ET-RR-1)

    International Nuclear Information System (INIS)

    Khattab, M.; Mariy, A.

    1986-01-01

    The steady state thermal performance of the ET-RR-1 core system is theoretically investigated by different models describing the heat flux and the coolant mass flow rate. The magnitude of the heat generated by a fuel element depends upon its position in the core. Normal and uniform distributions for heat flux and coolant mass flow rate are considered. The clad and coolant temperatures at different core positions are evaluated and compared with the experimental measurements at different operating conditions. The results indicated large discrepancy between the predicted and the experimental results. Therefore, the previous models and the experimental results are evaluated in order to develop the best model that describes the thermal performance of the ET-RR-1 core. The adapted model gives 99.5% significant confidence limit. The effect of increasing the heat flux or decreasing the mass flow rate by 20% from its maximum recommended operating condition is tested and discussed. Also, the thermal behaviour towards increasing the reactor power more than its maximum operating condition is discussed. The present work could also be used in extending the investigation to other PWR reactor operating conditions

  19. Dynamo Tests for Stratification Below the Core-Mantle Boundary

    Science.gov (United States)

    Olson, P.; Landeau, M.

    2017-12-01

    Evidence from seismology, mineral physics, and core dynamics points to a layer with an overall stable stratification in the Earth's outer core, possibly thermal in origin, extending below the core-mantle boundary (CMB) for several hundred kilometers. In contrast, energetic deep mantle convection with elevated heat flux implies locally unstable thermal stratification below the CMB in places, consistent with interpretations of non-dipole geomagnetic field behavior that favor upwelling flows below the CMB. Here, we model the structure of convection and magnetic fields in the core using numerical dynamos with laterally heterogeneous boundary heat flux in order to rationalize this conflicting evidence. Strongly heterogeneous boundary heat flux generates localized convection beneath the CMB that coexists with an overall stable stratification there. Partially stratified dynamos have distinctive time average magnetic field structures. Without stratification or with stratification confined to a thin layer, the octupole component is small and the CMB magnetic field structure includes polar intensity minima. With more extensive stratification, the octupole component is large and the magnetic field structure includes intense patches or high intensity lobes in the polar regions. Comparisons with the time-averaged geomagnetic field are generally favorable for partial stratification in a thin layer but unfavorable for stratification in a thick layer beneath the CMB.

  20. Heat transfer for ultrahigh flux reactor

    International Nuclear Information System (INIS)

    Wadkins, R.P.; Lake, J.A.; Oh, C.H.

    1987-01-01

    The use of a uniquely designed nuclear reactor to supply neutrons for materials research is the focus of recent reactor design efforts. The biological, materials, and fundamental physics aspects of research require neutron fluxes much higher than present research and testing facilities can produce. The most advanced research using neutrons as probing detectors is being done in the High Flux Reactor at the Institut Laue Langeuin, France. The design of a reactor that can produce neutron fluxes of 1.0 x 10 16 n/cm 2 .s requires a relatively high power (300 MW range) and a small core volume (approximately 30 liters). This combination of power and volume leads to a high power density which places increased demands on thermal hydraulic margins

  1. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  2. Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

    International Nuclear Information System (INIS)

    Taryo, Taswanda

    2002-01-01

    Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

  3. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    Rousseau, A.; Genthon, J.P.

    1958-01-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  4. Improved HOR fuel management by flux measurement data feedback

    Energy Technology Data Exchange (ETDEWEB)

    Serov, I.V.; Leege, P.F.A. de; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft University of Technology, Reactor Physics Dep., Interfaculty Reactor Inst., Delft (Netherlands)

    1997-07-01

    Flux distribution in a nuclear reactor can be obtained by utilizing different calculational and experimental methods. The obtained flux distributions are associated with uncertainties and therefore always differ from each other. By combining information from the calculation and experiment using the confluence method, it is possible to obtain a more reliable estimate of the flux distribution than exhibited by the calculation or experiment separately. As a feedback, the fuel burnup distribution, which is used as initial data to the calculation can be improved as well. The confluence method is applied to improvement of the burnup distribution estimates for the HOR research reactor of the Delft University of Technology. An integrated code system CONHOR is developed to match the CITATION results of in-core foil activation rate calculations with in-core experimental data through confluence. The system forms the basis for the advanced fuel management of the reactor. (author) 1 fig., 8 refs.

  5. Temperature-dependent attenuation of ex-vessel flux measurements at the Hanford Fast Flux Test Facility

    International Nuclear Information System (INIS)

    McLane, F.E.; Wood, M.R.; Rathbun, J.L.

    1982-01-01

    Indicated nuclear power, developed by measuring leakage neutrons, has been found to be temperature dependent at the Hanford Fast Flux Test Facility (FFTF). The magnitude, sense and speed of response of the effect suggest that hot sodium above th core and shield is a significant cause. Future designs which may minimize this effect are discussed

  6. Generalized flux states of the t-J model

    International Nuclear Information System (INIS)

    Nori, F.; Abrahams, E.; Zimanyi, G.T.

    1990-01-01

    We investigate certain generalized flux phases arising in a mean-field approach to the t-J model. First, we establish that the energy of noninteracting electrons moving in a uniform magnetic field has an absolute minimum as a function of the flux at exactly one flux quantum per particle. Using this result, we show that if the hard-core nature of the hole bosons is taken into account, then the slave-boson mean-field approximation for the t-J Hamiltonian allows for a solution where both the spinons and the holons experience an average flux of one flux quantum per particle. This enables them to achieve the lowest possible energy within the manifold of spatially uniform flux states. In the case of the continuum model, this is possible only for certain fractional fillings and we speculate that the system may react to this frustration effect by phase separation

  7. THE JCMT GOULD BELT SURVEY: A FIRST LOOK AT DENSE CORES IN ORION B

    International Nuclear Information System (INIS)

    Kirk, H.; Francesco, J. Di; Johnstone, D.; Duarte-Cabral, A.; Hatchell, J.; Sadavoy, S.; Mottram, J. C.; Buckle, J.; Salji, C.; Berry, D. S.; Currie, M. J.; Jenness, T.; Broekhoven-Fiene, H.; Fich, M.; Tisi, S.; Nutter, D.; Quinn, C.; Pattle, K.; Pineda, J. E.; Hogerheijde, M. R.

    2016-01-01

    We present a first look at the SCUBA-2 observations of three sub-regions of the Orion B molecular cloud: LDN 1622, NGC 2023/2024, and NGC 2068/2071, from the JCMT Gould Belt Legacy Survey. We identify 29, 564, and 322 dense cores in L1622, NGC 2023/2024, and NGC 2068/2071 respectively, using the SCUBA-2 850 μm map, and present their basic properties, including their peak fluxes, total fluxes, and sizes, and an estimate of the corresponding 450 μm peak fluxes and total fluxes, using the FellWalker source extraction algorithm. Assuming a constant temperature of 20 K, the starless dense cores have a mass function similar to that found in previous dense core analyses, with a Salpeter-like slope at the high-mass end. The majority of cores appear stable to gravitational collapse when considering only thermal pressure; indeed, most of the cores which have masses above the thermal Jeans mass are already associated with at least one protostar. At higher cloud column densities, above 1–2 × 10 23 cm −2 , most of the mass is found within dense cores, while at lower cloud column densities, below 1 × 10 23 cm −2 , this fraction drops to 10% or lower. Overall, the fraction of dense cores associated with a protostar is quite small (<8%), but becomes larger for the densest and most centrally concentrated cores. NGC 2023/2024 and NGC 2068/2071 appear to be on the path to forming a significant number of stars in the future, while L1622 has little additional mass in dense cores to form many new stars

  8. Design of a PM Vernier Machine with Consideration for Modulation Flux and Comparison with Conventional PM motors

    Directory of Open Access Journals (Sweden)

    Byungtaek Kim

    2017-11-01

    Full Text Available This study deals with the core design of a PM vernier machine considering modulation flux effects, and the comparative investigation on volume and performance characteristics of the vernier over conventional PM machines are addressed. To these ends, for a PM vernier machine in operation at the base-speed, the flux density equations for teeth and yokes considering the flux modulation effects are derived, where the air gap harmonic permeance function is used. Using the derived equations, a PM vernier motor with specified yoke flux densities is designed. To identify the predicted flux yoke densities, the flux distribution and iron losses in core parts are analyzed through time-step finite element (FE simulations. Through Fourier series expansion of the air gap flux waves obtained by FE analysis at several specified times, the harmonic components constituting the flux waves are investigated and their speeds are also evaluated in numerical ways. Finally, to estimate the competitiveness of vernier machines versus conventional machines, the designed PM vernier motor is compared against two different conventional PM motors designed through the same design procedures in various aspects such as volume, torque capacity, efficiency, and power factor, in which, in particular, the core losses are included in efficiency calculation.

  9. Core losses of an inverter-fed permanent magnet synchronous motor with an amorphous stator core under no-load

    Directory of Open Access Journals (Sweden)

    Nicolas Denis

    2016-05-01

    Full Text Available In this paper, an interior permanent magnet synchronous motor (IPMSM with a stator core made of amorphous magnetic material (AMM is presented. The IPMSM is driven by a voltage source three-phase inverter with classical pulse width modulation (PWM control. The core losses under no-load condition are measured by experiment and compared to an equivalent IPMSM with a stator core made of NO steel. Under these conditions, the core losses are influenced by the stator, rotor and magnet shapes but also by the PWM carrier signal that implies a high frequency harmonic in the magnetic flux density. It is demonstrated that the AMM can reduce the core losses by about 56 %.

  10. FFTF [Fast Flux Test Facility] fuel handling experience (1979--1986)

    International Nuclear Information System (INIS)

    Romrell, D.M.; Art, D.M.; Redekopp, R.D.; Waldo, J.B.

    1987-05-01

    The Fast Flux Test Facility (FFTF)is a 400 MW (th) sodium-cooled fast flux test reactor located on the Hanford Site in southeastern Washington State. The FFTF is operated by the Westinghouse Hanford Company for the United States Department of Energy. The FFTF is a three loop plant designed primarily for the purpose of testing full-scale core components in an environment prototypic of future liquid metal reactors. The plant design emphasizes features to enhance this test capability, especially in the area of the core, reactor vessel, and refueling system. Eight special test positions are provided in the vessel head to permit contact instrumented experiments to be installed and irradiated. These test positions effectively divide the core into three sectors. Each sector requires its own In-Vessel Handling Machine (IVHM) to access all the core positions. Since the core and the in-vessel refueling components are submerged under sodium, all handling operations must be performed blind. This puts severe requirements on the positioning ability are reliability of the refueling components. This report addresses the operating experience with the fuel handling system from initial core loading in November, 1979 through 1986. This includes 9 refueling cycles. 2 refs., 8 figs

  11. SHREDI, Neutron Flux and Neutron Activation in 2-D Shields by Removal Diffusion

    International Nuclear Information System (INIS)

    Daneri, A.; Toselli, G.

    1976-01-01

    1 - Nature of physical problem solved: SHREDI is a removal - diffusion neutron shielding code. The program computes neutron fluxes and activations in bidimensional sections (x,y or r,z) of the shield. It is also possible to consider shielding points with the same y or z coordinate (mono-dimensional problems). 2 - Method of solution: The integrals which define the removal fluxes are computed in some shield points by means of a particular algorithm based on the Simpson's and trapezoidal rules. For the diffusion calculation the finite difference method is used. The removal sources are interpolated in all diffusion points by Chebyshev polynomials. 3 - Restrictions on the complexity of the problem: Maxima: number of removal energy groups NGR = 40; number of diffusion energy groups NGD = 40; number of the reactor core and shield materials NCMP = 50; number of core mesh points in r (or x) direction for integral calculation = 75; number of core mesh points in z (or y) direction for integral calculation = 75; number of core mesh points in theta (or z) direction for integral calculation = 75; number of shield mesh points for the neutron flux calculation in r (or x) direction NPX = 200; number of shield mesh points for the neutron flux calculation in z (or y) direction NPY = 200; n.b. (NPX * NPY) le 12000

  12. Simulation of the Point Lepreau core-follow history with SORO

    Energy Technology Data Exchange (ETDEWEB)

    Shanes, F.C.; Olive, C.G.; Cheng, I. [Nuclear Safety Solutions Limited, Toronto, Ontario (Canada); Banica, C. [Ontario Power Generation, Ajax, Ontario (Canada); Newman, C. [NB Power Nuclear, Point Lepreau Generating Station, Lepreau, New Brunswick (Canada); Nainer, O. [Bruce Power, Toronto, Ontario (Canada)

    2006-07-01

    The core tracking SORO program calculates the flux, burn-ups, bundle and channel powers, and is used for Ontario Power Generation and Bruce Power reactors. This paper presents the results of a comparison of the SORO cell fluxes against the Point Lepreau CANDU-6 measured flux data. A SORO model was created for the Point Lepreau reactor, and simulations were carried out to compare the SORO fluxes to the Travelling Flux Detector (TFD) scans and one year of operating history. Considering the good agreement between measured and computed fluxes, the results provide confidence that SORO accurately calculates the cell fluxes and bundle powers. (author)

  13. Simulation of the Point Lepreau core-follow history with SORO

    International Nuclear Information System (INIS)

    Shanes, F.C.; Olive, C.G.; Cheng, I.; Banica, C.; Newman, C.; Nainer, O.

    2006-01-01

    The core tracking SORO program calculates the flux, burn-ups, bundle and channel powers, and is used for Ontario Power Generation and Bruce Power reactors. This paper presents the results of a comparison of the SORO cell fluxes against the Point Lepreau CANDU-6 measured flux data. A SORO model was created for the Point Lepreau reactor, and simulations were carried out to compare the SORO fluxes to the Travelling Flux Detector (TFD) scans and one year of operating history. Considering the good agreement between measured and computed fluxes, the results provide confidence that SORO accurately calculates the cell fluxes and bundle powers. (author)

  14. Preliminary study of a flux converter for experimental reactor

    International Nuclear Information System (INIS)

    Malouch, M.F.

    1998-01-01

    The purpose of this project is to define the characteristics of a flux converter dedicated to increase the fast neutron flux in irradiation devices placed in the core of Osiris experimental reactor. This preliminary work has dealt with the neutronic and thermal-hydraulic aspects of this problem. The synthesis of the results produced by the codes APOLLO2, DAIXY, MERCURE5.3 and FLICA-3M shows that a cylindrical converter equipped with 5 fissile rings can enhance the fast flux by a 35% factor in an experimental device set in its center. (A.C.)

  15. Constraining genome-scale models to represent the bow tie structure of metabolism for 13C metabolic flux analysis

    DEFF Research Database (Denmark)

    Backman, Tyler W.H.; Ando, David; Singh, Jahnavi

    2018-01-01

    for a minimum of fluxes into core metabolism to satisfy these experimental constraints. Together, these methods accelerate and automate the identification of a biologically reasonable set of core reactions for use with 13C MFA or 2S- 13C MFA, as well as provide for a substantially lower set of flux bounds......Determination of internal metabolic fluxes is crucial for fundamental and applied biology because they map how carbon and electrons flow through metabolism to enable cell function. 13C Metabolic Flux Analysis (13C MFA) and Two-Scale 13C Metabolic Flux Analysis (2S-13C MFA) are two techniques used...

  16. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    Directory of Open Access Journals (Sweden)

    Imam Mahmoud M.

    2002-01-01

    Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

  17. Flux distribution measurements in the Bruce B Unit 6 reactor using a transportable traveling flux detector system

    International Nuclear Information System (INIS)

    Leung, T.C.; Drewell, N.H.; Hall, D.S.; Lopez, A.M.

    1987-01-01

    A transportable traveling flux detector (TFD) system for use in power reactors has been developed and tested at Chalk River Nuclear Labs. in Canada. It consists of a miniature fission chamber, a motor drive mechanism, a computerized control unit, and a data acquisition subsystem. The TFD system was initially designed for the in situ calibration of fixed self-powered detectors in operating power reactors and for flux measurements to verify reactor physics calculations. However, this system can also be used as a general diagnostic tool for the investigation of apparent detector failures and flux anomalies and to determine the movement of reactor internal components. This paper describes the first successful use of the computerized TFD system in an operating Canada deuterium uranium (CANDU) power reactor and the results obtained from the flux distribution measurements. An attempt is made to correlate minima in the flux profile with the locations of fuel channels so that future measurements can be used to determine the sag of the channels. Twenty-seven in-core flux detector assemblies in the 855-MW (electric) Unit 6 reactor of the Ontario Hydro Bruce B Generating Station were scanned

  18. Advantages of iron core in a tokamak

    International Nuclear Information System (INIS)

    Bettis, E.S.; Ballou, J.K.; Becraft, W.R.; Peng, Y.K.M.; Watts, H.L.

    1977-01-01

    A quantitative comparison of the iron core vs air core concepts was carried out on a preliminary basis by using a representative tokamak reactor design with the following self-consistent reference parameters. In the area of plasma engineering, poloidal field and MHD equilibrium considerations with an unsaturated iron core is discussed. The question of proper poloidal field coils to maintain D-shaped plasmas of relatively high anti β (7%) with a saturated iron core is also discussed. Estimates of the required iron core size, volt seconds, magnetic flux and its influence on force loading on the superconducting toroidal field coils are shown. Conceptual designs of the mechanical structure of an iron core device are presented. Favorable impacts on the OH power supply cost and complexity are indicated

  19. Plasma crowbars in cylindrical flux compression experiments

    International Nuclear Information System (INIS)

    Suter, L.J.

    1979-01-01

    We have done a series of one- and two-dimensional calculations of hard-core Z-pinch flux compression experiments in order to study the effect of a plasma on these systems. These calculations show that including a plasma can reduce the amount of flux lost during the compression. Flux losses to the outer wall of such experiments can be greatly reduced by a plasma conducting sheath which forms along the wall. This conducting sheath consists of a cold, dense high β, unmagnetized plasma which has enough pressure to balance a large field gradient. Flux which is lost into the center conductor is not effectively stopped by this plasma sheath until late in the implosion, at which time a layer similar to the one formed at the outer wall is created. Two-dimensionl simulations show that flux losses due to arching along the sliding contact of the experiment can be effectively stopped by the formation of a plasma conducting sheath

  20. 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)

  1. Fluid-structure-interaction of the pressurized water reactor core internals during blowdown - numerical simulation with a homogenization model

    International Nuclear Information System (INIS)

    Benner, J.

    1984-03-01

    A method for the numerical simulation of the Pressurized Water Reactor (PWR) core internal's behaviour during a blowdown accident is described, by which the motion of the reactor core and the interaction of the fuel elements with the core barrel and the coolant medium is calculated. Furthermore, some simple models for the support columns, lower and upper core support and the grid plate are provided. All these models have been implemented into the code Flux-4. For the solution of the very complex, coupled equations of motions for fluid and fuel rods an efficient numerical solution technique has been developed. With the new code-version Flux-5 the PWR-blowdown is parametically investigated. The calculated core barrel loadings are compared with Flux-4 results, simulating the core's inertia by a mass ring of HDR type. (orig.) [de

  2. A conceptual high flux reactor design with scope for use in ADS ...

    Indian Academy of Sciences (India)

    By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to ...

  3. Operation of SPHEX in helium and deuterium

    International Nuclear Information System (INIS)

    Gee, S.J.; Rusbridge, M.G.

    1994-01-01

    In the SPHEX spheromak a magnetised Marshall gun generates a toroidal plasma with embedded toroidal and poloidal magnetic fields. The resulting spheromak plasma is almost entirely governed by the programmed gun flux, Ψ G . Surprisingly then following ejection, the voltage V G between the gun electrodes attains a value which is largely independent of Ψ G . The SPHEX power supply acts as an approximate current source, consequently V G is entirely determined by the non linear load presented to it by the plasma, for any given gun current. The signature of initial gas breakdown in the gun is a period of low V G (∼ 140 V) called the pedestal. This is interpreted as the electrode sheath voltage, and persists throughout the discharge. After the initial gas breakdown, the gun voltage remains at this pedestal voltage as the gun current builds up, until the instant of plasma ejection when it starts to rise rapidly. Ejection is governed by a simple force balance at the gun muzzle, occurring when the internal azimuthal field due to the gun current exceeds the external radial field from the gun solenoid. The sharp rise in V G continues until it reaches some critical value when it abruptly stops, suggesting that a breakdown of the 'magnetic insulation' occurs, which is then subsequently maintained at a finite V G throughout the discharge. This implied breakdown cannot be 'gas like' which would tend to reduce V G to the pedestal voltage and would divert all of the gun current from the spheromak, neither of which are seen. Nor can it be due to 'short flux' linking the gun electrodes (effectively bypassing the spheromak) since this would short out the electrodes at any voltage. We propose that the observed gun voltage is limited by breakdown of the azimuthal magnetic insulation between the gun electrodes, by ions which acquire enough energy so that their Larmor diameter equals the inter electrode gap. (author) 7 refs., 4 figs

  4. Operation of SPHEX in helium and deuterium

    Energy Technology Data Exchange (ETDEWEB)

    Gee, S.J.; Rusbridge, M.G. [Manchester Univ. (United Kingdom). Dept. of Physics

    1994-12-31

    In the SPHEX spheromak a magnetised Marshall gun generates a toroidal plasma with embedded toroidal and poloidal magnetic fields. The resulting spheromak plasma is almost entirely governed by the programmed gun flux, {Psi}{sub G}. Surprisingly then following ejection, the voltage V{sub G} between the gun electrodes attains a value which is largely independent of {Psi}{sub G}. The SPHEX power supply acts as an approximate current source, consequently V{sub G} is entirely determined by the non linear load presented to it by the plasma, for any given gun current. The signature of initial gas breakdown in the gun is a period of low V{sub G} ({approx} 140 V) called the pedestal. This is interpreted as the electrode sheath voltage, and persists throughout the discharge. After the initial gas breakdown, the gun voltage remains at this pedestal voltage as the gun current builds up, until the instant of plasma ejection when it starts to rise rapidly. Ejection is governed by a simple force balance at the gun muzzle, occurring when the internal azimuthal field due to the gun current exceeds the external radial field from the gun solenoid. The sharp rise in V{sub G} continues until it reaches some critical value when it abruptly stops, suggesting that a breakdown of the `magnetic insulation` occurs, which is then subsequently maintained at a finite V{sub G} throughout the discharge. This implied breakdown cannot be `gas like` which would tend to reduce V{sub G} to the pedestal voltage and would divert all of the gun current from the spheromak, neither of which are seen. Nor can it be due to `short flux` linking the gun electrodes (effectively bypassing the spheromak) since this would short out the electrodes at any voltage. We propose that the observed gun voltage is limited by breakdown of the azimuthal magnetic insulation between the gun electrodes, by ions which acquire enough energy so that their Larmor diameter equals the inter electrode gap. (author) 7 refs., 4 figs.

  5. Theoretical aspects of magnetic helicity

    International Nuclear Information System (INIS)

    Hammer, J.H.

    1985-01-01

    The magnetic helicity, usually defined as K=integralA.Bdv, where A is the vector potential and B the magnetic field, measures the topological linkage of magnetic fluxes. Helicity manifests itself in the twistedness and knottedness of flux tubes. Its significance is that it is an ideal MHD invariant. While the helicity formalism has proven very useful in understanding reversed field pinch and spheromak behavior, some problems exist in applying the method consistently for complex (e.g., toroidal) conductor geometries or in situations where magnetic flux penetrates conducting walls. Recent work has attempted to generalize K to allow for all possible geometries

  6. Detecting the position of the moving-iron solenoid by non-displacement sensor based on parameter identification of flux linkage characteristics

    Science.gov (United States)

    Wang, Xuping; Quan, Long; Xiong, Guangyu

    2013-11-01

    Currently, most researches use signals, such as the coil current or voltage of solenoid, to identify parameters; typically, parameter identification method based on variation rate of coil current is applied for position estimation. The problem exists in these researches that the detected signals are prone to interference and difficult to obtain. This paper proposes a new method for detecting the core position by using flux characteristic quantity, which adds a new group of secondary winding to the coil of the ordinary switching electromagnet. On the basis of electromagnetic coupling theory analysis and simulation research of the magnetic field regarding the primary and secondary winding coils, and in accordance with the fact that under PWM control mode varying core position and operating current of windings produce different characteristic of flux increment of the secondary winding. The flux increment of the electromagnet winding can be obtained by conducting time domain integration for the induced voltage signal of the extracted secondary winding, and the core position from the two-dimensional fitting curve of the operating winding current and flux-linkage characteristic quantity of solenoid are calculated. The detecting and testing system of solenoid core position is developed based on the theoretical research. The testing results show that the flux characteristic quantity of switching electromagnet magnetic circuit is able to effectively show the core position and thus to accomplish the non-displacement transducer detection of the said core position of the switching electromagnet. This paper proposes a new method for detecting the core position by using flux characteristic quantity, which provides a new theory and method for switch solenoid to control the proportional valve.

  7. Quality of Metal Deposited Flux Cored Wire With the System Fe-C-Si-Mn-Cr-Mo-Ni-V-Co

    Science.gov (United States)

    Gusev, Aleksander I.; Kozyrev, Nikolay A.; Osetkovskiy, Ivan V.; Kryukov, Roman E.; Kozyreva, Olga A.

    2017-10-01

    Studied the effect of the introduction of vanadium and cobalt into the charge powder fused wire system Fe-C-Si-Mn-Cr-Ni-Mo-V, used in cladding assemblies and equipment parts and mechanisms operating under abrasive and abrasive shock loads. the cored wires samples were manufactured in the laboratory conditions and using appropriate powder materials and as a carbonfluoride contained material were used the dust from gas purification of aluminum production, with the following components composition, %: Al2O3 = 21-46.23; F = 18-27; Na2O = 8-15; K2O = 0.4-6; CaO = 0.7-2.3; Si2O = 0.5-2.48; Fe2O3 = 2.1-3.27; C = 12.5-30.2; MnO = 0.07-0.9; MgO = 0.06-0.9; S = 0.09-0.19; P = 0.1-0.18. Surfacing was produced on the St3 metal plates in 6 layers under the AN-26C flux by welding truck ASAW-1250. Cutting and preparation of samples for research had been implemented. The chemical composition and the hydrogen content of the weld metal were determined by modern methods. The hardness and abrasion rate of weld metal had been measured. Conducted metallographic studies of weld metal: estimated microstructure, grain size, contamination of oxide non-metallic inclusions. Metallographic studies showed that the microstructure of the surfaced layer by cored wire system Fe-C-Si-Mn-Cr-Mo-Ni-V-Co is uniform, thin dendrite branches are observed. The microstructure consists of martensite, which is formed inside the borders of the former austenite grain retained austenite present in small amounts in the form of separate islands, and thin layers of δ-ferrite, which is located on the borders of the former austenite grains. Carried out an assessment the effect of the chemical composition of the deposited metal on the hardness and wear and hydrogen content. In consequence of multivariate correlation analysis, it was determined dependence to the hardness of the deposited layer and the wear resistance of the mass fraction of the elements included in the flux-cored wires of the system Fe

  8. THE JCMT GOULD BELT SURVEY: A FIRST LOOK AT DENSE CORES IN ORION B

    Energy Technology Data Exchange (ETDEWEB)

    Kirk, H.; Francesco, J. Di; Johnstone, D. [NRC Herzberg Astronomy and Astrophysics, 5071 West Saanich Rd, Victoria, BC, V9E 2E7 (Canada); Duarte-Cabral, A.; Hatchell, J. [Physics and Astronomy, University of Exeter, Stocker Road, Exeter EX4 4QL (United Kingdom); Sadavoy, S.; Mottram, J. C. [Max Planck Institute for Astronomy, Königstuhl 17, D-69117 Heidelberg (Germany); Buckle, J.; Salji, C. [Astrophysics Group, Cavendish Laboratory, J J Thomson Avenue, Cambridge, CB3 0HE (United Kingdom); Berry, D. S.; Currie, M. J.; Jenness, T. [Joint Astronomy Centre, 660 N. A‘ohōkū Place, University Park, Hilo, Hawaii 96720 (United States); Broekhoven-Fiene, H. [Department of Physics and Astronomy, University of Victoria, Victoria, BC, V8P 1A1 (Canada); Fich, M.; Tisi, S. [Department of Physics and Astronomy, University of Waterloo, Waterloo, Ontario, N2L 3G1 (Canada); Nutter, D.; Quinn, C. [School of Physics and Astronomy, Cardiff University, The Parade, Cardiff, CF24 3AA (United Kingdom); Pattle, K. [Jeremiah Horrocks Institute, University of Central Lancashire, Preston, Lancashire, PR1 2HE (United Kingdom); Pineda, J. E. [European Southern Observatory (ESO), Garching (Germany); Hogerheijde, M. R. [Leiden Observatory, Leiden University, P.O. Box 9513, 2300 RA Leiden (Netherlands); and others

    2016-02-01

    We present a first look at the SCUBA-2 observations of three sub-regions of the Orion B molecular cloud: LDN 1622, NGC 2023/2024, and NGC 2068/2071, from the JCMT Gould Belt Legacy Survey. We identify 29, 564, and 322 dense cores in L1622, NGC 2023/2024, and NGC 2068/2071 respectively, using the SCUBA-2 850 μm map, and present their basic properties, including their peak fluxes, total fluxes, and sizes, and an estimate of the corresponding 450 μm peak fluxes and total fluxes, using the FellWalker source extraction algorithm. Assuming a constant temperature of 20 K, the starless dense cores have a mass function similar to that found in previous dense core analyses, with a Salpeter-like slope at the high-mass end. The majority of cores appear stable to gravitational collapse when considering only thermal pressure; indeed, most of the cores which have masses above the thermal Jeans mass are already associated with at least one protostar. At higher cloud column densities, above 1–2 × 10{sup 23} cm{sup −2}, most of the mass is found within dense cores, while at lower cloud column densities, below 1 × 10{sup 23} cm{sup −2}, this fraction drops to 10% or lower. Overall, the fraction of dense cores associated with a protostar is quite small (<8%), but becomes larger for the densest and most centrally concentrated cores. NGC 2023/2024 and NGC 2068/2071 appear to be on the path to forming a significant number of stars in the future, while L1622 has little additional mass in dense cores to form many new stars.

  9. Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

    International Nuclear Information System (INIS)

    Oakley, D.J.

    1978-01-01

    The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

  10. MTR core loading pattern optimization using burnup dependent group constants

    Directory of Open Access Journals (Sweden)

    Iqbal Masood

    2008-01-01

    Full Text Available A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.

  11. Prediction of human core body temperature using non-invasive measurement methods.

    Science.gov (United States)

    Niedermann, Reto; Wyss, Eva; Annaheim, Simon; Psikuta, Agnes; Davey, Sarah; Rossi, René Michel

    2014-01-01

    The measurement of core body temperature is an efficient method for monitoring heat stress amongst workers in hot conditions. However, invasive measurement of core body temperature (e.g. rectal, intestinal, oesophageal temperature) is impractical for such applications. Therefore, the aim of this study was to define relevant non-invasive measures to predict core body temperature under various conditions. We conducted two human subject studies with different experimental protocols, different environmental temperatures (10 °C, 30 °C) and different subjects. In both studies the same non-invasive measurement methods (skin temperature, skin heat flux, heart rate) were applied. A principle component analysis was conducted to extract independent factors, which were then used in a linear regression model. We identified six parameters (three skin temperatures, two skin heat fluxes and heart rate), which were included for the calculation of two factors. The predictive value of these factors for core body temperature was evaluated by a multiple regression analysis. The calculated root mean square deviation (rmsd) was in the range from 0.28 °C to 0.34 °C for all environmental conditions. These errors are similar to previous models using non-invasive measures to predict core body temperature. The results from this study illustrate that multiple physiological parameters (e.g. skin temperature and skin heat fluxes) are needed to predict core body temperature. In addition, the physiological measurements chosen in this study and the algorithm defined in this work are potentially applicable as real-time core body temperature monitoring to assess health risk in broad range of working conditions.

  12. Behaviour of aged and new flux detectors in Darlington reactors

    Energy Technology Data Exchange (ETDEWEB)

    Banica, C.; Foster, M., E-mail: Constantin.Banica@OPG.com [Ontario Power Generation, Darlington Nuclear, Bowmanville, Ontario (Canada)

    2013-07-01

    In-core neutron flux detectors are used for protective and safety functions in the Darlington NGS 'A' CANDU reactors. This paper presents new observations regarding the aging of flux detectors, including response to fuelling, response to unit shutdown and indicators of detector noise. Comparisons of detector signals before and after replacement confirm previous assumptions about aging effects. (author)

  13. Optimization of neutron flux distribution in Isotope Production Reactor

    International Nuclear Information System (INIS)

    Valladares, G.L.

    1988-01-01

    In order to optimize the thermal neutrons flux distribution in a Radioisotope Production and Research Reactor, the influence of two reactor parameters was studied, namely the Vmod / Vcomb ratio and the core volume. The reactor core is built with uranium oxide pellets (UO 2 ) mounted in rod clusters, with an enrichment level of ∼3 %, similar to LIGHT WATER POWER REATOR (LWR) fuel elements. (author) [pt

  14. Measure of thermal neutron flux in the IPEN/MB-01 reactor using 197 Au wire activation detectors

    International Nuclear Information System (INIS)

    Marques, Andre Luis Ferreira

    1995-01-01

    This dissertation has aimed at developing a neutron flux measurement technique by means of detectors activation analysis. The main task of this work was the implementation of this thermal neutron flux measurement technique, using gold wires as activation detectors in the IPEN/MB-01 reactor core. The neutron thermal flux spatial distribution was obtained by gold wire activation technique, with wire diameters of 0.125 mm and 0.250 mm in seven selected reactor experimental channels. The values of thermal flux were about 10 9 neutrons/cm 2 .s. This experiment has been the first one conducted with gold wires in the IPEN/MB-01 reactor, being this technique implemented for use by experiments in flux mapping of the core

  15. Empirical observations on the aging of flux detectors at Darlington

    Energy Technology Data Exchange (ETDEWEB)

    Banica, C. [Ontario Power Generation, Darlington Nuclear, Bowmanville, Ontario (Canada); Slovak, R. [Ontario Power Generation, IMandCS, Pickering, Ontario (Canada)

    2011-07-01

    In-core neutron flux detectors are used for protective and safety functions in the Darlington CANDU reactors. This paper presents observations to date regarding aging of detectors, including recent measurements of prompt fractions and lead cable behaviour during a reactor power rundown. Linear models have been used to estimate and predict the prompt fraction evolution in time using independent variables such as the integrated neutron flux at the detector location, the length of the detector lead cable and the residual current at near-zero flux. (author)

  16. Self-powered neutron and gamma-ray flux detector

    International Nuclear Information System (INIS)

    Allan, C.J.; Shields, R.B.; Lynch, G.F.; Cuttler, J.M.

    1980-01-01

    A new type of self-powered neutron detector was developed which is sensitive to both the neutron and gamma-ray fluxes. The emitter comprises two parts. The central emitter core is made of materials that generate high-energy electrons on exposure to neutrons. The outer layer acts as a gamma-ray/electron converter, and since it has a higher atomic number and higher back-scattering coefficient than the collector, increases the net outflow or emmission of electrons. The collector, which is around the emitter outer layer, is insulated from the outer layer electrically with dielectric insulation formed from compressed metal-oxide powder. The fraction of electrons given off by the emitter that is reflected back by the collector is less than the fraction of electrons emitted by the collector that is reflected back by the emitter. The thickness of the outer layer needed to achieve this result is very small. A detector of this design responds to external reactor gamma-rays as well as to neutron capture gamma-rays from the collector. The emitter core is either nickel, iron or titanium, or alloys based on these metals. The outer layer is made of platinum, tantalum, osmium, molybdenum or cerium. The detector is particularly useful for monitoring neutron and gamma ray flux intensities in nuclear reactor cores in which the neutron and gamma ray flux intensities are closely proportional, are unltimately related to the fission rate, and are used as measurements of nuclear reactor power. (DN)

  17. Historical deposition and fluxes of mercury in Narraguinnep Reservoir, southwestern Colorado, USA

    International Nuclear Information System (INIS)

    Gray, John E.; Fey, David L.; Holmes, Charles W.; Lasorsa, Brenda K.

    2005-01-01

    Narraguinnep Reservoir has been identified as containing fish with elevated Hg concentrations and has been posted with an advisory recommending against consumption of fish. There are presently no point sources of significant Hg contamination to this reservoir or its supply waters. To evaluate potential historical Hg sources and deposition of Hg to Narraguinnep Reservoir, the authors measured Hg concentrations in sediment cores collected from this reservoir. The cores were dated by the 137 Cs method and these dates were further refined by relating water supply basin hydrological records with core sedimentology. Rates of historical Hg flux were calculated (ng/cm 2 /a) based on the Hg concentrations in the cores, sediment bulk densities, and sedimentation rates. The flux of Hg found in Narraguinnep Reservoir increased by approximately a factor of 2 after about 1970. The 3 most likely sources of Hg to Narraguinnep Reservoir are surrounding bedrocks, upstream inactive Au-Ag mines, and several coal-fired electric power plants in the Four Corners region. Patterns of Hg flux do not support dominant Hg derivation from surrounding bedrocks or upstream mining sources. There are 14 coal-fired power plants within 320 km of Narraguinnep Reservoir that produce over 80 x 10 6 MWH of power and about 1640 kg-Hg/a are released through stack emissions, contributing significant Hg to the surrounding environment. Two of the largest power plants, located within 80 km of the reservoir, emit about 950 kg-Hg/a. Spatial and temporal patterns of Hg fluxes for sediment cores collected from Narraguinnep Reservoir suggest that the most likely source of Hg to this reservoir is from atmospheric emissions from the coal-fired electric power plants, the largest of which began operation in this region in the late-1960s and early 1970s

  18. 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)

  19. ETRR-2 in-core fuel management strategy

    International Nuclear Information System (INIS)

    Khalil, M.Y.; Amin, Esmat; Belal, M.G.

    2005-01-01

    The Egypt second research reactor has many irradiation channels, beam tubes and irradiation boxes, inside and outside the reactor core. The core reload configuration has great effect on the core performance and fluxes in the irradiation channels. This paper deals with the design and safety analysis that were performed for the determination of ETRR2 in-core fuel management strategy which fulfills neutronic design criteria, safety reactor operation, utility optimization and achieve the overall fuel management criteria. The core is divided into 8 zones, in order to obtain the minimum and adjacent fuel movement scheme that is recommended from the operational point of view. Then a search for the initial core using backward iteration, one get different initial cores, one initial core would assume the equilibrium core after 250 full power days of operation, while the other assumes equilibrium after 199 full power days, and shows a better performance of power peaking factor. (author)

  20. 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

  1. Iron Losses in Electrical Machines Due to Non Sinusoidal Alternating Fluxes

    DEFF Research Database (Denmark)

    Ritchie, Ewen; Walker, J.A.; Dorrell, D. G.

    2007-01-01

    This paper shows how the flux waveform in the core of an electrical machine can be vary non- sinusoidally which complicates the calculation of the iron loss in a machine. A set of tests are conducted on a steel sample using an Epstein square where harmonics are injected into the flux waveform which...... of a machine....

  2. Generating the flux map of Nigeria Research Reactor-1 for efficient ...

    African Journals Online (AJOL)

    One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

  3. Multiobjective optimization for design of multifunctional sandwich panel heat pipes with micro-architected truss cores

    International Nuclear Information System (INIS)

    Roper, Christopher S.

    2011-01-01

    A micro-architected multifunctional structure, a sandwich panel heat pipe with a micro-scale truss core and arterial wick, is modeled and optimized. To characterize multiple functionalities, objective equations are formulated for density, compressive modulus, compressive strength, and maximum heat flux. Multiobjective optimization is used to determine the Pareto-optimal design surfaces, which consist of hundreds of individually optimized designs. The Pareto-optimal surfaces for different working fluids (water, ethanol, and perfluoro(methylcyclohexane)) as well as different micro-scale truss core materials (metal, ceramic, and polymer) are determined and compared. Examination of the Pareto fronts allows comparison of the trade-offs between density, compressive stiffness, compressive strength, and maximum heat flux in the design of multifunctional sandwich panel heat pipes with micro-scale truss cores. Heat fluxes up to 3.0 MW/m 2 are predicted for silicon carbide truss core heat pipes with water as the working fluid.

  4. Reactivity anomaly surveillance in the Fast Flux Test Facility through cycle 3

    International Nuclear Information System (INIS)

    Knutson, B.J.; Harris, R.A.

    1984-08-01

    The technique for monitoring core reactivity during power operation used at the Fast Flux Test Facility (FFTF) is described. This technique relies on comparing predicted to measured rod positions to detect any anomalous (or unpredicted) core reactivity changes. It is implemented on the Plant Data System (PDS) computer and thus provides rapid indication of any abnormal core conditions. The prediction algorithms use thermal-hydraulic, control rod position and neutron flux sensor information to predict the core reactivity state. Initial results of using this technique based mainly on theoretical formulations is presented. The results show that the reactivity changes due to increasing reactor power (power defect) and burnup of the fuel were within approx. 16% of predicted values. To increase the sensitivity and accuracy of this technique, the prediction algorithms were calibrated to actual operating data. The work of calibrating this technique and the results of using the calibrated technique up through the third full operating cycle are summarized

  5. The Role of Water Movement and Spatial Scaling for Measurement of Dissolved Inorganic Nitrogen Fluxes in Intertidal Sediments

    Science.gov (United States)

    Asmus, R. M.; Jensen, M. H.; Jensen, K. M.; Kristensen, E.; Asmus, H.; Wille, A.

    1998-02-01

    Fluxes of dissolved inorganic nitrogen (ammonium and nitrate) across the sediment-water interface were determined at intertidal locations in Königshafen, northern Wadden Sea, North Sea. Three different incubation techniques were compared: closed sediment cores (small scale), closed bell jars (medium scale) and an open flow system (Sylt flume, large scale). Water movement in the two closed systems was maintained below the resuspension limit by spinning magnets (cores, incubated in the laboratory) or by transfer of wave action through flexible plastic foil (bell jars,in situ), whereas in the flume system (in situ) water movement was unidirectional, driven by currents and waves. Data sets from several years of core measurements (1992-94), bell jar measurements (1980) and flume campaigns (1990-93) served as the basis for a comparison of dissolved inorganic nitrogen fluxes. Fluxes of ammonium and nitrate were within the same order of magnitude in closed cores and bell jars, while flume rates of ammonium were considerably higher. The high flume rates were caused by advective flushing due to tidal water movement and wave action. The release of ammonium increased significantly with current velocity between 1 and 13 cm s-1. Fluxes of ammonium were higher in sediments withArenicola marinacompared to those without this bioturbating species. The influence of benthic microalgae was evident only in the small and medium scale core and bell jar systems as reduced ammonium release during light exposure. Nitrate was consumed by sediments in both closed systems at a rate proportional to the nitrate concentration in the overlying water. Nitrate fluxes in the large scale Sylt flume were low with an average of only 7% of the ammonium fluxes, probably due to low concentrations in tidal waters during measurements (summer). Both closed, small scale or open, large scale techniques can be applied successfully for benthic flux studies, but the actual choice depends on the purpose of the

  6. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  7. Critical experiments of JMTRC MEU cores

    International Nuclear Information System (INIS)

    Nagaoka, Y.; Takeda, K.; Shimakawa, S.; Koike, S.; Oyamada, R.

    1984-01-01

    The JMTRC, the critical facility of the Japan Materials Testing Reactor (JMTR), went critical on August 29, 1983, with 14 medium enriched uranium (MEU, 45%) fuel elements. Experiments are now being carried out to measure the change in various reactor characteristics between the previous HEU core and the new MEU fueled core. This paper describes the results obtained thus far on critical mass, excess reactivity, control rod worths and flux distribution, including preliminary neutronics calculations for the experiments using the SRAC code. (author)

  8. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    Energy Technology Data Exchange (ETDEWEB)

    Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  9. Thermal-hydraulic mixing in the split-core ANS reactor design

    International Nuclear Information System (INIS)

    Dorning, R.J.J.

    1988-01-01

    A design has been proposed for the advanced neutron source (ANS) reactor that incorporates a split core, one purpose of which is to create a mixing plenum between the upper and lower cores. It was hoped that in addition to introducing various desirable neutronics features, such as decreasing the fast neutron flux contamination of thermal and cold neutron beams located in the reactor midplane, this mixing plenum would make possible higher operating powers by lowering the maximum core temperature. This lower temperature was to be achieved as a result of the mixing, of the hot D 2 O coolant exiting the upper-core channels, and the cold D 2 O leaving the large upper core bypass. It was expected that this mixing would bring about a significantly reduced lower core maximum coolant inlet temperature. The authors have carried out large-scale computer calculations to determine the extent to which this mixing occurs in current split-core design geometry, which does not incorporate baffles, mixing devices, or other design features introduced to enhance mixing. The large-scale self-consistent calculations summarized here indicate that innovative design ideas to enhance mixing will be necessary if the split-core concept is to achieve the amount of thermal mixing needed to make possible significantly higher power operation and corresponding higher flux sources

  10. The inverse Numerical Computer Program FLUX-BOT for estimating Vertical Water Fluxes from Temperature Time-Series.

    Science.gov (United States)

    Trauth, N.; Schmidt, C.; Munz, M.

    2016-12-01

    Heat as a natural tracer to quantify water fluxes between groundwater and surface water has evolved to a standard hydrological method. Typically, time series of temperatures in the surface water and in the sediment are observed and are subsequently evaluated by a vertical 1D representation of heat transport by advection and dispersion. Several analytical solutions as well as their implementation into user-friendly software exist in order to estimate water fluxes from the observed temperatures. Analytical solutions can be easily implemented but assumptions on the boundary conditions have to be made a priori, e.g. sinusoidal upper temperature boundary. Numerical models offer more flexibility and can handle temperature data which is characterized by irregular variations such as storm-event induced temperature changes and thus cannot readily be incorporated in analytical solutions. This also reduced the effort of data preprocessing such as the extraction of the diurnal temperature variation. We developed a software to estimate water FLUXes Based On Temperatures- FLUX-BOT. FLUX-BOT is a numerical code written in MATLAB which is intended to calculate vertical water fluxes in saturated sediments, based on the inversion of measured temperature time series observed at multiple depths. It applies a cell-centered Crank-Nicolson implicit finite difference scheme to solve the one-dimensional heat advection-conduction equation. Besides its core inverse numerical routines, FLUX-BOT includes functions visualizing the results and functions for performing uncertainty analysis. We provide applications of FLUX-BOT to generic as well as to measured temperature data to demonstrate its performance.

  11. Vanadium Beta Emission Detectors for Reactor In-Core Neutron Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, I Oe; Soederlund, B

    1969-06-15

    In-core flux measurements are becoming increasingly important in both power reactors and test reactors. In particular power distribution measurements in large power reactors have to be performed with a great number of neutron detectors capable of withstanding high integrated flux values. This report presents a summary of the development and application of a new type of nuclear radiation sensor, a beta emission detector, for measurements at high neutron flux levels. The work has been carried out at the Section for Instrumentation and has been the basis for a type of neutron detector employed in the Marviken in-core system as well as for other types. The report describes the design and principle of operation, experiments and tests. Also included are the results and comments from a long-term irradiation of some detectors in the Halden reactor.

  12. Maximum neutron flux in thermal reactors

    International Nuclear Information System (INIS)

    Strugar, P.V.

    1968-12-01

    Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples

  13. Permanent-magnet switched-flux machine

    Science.gov (United States)

    Trzynadlowski, Andrzej M.; Qin, Ling

    2010-01-12

    A permanent-magnet switched-flux (PMSF) device has a ferromagnetic outer stator mounted to a shaft about a central axis extending axially through the PMSF device. Pluralities of top and bottom stator poles are respectively mounted in first and second circles, radially outwardly in first and second transverse planes extending from first and second sections of the central axis adjacent to an inner surface of the ferromagnetic outer stator. A ferromagnetic inner rotor is coupled to the shaft and has i) a rotor core having a core axis co-axial with the central axis; and ii) first and second discs having respective outer edges with first and second pluralities of permanent magnets (PMs) mounted in first and second circles, radially outwardly from the rotor core axis in the first and second transverse planes. The first and second pluralities of PMs each include PMs of alternating polarity.

  14. Land Use, Land Use History, and Soil Type Affect Soil Greenhouse Gas Fluxes From Agricultural Landscapes of the East African Highlands

    Science.gov (United States)

    Wanyama, I.; Rufino, M. C.; Pelster, D. E.; Wanyama, G.; Atzberger, C.; van Asten, P.; Verchot, Louis V.; Butterbach-Bahl, K.

    2018-03-01

    This study aims to explain effects of soil textural class, topography, land use, and land use history on soil greenhouse gas (GHG) fluxes in the Lake Victoria region. We measured GHG fluxes from intact soil cores collected in Rakai, Uganda, an area characterized by low-input smallholder (soil cores were air dried and rewetted to water holding capacities (WHCs) of 30, 55, and 80%. Soil CO2, CH4, and N2O fluxes were measured for 48 h following rewetting. Cumulative N2O fluxes were highest from soils under perennial crops and the lowest from soils under annual crops (P soils had lower N2O fluxes than the clay soils (P soil CO2 fluxes were highest from eucalyptus plantations and lowest from annual crops across multiple WHC (P = 0.014 at 30% WHC and P soil cores from the top soil. This study reveals that land use and soil type have strong effects on GHG fluxes from agricultural land in the study area. Field monitoring of fluxes is needed to confirm whether these findings are consistent with what happens in situ.

  15. Fast neutron flux in heavy water reactors; Flux de neutrons rapides dans les piles a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Brisbois, J; Katz, S [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France)

    1966-07-01

    The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author) [French] La possibilite de calculer le flux de neutrons rapides dans un reseau d'uranium naturel a eau lourde par superposition des apports des divers barreaux, a ete verifiee en utilisant un code Monte-Carlo monodimensionel. Les resultats obtenus concordent avec des mesures experimentales effectuees dans le coeur et reacteur de la pile Aquilon. (auteurs)

  16. Structure and optical properties of cored wurtzite (Zn,Mg)O heteroepitaxial nanowires

    International Nuclear Information System (INIS)

    Heo, Y.W.; Abernathy, C.; Pruessner, K.; Sigmund, W.; Norton, D.P.; Overberg, M.; Ren, F.; Chisholm, M.F.

    2004-01-01

    The synthesis, structure, and optical properties of one-dimensional heteroepitaxial cored (Zn,Mg)O semiconductor nanowires grown by a catalyst-driven molecular beam epitaxy technique are discussed. The structures form spontaneously in a Zn, Mg and O 2 /O 3 flux, consisting of a single crystal, Zn-rich Zn 1-x Mg x O(x 1-y Mg y O(y>>0.02) sheath. High resolution Z-contrast scanning transmission electron microscopy shows core diameters as small as 4 nm. The cored structure forms spontaneously under constant flux due to a bimodal growth mechanism in which the core forms via bulk like vapor-liquid-solid growth, while the outer sheath grows as a heteroepitaxial layer. Temperature-dependent photoluminescence shows a slight blueshift in the near band edge peak, which is attributed to a few percent Mg doping in the nanoscale ZnO core. The catalyst-driven molecular beam epitaxy technique provides for site-specific nanorod growth on arbitrary substrates

  17. SLAROM, Neutron Flux Distribution and Spectra in Lattice Cell

    International Nuclear Information System (INIS)

    Nakagawa, M.; Tsuchihashi, K.

    2002-01-01

    1 - Description of program or function: SLAROM solves the neutron integral transport equations to determine flux distribution and spectra in a lattice and calculates cell averaged effective cross sections. 2 - Method of solution: Collision probability method for cell calculation and 1D diffusion for core calculation. 3 - Restrictions on the complexity of the problem: Variable dimensions are used throughout the program so that computer core requirements depend on a variety of program parameters

  18. 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)

  19. 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

  20. SEARCH FOR GLOBAL f-MODES AND p-MODES IN THE {sup 8}B NEUTRINO FLUX

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Ilídio, E-mail: ilidio.lopes@ist.utl.pt, E-mail: ilopes@uevora.pt [Centro Multidisciplinar de Astrofísica, Instituto Superior Técnico, Universidade de Lisboa, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Departamento de Física, Escola de Ciências e Tecnologia, Universidade de Évora, Colégio Luis António Verney, 7002-554 Évora (Portugal)

    2013-11-01

    The impact of global acoustic modes on the {sup 8}B neutrino flux time series is computed for the first time. It is shown that the time fluctuations of the {sup 8}B neutrino flux depend on the amplitude of acoustic eigenfunctions in the region where the {sup 8}B neutrino flux is produced: modes with low n (or order) that have eigenfunctions with a relatively large amplitude in the Sun's core strongly affect the neutrino flux; conversely, modes with high n that have eigenfunctions with a minimal amplitude in the Sun's core have a very small impact on the neutrino flux. It was found that the global modes with a larger impact on the {sup 8}B neutrino flux have a frequency of oscillation in the interval 250 μHz to 500 μHz (or a period in the interval 30 minutes to 70 minutes), such as the f-modes (n = 0) for the low degrees, radial modes of order n ≤ 3, and the dipole mode of order n = 1. Their corresponding neutrino eigenfunctions are very sensitive to the solar inner core and are unaffected by the variability of the external layers of the solar surface. If time variability of neutrinos is observed for these modes, it will lead to new ways of improving the sound speed profile inversion in the central region of the Sun.

  1. Characteristic test of initial HTTR core

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Shimakawa, Satoshi; Fujimoto, Nozomu; Goto, Minoru

    2004-01-01

    This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations

  2. 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

  3. Estimating biological elementary flux modes that decompose a flux distribution by the minimal branching property

    DEFF Research Database (Denmark)

    Chan, Siu Hung Joshua; Solem, Christian; Jensen, Peter Ruhdal

    2014-01-01

    biologically feasible EFMs by considering their graphical properties. A previous study on the transcriptional regulation of metabolic genes found that distinct branches at a branch point metabolite usually belong to distinct metabolic pathways. This suggests an intuitive property of biologically feasible EFMs......, i.e. minimal branching. RESULTS: We developed the concept of minimal branching EFM and derived the minimal branching decomposition (MBD) to decompose flux distributions. Testing in the core Escherichia coli metabolic network indicated that MBD can distinguish branches at branch points and greatly...... knowledge, which facilitates interpretation. Comparison of the methods applied to a complex flux distribution in Lactococcus lactis similarly showed the advantages of MBD. The minimal branching EFM concept underlying MBD should be useful in other applications....

  4. JOYO MK-II core characteristics database

    International Nuclear Information System (INIS)

    Tabuchi, Shiro; Aoyama, Takafumi; Nagasaki, Hideaki; Kato, Yuichi

    1998-12-01

    The experimental fast reactor JOYO served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, extensive data were accumulated from the core 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. The code system MAGI has been developed and used for core management of JOYO MK-II, and the core characteristics and the irradiation test conditions were calculated using MAGI on the basis of three dimensional diffusion theory with seven neutron energy groups. The core management data include extensive data, which were recorded on CD-ROM for user convenience. The data are specifications and configurations of the core, and for about 300 driver fuel subassemblies and about 60 uninstrumented irradiation subassemblies are core composition before and after irradiation, neutron flux, neutron fluences, fuel and control rod burn-up, and temperature and power distributions. MK-II core characteristics and test conditions were stored in the database for post analysis. Core characteristics data include excess reactivities, control rod worths, and reactivity coefficients, e.g., temperature, power and burn-up. Test conditions include both measured and calculated data for irradiation conditions. (author)

  5. Energy selecting action of limiters on particle fluxes penetrating into the SOL-plasma

    International Nuclear Information System (INIS)

    Hildebrandt, D.

    1986-01-01

    A single model of the penetration of particle effluxes from the core plasma into the SOL-plasma of tokamaks is proposed. The assumptions made are free streaming of particles parallel to the magnetic field and anomalous particle transport perpendicular to the toroidal field with a constant radial velocity. The model has been proved for measured particle fluxes of Li which was injected into the core plasma of the tokamak T-10. The dependence of the Li-particle flux on the minor radius as well as toroidal asymmetries in the SOL-plasma can be explained by the results of the calculations. (author)

  6. Analysis of fuel management pattern of research reactor core of the MTR type design

    International Nuclear Information System (INIS)

    Lily Suparlina; Tukiran Surbakti

    2014-01-01

    Research reactor core design needs neutronics parameter calculation use computer codes. Research reactor MTR type is very interested because can be used as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thermal neutron flux in the core is 1.0x10 15 n/cm 2 s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of high and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronics parameter calculation is done for new U-9Mo-Al fuel with variation of densities.The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0 x 10 15 n/cm 2 s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. (author)

  7. Comparison of VLBI radio core and X-ray flux densities of extragalactic radio sources

    International Nuclear Information System (INIS)

    Bloom, S.D.; Marscher, A.P.

    1990-01-01

    The Einstein Observatory revealed that most quasars, selected in a variety of ways, are strong x-ray emitters. Radio bright quasars are statistically more luminous in the x-ray than their radio-quiet counterparts. It was also found that the 90 GHz to soft x-ray spectral index has a very small dispersion for sources selected by their strong millimeter emission. This implies a close relationship between compact radio flux density and x-ray emission. Strong correlations have been found between the arcsecond scale flux densities and soft x-ray fluxes. It is suggested that the correlation can be explained if the soft x-rays were produced by the synchrotron self-Compton (SSC) process within the compact radio emitting region. (author)

  8. Observations and modeling of magnetized plasma jets and bubbles launched into a transverse B-field

    Science.gov (United States)

    Fisher, Dustin M.; Zhang, Yue; Wallace, Ben; Gilmore, Mark; Manchester, Ward B., IV; van der Holst, Bart; Rogers, Barrett N.; Hsu, Scott C.

    2017-10-01

    Hot, dense, plasma structures launched from a coaxial plasma gun on the HelCat dual-source plasma device at the University of New Mexico drag frozen-in magnetic flux into the chamber's background magnetic field providing a rich set of dynamics to study magnetic turbulence, force-free magnetic spheromaks, shocks, as well as CME-like dynamics possibly relevant to the solar corona. Vector magnetic field data from an eleven-tipped B-dot rake probe and images from an ultra-fast camera will be presented in comparison with ongoing MHD modeling using the 3-D MHD BATS-R-US code developed at the University of Michigan. BATS-R-US employs an adaptive mesh refinement grid (AMR) that enables the capture and resolution of shock structures and current sheets and is uniquely suited for flux-rope expansion modeling. Recent experiments show a possible magnetic Rayleigh-Taylor (MRT) instability that appears asymmetrically at the interface between launched spheromaks (bubbles) and their entraining background magnetic field. Efforts to understand this instability using in situ measurements, new chamber boundary conditions, and ultra-fast camera data will be presented. Work supported by the Army Research Office Award No. W911NF1510480.

  9. Novel Transverse Flux Machine for Vehicle Traction Applications: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Wan, Z.; Ahmed, A.; Husain, I.; Muljadi, E.

    2015-04-02

    A novel transverse flux machine topology for electric vehicle traction applications using ferrite magnets is presented in this paper. The proposed transverse flux topology utilizes novel magnet arrangements in the rotor that are similar to the Halbach array to boost flux linkage; on the stator side, cores are alternately arranged around a pair of ring windings in each phase to make use of the entire rotor flux that eliminates end windings. Analytical design considerations and finite-element methods are used for an optimized design of a scooter in-wheel motor. Simulation results from finite element analysis (FEA) show that the motor achieved comparable torque density to conventional rare-earth permanent magnet (PM) machines. This machine is a viable candidate for direct-drive applications with low cost and high torque density.

  10. Observations of magnetic flux ropes during magnetic reconnection in the Earth's magnetotail

    Directory of Open Access Journals (Sweden)

    A. L. Borg

    2012-05-01

    Full Text Available We present an investigation of magnetic flux ropes observed by the four Cluster spacecraft during periods of magnetic reconnection in the Earth's magnetotail. Using a list of 21 Cluster encounters with the reconnection process in the period 2001–2006 identified in Borg et al. (2012, we present the distribution and characteristics of the flux ropes. We find 27 flux ropes embedded in the reconnection outflows of only 11 of the 21 reconnection encounters. Reconnection processes associated with no flux rope observations were not distinguishable from those where flux ropes were observed. Only 7 of the 27 flux ropes show evidence of enhanced energetic electron flux above 50 keV, and there was no clear signature of the flux rope in the thermal particle measurements. We found no clear correlation between the flux rope core field and the prevailing IMF By direction.

  11. Studying the Formation and Evolution of Eruptive Solar Magnetic Flux Ropes

    Science.gov (United States)

    Linton, M.

    2017-12-01

    Solar magnetic eruptions are dramatic sources of solar activity, and dangerous sources of space weather hazards. Many of these eruptions take the form of magnetic flux ropes, i.e., magnetic fieldlines wrapping around a core magnetic flux tube. Investigating the processes which form these flux ropes both prior to and during eruption, and investigating their evolution after eruption, can give us a critical window into understanding the sources of and processes involved in these eruptions. This presentation will discuss modeling and observational investigations into these various phases of flux rope formation, eruption, and evolution, and will discuss how these different explorations can be used to develop a more complete picture of erupting flux rope dynamics. This work is funded by the NASA Living with a Star program.

  12. INDIAN POINT REACTOR REACTIVITY AND FLUX DISTRIBUTION MEASUREMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Batch, M. L.; Fischer, F. E.

    1963-11-15

    The reactivity of the Indian Point core was measured near zero reactivity at various shim and control rod patterns. Flux distribution measurements were also made, and the results are expressed in terms of power peaking factors and normalized detector response during rod withdrawal. (D.L.C.)

  13. Clastic sediment flux to tropical Andean lakes: records of glaciation and soil erosion

    Science.gov (United States)

    Rodbell, Donald T.; Seltzer, Geoffrey O.; Mark, Bryan G.; Smith, Jacqueline A.; Abbott, Mark B.

    2008-08-01

    We developed records of clastic sediment flux to 13 alpine lakes in Peru, Ecuador, and Bolivia, and compared these with independently dated records of regional glaciation. Our objectives are to determine whether a strong relationship exists between the extent of ice cover in the region and the rate of clastic sediment delivery to alpine lakes, and thus whether clastic sediment records serve as reliable proxies for glaciation during the late Pleistocene. We isolated the clastic component in lake sediment cores by removing the majority of the biogenic and authigenic components from the bulk sediment record, and we dated cores by a combination of radiocarbon and tephrochronology. In order to partially account for intra-basin differences in sediment focusing, bedrock erosivity, and sediment availability, we normalized each record to the weighted mean value of clastic sediment flux for each respective core. This enabled the stacking of all 13 lake records to produce a composite record that is generally representative of the tropical Andes. There is a striking similarity between the composite record of clastic sediment flux and the distribution of ˜100 cosmogenic radionuclide (CRN) exposure ages for erratics on moraine crests in the central Peruvian and northern Bolivian Andes. The extent of ice cover thus appears to be the primary variable controlling the delivery of clastic sediment to alpine lakes in the region, which bolsters the increasing use of clastic sediment flux as a proxy for the extent of ice cover in the region. The CRN moraine record and the stacked lake core composite record together indicate that the expansion of ice cover and concomitant increase in clastic sediment flux began at least 40 ka, and the local last glacial maximum (LLGM) culminated between 30 and 20 ka. A decline in clastic sediment flux that began ˜20 ka appears to mark the onset of deglaciation from the LLGM, at least one millennium prior to significant warming in high latitude regions

  14. 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

  15. Trace element fluxes during the last 100 years in sediment near a nuclear power plant

    Science.gov (United States)

    Bojórquez-Sánchez, S.; Marmolejo-Rodríguez, A. J.; Ruiz-Fernández, A. C.; Sánchez-González, A.; Sánchez-Cabeza, J. A.; Bojórquez-Leyva, H.; Pérez-Bernal, L. H.

    2017-11-01

    The Salada coastal lagoon is located in Veracruz (Mexico) near the Laguna Verde Nuclear Power Plant (LVNPP). Currently, the lagoon receives the cooling waters used in the LVNPP. To evaluate the fluxes and mobilization of trace elements due to human activities in the area, two sediment cores from the coastal flood plains of Salada Lagoon were analysed. Cores were collected using PVC tubes. Sediments cores were analysed every centimetre for dating (210Pb by alpha detector) and trace metal analysis using ICP-Mass Spectrometry. The dating of both sediment cores covers the period from 1900 to 2013, which includes the construction of the LVNPP (1970's). The Normalized Enrichment Factor shows enrichment of Ag, As and Cr in both sediment cores. These enrichments correspond to the extent of mining activity (which reached a maximum in the 1900's) and to the geological setting of the coastal zone. The profiles of the element fluxes in both sediment cores reflected the construction and operation of the LVNPP; however, the elements content did not show evidence of pollution coming from the LVNPP.

  16. Analytic function expansion nodal method for nuclear reactor core design

    International Nuclear Information System (INIS)

    Noh, Hae Man

    1995-02-01

    In most advanced nodal methods the transverse integration is commonly used to reduce the multi-dimensional diffusion equation into equivalent one- dimensional diffusion equations when derving the nodal coupling equations. But the use of the transverse integration results in some limitations. The first limitation is that the transverse leakage term which appears in the transverse integration procedure must be appropriately approximated. The second limitation is that the one-dimensional flux shapes in each spatial direction resulted from the nodal calculation are not accurate enough to be directly used in reconstructing the pinwise flux distributions. Finally the transverse leakage defined for a non-rectangular node such as a hexagonal node or a triangular node is too complicated to be easily handled and may contain non-physical singular terms of step-function and delta-function types. In this thesis, the Analytic Function Expansion Nodal (AFEN) method and its two variations : the Polynomial Expansion Nodal (PEN) method and the hybrid of the AFEN and PEN methods, have been developed to overcome the limitations of the transverse integration procedure. All of the methods solve the multidimensional diffusion equation without the transverse integration. The AFEN method which we believe is the major contribution of this study to the reactor core analysis expands the homogeneous flux distributions within a node in non-separable analytic basis functions satisfying the neutron diffusion equations at any point of the node and expresses the coefficients of the flux expansion in terms of the nodal unknowns which comprise a node-average flux, node-interface fluxes, and corner-point fluxes. Then, the nodal coupling equations composed of the neutron balance equations, the interface current continuity equations, and the corner-point leakage balance equations are solved iteratively to determine all the nodal unknowns. Since the AFEN method does not use the transverse integration in

  17. On the use of flux-adjoint condensed nuclear data for 1-group AGR kinetics

    International Nuclear Information System (INIS)

    Hutt, P.K.

    1979-03-01

    Following previous work on the differences between one and two neutron group AGR kinetics the possible advantages of flux-adjoint condensed lattice data over the simple flux condensation procedure are investigated. Analytic arguments are given for expecting flux-adjoint condensation to give a better representation of rod worth slopes and flux shape changes associated with partially rodded cores. These areas have previously been found to yield most of the one to two neutron group differences. The validity of these arguments is demonstrated comparing various calculations. (U.K.)

  18. Three dimensions transport calculations for PWR core

    International Nuclear Information System (INIS)

    Richebois, E.

    2000-01-01

    The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

  19. Fast Flux Test Facility fuel and test management: The first 10 years

    International Nuclear Information System (INIS)

    Bennett, R.A.; Bennett, C.L.; Campbell, L.R.; Dobbin, K.D.; Tang, E.L.

    1991-07-01

    Core design and fuel and test management have been performed efficiently at the Fast Flux Test Facility. No outages have been extended to adjust core loadings. Development of mixed oxide fuels for advanced liquid metal breeder reactors has been carried out successfully. In fact, the fuel performance is extraordinary. Failures have been so infrequent that further development and refinement of fuel requirements seem appropriate and could lead to a significant reduction in projected electrical busbar costs. The Fast Flux Test Facility is also involved in early metal fuel development tests and appears to be an ideal test bed for any further fuel development or refinement testing. 3 refs., 4 figs., 2 tabs

  20. 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

  1. Monitoring core barrel motion by neutron noise diagnostics

    International Nuclear Information System (INIS)

    Por, G.

    1985-08-01

    The core barrel motion is detected by ionization chambers located around the reactor vessel. The method is based on the measurement of the neutron flux fluctuations. Calculations to determine the direction and the size of the motion are discussed. The identification of core barrel motion and its connection with the error of one of the main circulating pumps in the Rheinsberg nuclear power plant are described. Core barrel motion of 10 Hz with an amplitude less than 50 μm could be diagnozed at the Paks-1 reactor using the Dutch high accuracy evaluation system. (V.N.)

  2. Reliable estimation of neutron flux in BWR reactor vessel using the tort code (2) application to neutron and gamma flux estimation

    Energy Technology Data Exchange (ETDEWEB)

    Kurosawa, M. [Toshiba Corp., Yokohama (Japan); Tsukiyama, T.; Hayashi, K. [Hitachi Engineering Co. Ltd., Hitachi-shi (Japan)

    2001-07-01

    A neutron and gamma flux distribution around the core of BWR commercial plant in Japan was calculated, using a three-dimensional transport code, TORT in DOORS32 code system. In the external of the core, the bottom of the model was at an elevation of 150 cm below the bottom of active fuel, the top of the model was at an elevation of the top of the shroud head dome and the radial part of the model was to the outside of the reactor pressure vessel. The top guide beams were modeled explicitly to obtain the neutron and gamma flux distribution both in the beams and outside beams. The each control rod guide tube was also modeled with homogeneous region which included the blade wing and poison tubes so that we could obtain the neutron and gamma flux distribution around the each control rod guide tube. The calculation model mentioned above needed very large memory size which exceeded a few decade giga-bytes. As the using the splicing/coupling method had uncertainly at the splicing/coupling boundary, in this work the calculation was performed without this splicing/coupling method. On the other hand, radioactivity data were measured for a few pieces of the top guide beam, shroud and in-core monitor guide tube in the same plant which was analyzed in the above calculation. So the calculation results were able to be compared with those measured data as benchmarking and at the end of this task, the C/M values at these measured points were obtained and calculation model using TORT was evaluated. (authors)

  3. Functional requirements for core surveillance systems

    International Nuclear Information System (INIS)

    Andersson, T.

    2000-01-01

    Operating experience at Ringhals-2 has demonstrated the feasibility of a mixed core surveillance system comprised of fixed in-core detectors combined with the original movable detector system. A small number of fixed in-core detectors provide continuous measurement of the thermal margins while the movable detectors are used mainly at start-up to verify the expected power distribution. Reactor noise diagnostics and neural networks can further improve the monitoring system. The reliability of the movable detector system can be improved by mechanical simplification. Wear and maintenance costs are lowered if the required flux-mapping frequency is reduced. Improved computer codes make the measurement uncertainties less dependent on the number of instrumented positions. A mixed system requires new types of technical specifications. (author)

  4. Some characteristics of two heterogeneous cores and their experimental confirmation

    International Nuclear Information System (INIS)

    Giese, H.; Henneges, G.; Pilate, S.

    1979-01-01

    Heterogeneous core geometries have been investigated at the zero power facility ZEBRA with the objective of developing an improved understanding of the basic physics parameters of future non-conventional fast breeder reactors. Three assemblies were investigated, comprising similar fissile volumes and numbers of in-core fertile elements. First, BZC, a Salt-and-Pepper Core with fertile elements arranged in groups of sizes ranging from about one to eight power reactor subassemblies. Second, BZD, a Single Annular Core with the fertile elements collected to form a large central island of about 95 cm diameter, equivalent to about 36 subassemblies. Third, BZD/1A, a modified version of assembly BZD with a central island diameter of about 67 cm and an additional thin breeder ring. All assemblies used 24% Pu/Pu + U enriched fuel, with a core height of about 90 cm and a core diameter of about 200 cm. Extensive sodium voiding, flux-tilt and reactivity-interaction measurements have been carried out. The results indicate that relative to homogeneous designs a Single Annular arrangement yields a more marked improvement in sodium voiding performance than a Salt-and-Pepper arrangement, at the expense, however, of an increased sensitivity of the flux distribution to local reactivity perturbations. First analysis-level calculations on sodium-void yield C/E values with a significantly wider dispersion than in previous homogeneous assemblies. (author)

  5. Fast neutrino flavor conversions near the supernova core with realistic flavor-dependent angular distributions

    Energy Technology Data Exchange (ETDEWEB)

    Dasgupta, Basudeb; Sen, Manibrata [Tata Institute of Fundamental Research, Homi Bhabha Road, Mumbai, 400005 (India); Mirizzi, Alessandro, E-mail: bdasgupta@theory.tifr.res.in, E-mail: alessandro.mirizzi@ba.infn.it, E-mail: manibrata.sen@gmail.com [Dipartimento Interateneo di Fisica ' Michelangelo Merlin' , Via Amendola 173, 70126 Bari (Italy)

    2017-02-01

    It has been recently pointed out that neutrino fluxes from a supernova can show substantial flavor conversions almost immediately above the core. Using linear stability analyses and numerical solutions of the fully nonlinear equations of motion, we perform a detailed study of these fast conversions , focussing on the region just above the supernova core. We carefully specify the instabilities for evolution in space or time, and find that neutrinos travelling towards the core make fast conversions more generic, i.e., possible for a wider range of flux ratios and angular asymmetries that produce a crossing between the zenith-angle spectra of ν {sub e} and ν-bar {sub e} . Using fluxes and angular distributions predicted by supernova simulations, we find that fast conversions can occur within tens of nanoseconds, only a few meters away from the putative neutrinospheres. If these fast flavor conversions indeed take place, they would have important implications for the supernova explosion mechanism and nucleosynthesis.

  6. The Fuzziness of Giant Planets’ Cores

    Energy Technology Data Exchange (ETDEWEB)

    Helled, Ravit [Institute for Computational Science, University of Zurich, Zurich (Switzerland); Stevenson, David [Division of Geological and Planetary Sciences, Caltech, Pasadena, CA (United States)

    2017-05-01

    Giant planets are thought to have cores in their deep interiors, and the division into a heavy-element core and hydrogen–helium envelope is applied in both formation and structure models. We show that the primordial internal structure depends on the planetary growth rate, in particular, the ratio of heavy elements accretion to gas accretion. For a wide range of likely conditions, this ratio is in one-to-one correspondence with the resulting post-accretion profile of heavy elements within the planet. This flux ratio depends sensitively on the assumed solid-surface density in the surrounding nebula. We suggest that giant planets’ cores might not be distinct from the envelope and includes some hydrogen and helium, and the deep interior can have a gradual heavy-element structure. Accordingly, Jupiter’s core may not be well defined. Accurate measurements of Jupiter’s gravitational field by Juno could put constraints on Jupiter’s core mass. However, as we suggest here, the definition of Jupiter’s core is complex, and the core’s physical properties (mass, density) depend on the actual definition of the core and on the planet’s growth history.

  7. The Fuzziness of Giant Planets’ Cores

    International Nuclear Information System (INIS)

    Helled, Ravit; Stevenson, David

    2017-01-01

    Giant planets are thought to have cores in their deep interiors, and the division into a heavy-element core and hydrogen–helium envelope is applied in both formation and structure models. We show that the primordial internal structure depends on the planetary growth rate, in particular, the ratio of heavy elements accretion to gas accretion. For a wide range of likely conditions, this ratio is in one-to-one correspondence with the resulting post-accretion profile of heavy elements within the planet. This flux ratio depends sensitively on the assumed solid-surface density in the surrounding nebula. We suggest that giant planets’ cores might not be distinct from the envelope and includes some hydrogen and helium, and the deep interior can have a gradual heavy-element structure. Accordingly, Jupiter’s core may not be well defined. Accurate measurements of Jupiter’s gravitational field by Juno could put constraints on Jupiter’s core mass. However, as we suggest here, the definition of Jupiter’s core is complex, and the core’s physical properties (mass, density) depend on the actual definition of the core and on the planet’s growth history.

  8. Optimizing a three-element core design for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    West, C.D.

    1995-01-01

    Source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5-10 times the flux, for neutron beams, of the best existing facilities. Baseline design for the reactor core, based on the ''no new inventions'' rule, was an assembly of two annular fuel elements similar to those used in the Oak Ridge and Grenoble high flux reactors, containing highly enriched U silicide particles. DOE commissioned a study of the use of medium- or low-enriched U; a three-element core design was studied as a means to provide extra volume to accommodate the additional U compound required when the fissionable 235 U has to be diluted with 238 U to reduce the enrichment. This paper describes the design and optimization of that three-element core

  9. Performance testing of a mixed TRIGA core

    Energy Technology Data Exchange (ETDEWEB)

    Schumacher, R F; Godsey, T A; Feltz, D E; Randall, J D [Texas A and M University (United States)

    1974-07-01

    The major operational problem experienced by the Nuclear Science Center Reactor at Texas A and M University is full burnup. With two shift operation caused by the high utilization of the facility, the reactor is operated more than 100 megawatt days per year. The solution chosen for this problem was conversion to FLIP fuel. Since funds were not available to load an entire FLIP core, a mixed core comprised of approximately one third FLIP fuel located in the central region was designed. The design core was loaded and went critical on July 1, 1973. The results of the following measurements on the mixed core are presented: Determination of Rod worths; Measurement of Reactivity Effects; Determination of Flux values; Measurement of Fuel temperatures; Preliminary Fuel Burnup Rate; Pulsing Calibration. (author)

  10. Effect of thermohydraulic parameter on the flux distribution and the effective multiplication factor

    International Nuclear Information System (INIS)

    Mello, J.C.; Valladares, G.L.

    1990-01-01

    The influence of two thermohydraulics parameters; the coolant flow velocity along the reactor channels and the increase of the average water temperature through the core, on the thermal flux distribution and on the effective multiplication factor, was studied in a radioisotopes production reactor. The results show that, for a fixed values of the thermohydraulics parameters reffered above, there are limits for the reactor core volume reduction for each value of the V sub(mod)/V sub(comb) ratio. These thermohydraulics conditions determine the higher termal flux value in the flux-trap and the lower value of the reactor effective multiplication factor. It is also show that there is a V sub(mod)/V sub(comb) ratio value that correspond to the higher value of the lower effective multiplication factor. These results was interpreted and comment using fundamentals concepts and relations of reactor physics. (author)

  11. Comparison of the Reactor Core Characteristics of the AHR and the OPAL

    International Nuclear Information System (INIS)

    Seo, Chul Gyo; Lee, Byung Chul; Park, C.; Chae, Hee Taek

    2008-09-01

    The AHR (Advanced HANARO research Reactor) was designed using the experiences from the design, operation and utilization of HANARO. Its neutronic performance was compared to that of the OPAL with a 20 MW power which started its operation recently in Australia. As the OPAL does not have any in-core irradiation hole, a modified core model of the AHR, in which an in-core irradiation hole was changed into a fuel channel, was used for the comparison. For a clean, unperturbed core condition with all fresh fuels in the core and no irradiation holes in the reflector region, the maximum thermal neutron flux (E n 14 n/cm 2 s and that in the OPAL reaches 3.96x10 14 n/cm 2 s in the reflector region. The maximum flux in the AHR is 10.3% higher than that in the OPAL. The thermal flux region above 4.0x10 14 n/cm 2 s is widely distributed in the reflector of the AHR, but is not observed at all in the reflector of the OPAL. The uranium loading of the AHR core is 45.7 kgU, which is 16.3% higher than the 39.3 kgU of OPAL. For a clean core state, the excess reactivity of the AHR is higher than that of the OPAL. The assembly-average discharge burnup in the OPAL is estimated to be 49.1%U-235 whereas that in the AHR is 62.4%U-235. The difference for the discharge burnup is significant. For the conditions with the same cycle length of 30 days, the number of fuel assemblies consumed in the AHR is only 3/4 that of the OPAL

  12. Comparison of the Reactor Core Characteristics of the AHR and the OPAL

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Lee, Byung Chul; Park, C.; Chae, Hee Taek

    2008-09-15

    The AHR (Advanced HANARO research Reactor) was designed using the experiences from the design, operation and utilization of HANARO. Its neutronic performance was compared to that of the OPAL with a 20 MW power which started its operation recently in Australia. As the OPAL does not have any in-core irradiation hole, a modified core model of the AHR, in which an in-core irradiation hole was changed into a fuel channel, was used for the comparison. For a clean, unperturbed core condition with all fresh fuels in the core and no irradiation holes in the reflector region, the maximum thermal neutron flux (E{sub n}<0.625 eV) in the AHR reaches 4.41x10{sup 14} n/cm{sup 2} s and that in the OPAL reaches 3.96x10{sup 14} n/cm{sup 2}s in the reflector region. The maximum flux in the AHR is 10.3% higher than that in the OPAL. The thermal flux region above 4.0x10{sup 14} n/cm{sup 2}s is widely distributed in the reflector of the AHR, but is not observed at all in the reflector of the OPAL. The uranium loading of the AHR core is 45.7 kgU, which is 16.3% higher than the 39.3 kgU of OPAL. For a clean core state, the excess reactivity of the AHR is higher than that of the OPAL. The assembly-average discharge burnup in the OPAL is estimated to be 49.1%U-235 whereas that in the AHR is 62.4%U-235. The difference for the discharge burnup is significant. For the conditions with the same cycle length of 30 days, the number of fuel assemblies consumed in the AHR is only 3/4 that of the OPAL.

  13. GNPS 18-months fuel cycles core thermal hydraulic design

    International Nuclear Information System (INIS)

    Liu Changwen; Zhou Zhou

    2002-01-01

    GNPS begins to implement the 18-month fuel cycles from the initial annual reload at cycle 9, thus the initial core thermal hydraulic design is not valid any more. The new critical heat flux (CHF) correlation, FC, which is developed by Framatome, is used in the design, and the generalized statistical methodology (GSM) instead of the initial deterministic methodology is used to determine the DNBR design limit. As the AFA 2G and AFA 3G are mixed loaded in the transition cycle, it will result that the minimum DNBR in the mixed core is less than that of AFA 3G homogenous core, the envelop mixed core DNBR penalty is given. Consequently the core physical limit for mixed core and equilibrium cycles, and the new over temperature ΔT overpower ΔT are determined

  14. 3-D flux distribution and criticality calculation of TRIGA Mark-II

    International Nuclear Information System (INIS)

    Can, B.

    1982-01-01

    In this work, the static calculation of the (I.T.U. TRIGA Mark-II) flux distribution has been made. The three dimensional, r-θ-z, representation of the core has been used. In this representation, for different configuration, the flux distribution has been calculated depending on two group theory. The thermal-hydraulics, the poisoning effects have been ignored. The calculations have been made by using the three dimensional and multigroup code CAN. (author)

  15. Design and performance of a pulse transformer based on Fe-based nanocrystalline core.

    Science.gov (United States)

    Yi, Liu; Xibo, Feng; Lin, Fuchang

    2011-08-01

    A dry-type pulse transformer based on Fe-based nanocrystalline core with a load of 0.88 nF, output voltage of more than 65 kV, and winding ratio of 46 is designed and constructed. The dynamic characteristics of Fe-based nanocrystalline core under the impulse with the pulse width of several microseconds were studied. The pulse width and incremental flux density have an important effect on the pulse permeability, so the pulse permeability is measured under a certain pulse width and incremental flux density. The minimal volume of the toroidal pulse transformer core is determined by the coupling coefficient, the capacitors of the resonant charging circuit, incremental flux density, and pulse permeability. The factors of the charging time, ratio, and energy transmission efficiency in the resonant charging circuit based on magnetic core-type pulse transformer are analyzed. Experimental results of the pulse transformer are in good agreement with the theoretical calculation. When the primary capacitor is 3.17 μF and charge voltage is 1.8 kV, a voltage across the secondary capacitor of 0.88 nF with peak value of 68.5 kV, rise time (10%-90%) of 1.80 μs is obtained.

  16. 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)

  17. Generalized helicity and its time derivative

    International Nuclear Information System (INIS)

    Jarboe, T.R.; Marklin, G.J.

    1985-01-01

    Spheromaks can be sustained against resistive decay by helicity injection because they tend to obey the minimum energy principle. This principle states that a plasma-laden magnetic configuration will relax to a state of minimum energy subject to the constraint that the magnetic helicity is conserved. Use of helicity as a constraint on the minimization of energy was first proposed by Woltjer in connection with astrophysical phenomena. Helicity does decay on the resistive diffusion time. However, if helicity is created and made to flow continuoiusly into a confinement geometry, these additional linked fluxes can relax and sustain the configuration indefinitely against the resistive decay. In this paper we will present an extension of the definition of helicity to include systems where B vector can penetrate the boundary and the penetration can be varying in time. We then discuss the sustainment of RFPs and spheromaks in terms of helicity injection

  18. A wet, heterogeneous lunar interior: Lower mantle and core dynamo evolution

    Science.gov (United States)

    Evans, A. J.; Zuber, M. T.; Weiss, B. P.; Tikoo, S. M.

    2014-05-01

    While recent analyses of lunar samples indicate the Moon had a core dynamo from at least 4.2-3.56 Ga, mantle convection models of the Moon yield inadequate heat flux at the core-mantle boundary to sustain thermal core convection for such a long time. Past investigations of lunar dynamos have focused on a generally homogeneous, relatively dry Moon, while an initial compositionally stratified mantle is the expected consequence of a postaccretionary lunar magma ocean. Furthermore, recent re-examination of Apollo samples and geophysical data suggests that the Moon contains at least some regions with high water content. Using a finite element model, we investigate the possible consequences of a heterogeneously wet, compositionally stratified interior for the evolution of the Moon. We find that a postoverturn model of mantle cumulates could result in a core heat flux sufficiently high to sustain a dynamo through 2.5 Ga and a maximum surface, dipolar magnetic field strength of less than 1 μT for a 350-km core and near ˜2 μT for a 450-km core. We find that if water was transported or retained preferentially in the deep interior, it would have played a significant role in transporting heat out of the deep interior and reducing the lower mantle temperature. Thus, water, if enriched in the lower mantle, could have influenced core dynamo timing by over 1.0 Gyr and enhanced the vigor of a lunar core dynamo. Our results demonstrate the plausibility of a convective lunar core dynamo even beyond the period currently indicated by the Apollo samples.

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

    International Nuclear Information System (INIS)

    Matsumoto, Tetsuo

    1999-01-01

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

  20. Use of the gapped bead-on-plate test to investigate hydrogen induced cracking of flux cored arc welds of a quenched and tempered steel

    International Nuclear Information System (INIS)

    Chen, Liang; Dunne, Druce; Davidson, Len

    2014-01-01

    Gapped bead-on-plate (G-BOP) testing of flux cored arc welds was conducted to assess the susceptibility to hydrogen induced cold cracking (HICC) of weld metal deposited on a high strength quenched and tempered steel. For preheat temperatures higher than 40°C, no weld metal cracking was observed using a shielding gas consisting of argon with 20% carbon dioxide. In contrast, the no-crack condition was not achieved for a shielding gas consisting of argon-5% carbon dioxide for preheat temperatures lower than 100°C. This extraordinary difference in weld metal HICC resistance indicates that, in general, the shielding gas mixture can exert a major influence on weld metal transverse cold cracking behaviour

  1. Dosimetry Characteristics of Coupled Fast-Thermal Core 'HERBE'

    International Nuclear Information System (INIS)

    Pesic, M.; Milosevic, M.; Milovanovic, S.

    1996-01-01

    The 'HERBE' is new coupled fast-thermal core, designed in 1991, at the 'RB' reactor in the 'Vinca' Institute. It is used for verification of designed oriented computer codes developed in the Institute, training and sample irradiation in fast neutron field. For the last purpose a vertical experimental channel (VCH) is placed in the central axis of the fast core. Neutron spectrum in the centre of the VCR is calculated in 44 energy groups. Space distributions of two energy group neutron flux in the 'HERBE' are measured using gold foils and converted into the neutron absorbed dose (in air and tissue) using group flux-dose conversion factors. Gamma absorption doses in the air in the centre of the VCH are measured using calibrated small ionisation chamber filled with air. Determined dose rates are related to the reactor power. The first preliminary irradiations of silicon diodes (designed for production of the neutron dosemeters) in the centre of the VCH of the 'HERBE' fast core are carried out in 1994 and 1995. This paper describes calculation methods and measurement techniques applied to determination of the irradiation performance and dosimetry characteristics of the 'HERBE' system. (author)

  2. Dynamic behavior of homogeneous and heterogeneous LMFBR core-design concepts

    International Nuclear Information System (INIS)

    Chang, Y.I.; Henryson, H. II; Orechwa, Y.; Su, S.F.; Greenman, G.; Blomquist, R.

    1981-01-01

    The emphasis is placed on obtaining an understanding of the inherent difference between homogeneous and heterogeneous core configurations regarding neutronic characteristics related to the dynamic behavior. The space-time neutronic and thermal-hydraulic behavior was analyzed in detail for various core configurations by using the FX2-TH, a two-dimensional kinetics code with thermal-hydraulic feedback. In addition, the relationship between the flux tilt and the fundamental-to-first harmonic eigenvalue separation, and the sodium void reactivity in heterogeneous cores were also sutdied

  3. Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

    International Nuclear Information System (INIS)

    Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

    1988-04-01

    As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

  4. Feasibility study of incore fission chamber application for neutron flux measurements on the NET blanket

    International Nuclear Information System (INIS)

    Bertalot, L.

    1987-01-01

    A feasibility study has been carried out on the use of in-core fission chambers as neutron diagnostic tools to perform neutron flux measurements on the blanket component of NET. The high neutron and gamma fluxes and the severe thermal-mechanical and magnetic conditions of the blanket structure have been taken into account in this analysis. Preliminary design criteria and specifications of an in-core detector are presented for NET application. A research and development programme is outlined which aims to obtain more information on the tecnological constraints arising from the severe conditions of the NET blanket

  5. A method for prompt calculation of neutron flux from measured SPND [self-powered neutron detectors] currents

    International Nuclear Information System (INIS)

    Kulacsy, K.; Lux, I.

    1997-01-01

    A new, approximate method is given to calculate the in-core flux from the current of SPNDs, with a delay of only a few seconds. The stability of this stepwise algorithm is proven to be satisfactory, and the results of tests performed both on synthetic and on real data are presented. The reconstructed flux is found to follow both steady state and transient fluxes well. (author)

  6. Flux coupling and transcriptional regulation within the metabolic network of the photosynthetic bacterium Synechocystis sp. PCC6803

    DEFF Research Database (Denmark)

    Montagud, Arnau; Zelezniak, Aleksej; Navarro, Emilio

    2011-01-01

    networks, surrounded by a stable core of pathways leading to biomass building blocks. This analysis identified potential bottlenecks for hydrogen and ethanol production. Integration of transcriptomic data with the Synechocystis flux coupling networks lead to identification of reporter flux coupling pairs...

  7. Role of magnetic flux perturbations in confinement bifurcations in TUMAN-3M

    International Nuclear Information System (INIS)

    Lebedev, S.V.; Andreiko, M.V.; Askinazi, L.G.

    2003-01-01

    Poloidal magnetic flux variations in the small tokamak TUMAN-3M allowed observation of transitions between different confinement modes. The possibility of switching on/off the ohmic H-mode by edge poloidal magnetic flux perturbations has been found. The flux perturbations were created by fast current ramp up/down or by magnetic compression/decompression produced by fast increase/decrease in the toroidal magnetic field. It was found that positive flux perturbations (current ramp-up and magnetic compression scenarios) are useful means of H-mode triggering. If a negative flux perturbation (current ramp-down or magnetic decompression) is applied, the H-mode terminated. Various mechanisms involved in the L-H and H-L transition physics in the flux perturbation experiments were analyzed. The experimental observations of the transitions between confinement modes might be understood in terms of the model of a sheared radial electric field generation, which takes into account the electron Ware drift in a perturbed longitudinal electric field. Another scenario of improved confinement was observed in the initial phase of an ohmic discharge, when change in the poloidal flux is associated with current ramp-up. Variation of the rates of current ramp-up and working gas puffing in the beginning of a discharge resulted in a fast increase in the electron temperature near the axis. The increase correlates with low m/n MHD mode growth. The observed core electron confinement improvement is apparently connected with the rate of current ramp. Deviation from the optimal rate results in disappearance of the improvement. The role of magnetic shear profile and rational magnetic surfaces in the core electron confinement improvement in the initial phase of ohmic discharges is discussed. (author)

  8. Current Sheet Structures Observed by the TESIS EUV Telescope during a Flux Rope Eruption on the Sun

    Science.gov (United States)

    Reva, A. A.; Ulyanov, A. S.; Kuzin, S. V.

    2016-11-01

    We use the TESIS EUV telescope to study the current sheet signatures observed during flux rope eruption. The special feature of the TESIS telescope was its ability to image the solar corona up to a distance of 2 {R}⊙ from the Sun’s center in the Fe 171 Å line. The Fe 171 Å line emission illuminates the magnetic field lines, and the TESIS images reveal the coronal magnetic structure at high altitudes. The analyzed coronal mass ejection (CME) had a core with a spiral—flux rope—structure. The spiral shape indicates that the flux rope radius varied along its length. The flux rope had a complex temperature structure: cold legs (70,000 K, observed in He 304 Å line) and a hotter core (0.7 MK, observed in Fe 171 Å line). Such a structure contradicts the common assumption that the CME core is a cold prominence. When the CME impulsively accelerated, a dark double Y-structure appeared below the flux rope. The Y-structure timing, location, and morphology agree with the previously performed MHD simulations of the current sheet. We interpreted the Y-structure as a hot envelope of the current sheet and hot reconnection outflows. The Y-structure had a thickness of 6.0 Mm. Its length increased over time from 79 Mm to more than 411 Mm.

  9. Neutronics calculation of RTP core

    Science.gov (United States)

    Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

    2017-01-01

    Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) 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. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

  10. Gamma flux responsive self-powered detector with a tubular emitter

    International Nuclear Information System (INIS)

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

    1982-01-01

    A gamma-sensitive flux detector comprises tubular emitter, an insulating core within the emitter and an insulating layer about the emitter, and a tubular conductive collector electrode about the insulating layer. The emitter material may be platinum, lead, bismuth, tantalum, tungsten; platinum preferred

  11. Microprocessor-based integrated LMFBR core surveillance

    International Nuclear Information System (INIS)

    Gmeiner, L.

    1984-06-01

    This report results from a joint study of KfK and INTERATOM. The aim of this study is to explore the advantages of microprocessors and microelectronics for a more sophisticated core surveillance, which is based on the integration of separate surveillance techniques. Due to new developments in microelectronics and related software an approach to LMFBR core surveillance can be conceived that combines a number of measurements into a more intelligent decision-making data processing system. The following techniques are considered to contribute essentially to an integrated core surveillance system: - subassembly state and thermal hydraulics performance monitoring, - temperature noise analysis, - acoustic core surveillance, - failure characterization and failure prediction based on DND- and cover gas signals, and - flux tilting techniques. Starting from a description of these techniques it is shown that by combination and correlation of these individual techniques a higher degree of cost-effectiveness, reliability and accuracy can be achieved. (orig./GL) [de

  12. Mathematical model and simulations of radiation fluxes from buried radionuclides

    International Nuclear Information System (INIS)

    Ahmad Saat

    1999-01-01

    A mathematical model and a simple Monte Carlo simulations were developed to predict radiation fluxes from buried radionuclides. The model and simulations were applied to measured (experimental) data. The results of the mathematical model showed good acceptable order of magnitude agreement. A good agreement was also obtained between the simple simulations and the experimental results. Thus, knowing the radionuclide distribution profiles in soil from a core sample, it can be applied to the model or simulations to estimate the radiation fluxes emerging from the soil surface. (author)

  13. Visible Spectrometer at the Compact Toroid Injection Experiment, the Sustained Spheromak Plasma Experiment and the Alcator C-Mod Tokamak for Doppler Width and Shift Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Graf, A; Howard, S; Horton, R; Hwang, D; May, M; Beiersdorfer, P; McLean, H; Terry, J

    2006-05-15

    A novel Doppler spectrometer is currently being used for ion or neutral velocity and temperature measurements on the Alcator C-Mod Tokamak. The spectrometer has an f/No. of {approx}3.1 and is appropriate for visible light (3500-6700 {angstrom}). The full width at half maximum from a line emitting calibration source has been measured to be as small as 0.4 {angstrom}. The ultimate time resolution is line brightness light limited and on the order of ms. A new photon efficient detector is being used for the setup at C-Mod. Time resolution is achieved by moving the camera during a plasma discharge in a perpendicular direction through the dispersion plane of the spectrometer causing a vertical streaking across the camera face. Initial results from C-Mod as well as previous measurements from the Compact Toroid Injection Experiment (CTIX) and the Sustained Spheromak Plasma Experiment (SSPX) are presented.

  14. Estimation of ex-core detector responses by adjoint Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J. E. [Delft Univ. of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)

    2006-07-01

    Ex-core detector responses can be efficiently calculated by combining an adjoint Monte Carlo calculation with the converged source distribution of a forward Monte Carlo calculation. As the fission source distribution from a Monte Carlo calculation is given only as a collection of discrete space positions, the coupling requires a point flux estimator for each collision in the adjoint calculation. To avoid the infinite variance problems of the point flux estimator, a next-event finite-variance point flux estimator has been applied, witch is an energy dependent form for heterogeneous media of a finite-variance estimator known from the literature. To test the effects of this combined adjoint-forward calculation a simple geometry of a homogeneous core with a reflector was adopted with a small detector in the reflector. To demonstrate the potential of the method the continuous-energy adjoint Monte Carlo technique with anisotropic scattering was implemented with energy dependent absorption and fission cross sections and constant scattering cross section. A gain in efficiency over a completely forward calculation of the detector response was obtained, which is strongly dependent on the specific system and especially the size and position of the ex-core detector and the energy range considered. Further improvements are possible. The method works without problems for small detectors, even for a point detector and a small or even zero energy range. (authors)

  15. Princeton Plasma Physics Laboratory annual report, October 1, 1983-September 30, 1984

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, C.A. (ed.)

    1984-01-01

    Progress made during this reporting period is reported for each of the following areas: (1) principal parameters achieved in experimental devices, (2) TFTR, (3) PLT, (4) PBX, (5) S-1 Spheromak, (6) advanced concepts Torus-1, (7) x-ray laser studies, (8) theory, (9) tokamak modeling, (10) reactor studies, (11) spin-polarized fusion program, (12) tokamak fusion core experiment, and (13) engineering. (MOW)

  16. Princeton Plasma Physics Laboratory annual report, October 1, 1983-September 30, 1984

    International Nuclear Information System (INIS)

    Phillips, C.A.

    1984-01-01

    Progress made during this reporting period is reported for each of the following areas: (1) principal parameters achieved in experimental devices, (2) TFTR, (3) PLT, (4) PBX, (5) S-1 Spheromak, (6) advanced concepts Torus-1, (7) x-ray laser studies, (8) theory, (9) tokamak modeling, (10) reactor studies, (11) spin-polarized fusion program, (12) tokamak fusion core experiment, and (13) engineering

  17. Accelerator-driven sub-critical research facility with low-enriched fuel in lead matrix: Neutron flux calculation

    Directory of Open Access Journals (Sweden)

    Avramović Ivana

    2007-01-01

    Full Text Available The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF being developed over the last couple of years at the Vinča Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step. The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield for the selected sub-critical core cells for both beams have also been presented in this paper.

  18. Evaluation of upward heat flux in ex-vessel molten core heat transfer using MELCOR

    International Nuclear Information System (INIS)

    Park, S.Y.; Park, J.H.; Kim, S.D.; Kim, D.H.; Kim, H.D.

    2000-01-01

    The purpose of this study is to share experiences of MELCOR application to resolve the molten corium-concrete interaction (MCCI) issue in the Korea Next Generation Reactor (KNGR). In the evaluation of concrete erosion, the heat transfer modeling from the molten corium internal to the corium pool surface is very important and uncertain. MELCOR employs Kutateladze or Greene's bubble-enhanced heat transfer model for the internal heat transfer. The phenomenological uncertainty is so large that the model provides several model parameters in addition to the phenomenological model for user flexibility. However, the model parameters do not work on Kutateladze correlation at the top of the molten layer. From our experience, a code modification is suggested to match the upward heat flux with the experimental results. In this analysis, minor modification was carried out to calculate heat flux from the top molten layer to corium surface, and efforts were made to find out the best value of the model parameter based on upward heat flux of MACE test M1B. Discussion also includes its application to KNGR. (author)

  19. Analysis of Neutron Flux Distribution in Rsg-Gas Reactor With U-Mo Fuels

    Directory of Open Access Journals (Sweden)

    Taswanda Taryo

    2004-01-01

    Full Text Available The use of U-Mo fuels in research reactors seems to be promising and, recently, world researchers have carried out these such activities actively. The National Nuclear Energy Agency (BATAN which owns RSG-GAS reactor available in Serpong Research Center for Atomic Energy should anticipate this trend. It is, therefore, this research work on the use of U-Mo fuels in RSG-GAS reactor should be carried out. The work was focused on the analysis of neutron flux distribution in the RSG-GAS reactor using different content of molybdenum in U-Mo fuels. To begin with, RSG-GAS reactor core model was developed and simulated into X, Y and Z dimensions. Cross section of materials based on the developed cells of standard and control fuels was then generated using WIMS-D5-B. The criticality calculations were finally carried out applying BATAN-2DIFF code. The results showed that the neutron flux distribution obtained in U-Mo-fuel-based RSG-GAS core is very similar to those achieved in the 300-gram sillicide-fuel-based RSG-GAS reactor core. Indeed, the utilization of the U-Mo RSG-GAS core can be very similar to that of the high-density sillicide reactor core and even could be better in the future.

  20. Gas core reactors for actinide transmutation and breeder applications. Annual report

    International Nuclear Information System (INIS)

    Clement, J.D.; Rust, J.H.

    1978-01-01

    This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions

  1. Compilation of neutron flux density spectra and reaction rates in different neutron fields

    International Nuclear Information System (INIS)

    Ertek, C.

    1979-07-01

    Upon the recommendation of International Working Group of Reactor Radiation Measurements (IWGRRM), the compilation of neutron flux density spectra and the reaction rates obtained by activation and fission foils in different neutron fields is presented. The neutron fields considered are as follows: 1/E; iron block; LWR core and pressure vessel; LMFBR core and blanket; CTR first wall and blanket; fission spectrum

  2. 210Pb and Mass Flux Imbalance Between the Settling Particulates and Sediments at a Sediment Trap

    Science.gov (United States)

    Wu, C.; Chung, Y.; Hung, G.

    2005-05-01

    In order to evaluate the mass balance problems between the settling particulates and the underlying sediments, sediment traps were deployed at M1 site (21o32¡¦N, 119o28¡¦E; 2948m) where a box core was taken recently. The sediment trap results were previously published (Chung et al., 2004: Continental Shelf Research). We report here the 210Pb profile measured on this box core by determining its daughter, 210Po, with alpha spectrometry, assuming the paired nuclides are in radioactive equilibrium. The box core (33 cm long, taken in June, 2004) was analyzed for the distributions of its water content, loss on ignition (LOI), and 210Pb. The water content decreases with depth and averages about 55 percent; the LOI as a measure of the total organic matter (TOM) is fairly constant at about 12 percent. The 210Pb profile shows a general exponential decrease with depth, yielding a maximum sedimentation rate of 0.26cm/y, applying a constant flux and constant sedimentation rate model. This rate translates to a mass flux of 0.31g/cm2/y or about 8.5g/m2/d with a dry bulk density of 1.2g/cm3. Based on the excess 210Pb inventory integrated over the core length, the 210Pb flux is estimated at about 19.2dpm/cm2/y or 526dpm/cm2/d assuming at steady state. The 210Pb and mass fluxes obtained from the deepest trap at M1 were only 129dpm/m2/d and 0.55g/m2/d, respectively. The large ¡excess¡" of the 210Pb and mass fluxes in the sediments over those measured from the sediment trap (4 times in 210Pb flux and 14 times in mass flux) suggests either the sedimentation rate was overestimated due to neglect of the mixing effect, and/or large additional particulates, as resuspended sediments which contain less 210Pb, have been transported laterally near the bottom from elsewhere.

  3. The influence of spatial effects on the measurement results of reactivity in 'fast disturbances' of core parameters

    International Nuclear Information System (INIS)

    Tsyganov, S.V.; Shishkov, L.K.

    2001-01-01

    The analysis of methods for the determination of reactivity revealed an essential influence of spatial effect on the measurement precision. Using of reverse point kinetic equation for reactivity meter is assumed that the average neutron flux weigh with the importance function is known at every moment of the transient. In fact, reactivity meter represent behaviour of the neutron flux only of the part of the core, so measured value of reactivity can differ from really reactivity. Three-dimensional dynamic model of the core allow to evaluate such difference. It is supposed to evaluate correction factor for the neutron flux measured at the place where ion chamber situated with the three-dimensional model NOSTRA of the WWER core. On the basis of such algorithm we propose to build module allowing the influence of spatial effects on the results of the reactivity meter to be eliminated at real time regime. This code will be incorporated into the core monitoring system 'BLOK' (SCORPIO type) which is being developed for the Kola and Rostov NPP. The report illustrates utilization of such algorithm (Authors)

  4. Diffusion piecewise homogenization via flux discontinuity ratios

    International Nuclear Information System (INIS)

    Sanchez, Richard; Dante, Giorgio; Zmijarevic, Igor

    2013-01-01

    We analyze piecewise homogenization with flux-weighted cross sections and preservation of averaged currents at the boundary of the homogenized domain. Introduction of a set of flux discontinuity ratios (FDR) that preserve reference interface currents leads to preservation of averaged region reaction rates and fluxes. We consider the class of numerical discretizations with one degree of freedom per volume and per surface and prove that when the homogenization and computing meshes are equal there is a unique solution for the FDRs which exactly preserve interface currents. For diffusion sub-meshing we introduce a Jacobian-Free Newton-Krylov method and for all cases considered obtain an 'exact' numerical solution (eight digits for the interface currents). The homogenization is completed by extending the familiar full assembly homogenization via flux discontinuity factors to the sides of regions laying on the boundary of the piecewise homogenized domain. Finally, for the familiar nodal discretization we numerically find that the FDRs obtained with no sub-mesh (nearly at no cost) can be effectively used for whole-core diffusion calculations with sub-mesh. This is not the case, however, for cell-centered finite differences. (authors)

  5. Experience with mixed cores in the IRR-1

    International Nuclear Information System (INIS)

    Gilat, J.; Hirshfeld, H.; Wiener, A.

    1985-01-01

    Over twenty mixed core configurations composed of 'old' (18 curved plate) and 'new' 23 flat plate) MTR type fuel elements were irradiated in the IRR-1 swimming pool reactor. The number of 'new' fuel elements in the core varied from one to twenty. To establish the safety of these configurations, thermohydraulic calculations were carried out to derive the maximum allowed hot channel power, determined by the onset of flow instabilities. A core is considered safe if its hot channel power, obtained from a two-dimensional neutronic calculation of power distribution in the core, does not exceed the maximum allowed value. The conservative nature of the assumptions used in the above safety evaluation procedure was verified by measurements of pressure drops vs. coolant flow rates as well as of temperature and neutron flux distributions. (author)

  6. MAGNETIZATION OF CLOUD CORES AND ENVELOPES AND OTHER OBSERVATIONAL CONSEQUENCES OF RECONNECTION DIFFUSION

    International Nuclear Information System (INIS)

    Lazarian, A.; Esquivel, A.; Crutcher, R.

    2012-01-01

    Recent observational results for magnetic fields in molecular clouds reviewed by Crutcher seem to be inconsistent with the predictions of the ambipolar diffusion theory of star formation. These include the measured decrease in mass to flux ratio between envelopes and cores, the failure to detect any self-gravitating magnetically subcritical clouds, the determination of the flat probability distribution function (PDF) of the total magnetic field strengths implying that there are many clouds with very weak magnetic fields, and the observed scaling B∝ρ 2/3 that implies gravitational contraction with weak magnetic fields. We consider the problem of magnetic field evolution in turbulent molecular clouds and discuss the process of magnetic field diffusion mediated by magnetic reconnection. For this process that we termed 'reconnection diffusion', we provide a simple physical model and explain that this process is inevitable in view of the present-day understanding of MHD turbulence. We address the issue of the expected magnetization of cores and envelopes in the process of star formation and show that reconnection diffusion provides an efficient removal of magnetic flux that depends only on the properties of MHD turbulence in the core and the envelope. We show that as the amplitude of turbulence as well as the scale of turbulent motions decrease from the envelope to the core of the cloud, the diffusion of the magnetic field is faster in the envelope. As a result, the magnetic flux trapped during the collapse in the envelope is being released faster than the flux trapped in the core, resulting in much weaker fields in envelopes than in cores, as observed. We provide simple semi-analytical model calculations which support this conclusion and qualitatively agree with the observational results. Magnetic reconnection is also consistent with the lack of subcritical self-gravitating clouds, with the observed flat PDF of field strengths, and with the scaling of field strength

  7. MAGNETIZATION OF CLOUD CORES AND ENVELOPES AND OTHER OBSERVATIONAL CONSEQUENCES OF RECONNECTION DIFFUSION

    Energy Technology Data Exchange (ETDEWEB)

    Lazarian, A. [Astronomy Department, University of Wisconsin, Madison, WI 53706 (United States); Esquivel, A. [Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, Apartado Postal 70-543, 04510 Mexico D.F. (Mexico); Crutcher, R. [Department of Astronomy, University of Illinois at Urbana-Champaign, 1002 W. Green Street, Urbana, IL 61801 (United States)

    2012-10-01

    Recent observational results for magnetic fields in molecular clouds reviewed by Crutcher seem to be inconsistent with the predictions of the ambipolar diffusion theory of star formation. These include the measured decrease in mass to flux ratio between envelopes and cores, the failure to detect any self-gravitating magnetically subcritical clouds, the determination of the flat probability distribution function (PDF) of the total magnetic field strengths implying that there are many clouds with very weak magnetic fields, and the observed scaling B{proportional_to}{rho}{sup 2/3} that implies gravitational contraction with weak magnetic fields. We consider the problem of magnetic field evolution in turbulent molecular clouds and discuss the process of magnetic field diffusion mediated by magnetic reconnection. For this process that we termed 'reconnection diffusion', we provide a simple physical model and explain that this process is inevitable in view of the present-day understanding of MHD turbulence. We address the issue of the expected magnetization of cores and envelopes in the process of star formation and show that reconnection diffusion provides an efficient removal of magnetic flux that depends only on the properties of MHD turbulence in the core and the envelope. We show that as the amplitude of turbulence as well as the scale of turbulent motions decrease from the envelope to the core of the cloud, the diffusion of the magnetic field is faster in the envelope. As a result, the magnetic flux trapped during the collapse in the envelope is being released faster than the flux trapped in the core, resulting in much weaker fields in envelopes than in cores, as observed. We provide simple semi-analytical model calculations which support this conclusion and qualitatively agree with the observational results. Magnetic reconnection is also consistent with the lack of subcritical self-gravitating clouds, with the observed flat PDF of field strengths, and

  8. Simulating Neutronic Core Parameters in a Research and Test Reactor

    International Nuclear Information System (INIS)

    Selim, H.K.; Amin, E.A.; Koutb, M.E.

    2011-01-01

    The present study proposes an Artificial Neural Network (ANN) modeling technique that predicts the control rods positions in a nuclear research reactor. The neutron, flux in the core of the reactor is used as the training data for the neural network model. The data used to train and validate the network are obtained by modeling the reactor core with the neutronic calculation code: CITVAP. The type of the network used in this study is the feed forward multilayer neural network with the backpropagation algorithm. The results show that the proposed ANN has good generalization capability to estimate the control rods positions knowing neutron flux for a research and test reactor. This method can be used to predict critical control rods positions to be used for reactor operation after reload

  9. Interaction of a spheromak-like compact toroid with a high beta spherical tokamak plasma

    International Nuclear Information System (INIS)

    Hwang, D.Q.; McLean, H.S.; Baker, K.L.; Evans, R.W.; Horton, R.D.; Terry, S.D.; Howard, S.; Schmidt, G.L.

    2000-01-01

    Recent experiments using accelerated spheromak-like compact toroids (SCTs) to fuel tokamak plasmas have quantified the penetration mechanism in the low beta regime; i.e. external magnetic field pressure dominates plasma thermal pressure. However, fusion reactor designs require high beta plasma and, more importantly, the proper plasma pressure profile. Here, the effect of the plasma pressure profile on SCT penetration, specifically, the effect of diamagnetism, is addressed. It is estimated that magnetic field pressure dominates penetration even up to 50% local beta. The combination of the diamagnetic effect on the toroidal magnetic field and the strong poloidal field at the outer major radius of a spherical tokamak will result in a diamagnetic well in the total magnetic field. Therefore, the spherical tokamak is a good candidate to test the potential trapping of an SCT in a high beta diamagnetic well. The diamagnetic effects of a high beta spherical tokamak discharge (low aspect ratio) are computed. To test the penetration of an SCT into such a diamagnetic well, experiments have been conducted of SCT injection into a vacuum field structure which simulates the diamagnetic field effect of a high beta tokamak. The diamagnetic field gradient length is substantially shorter than that of the toroidal field of the tokamak, and the results show that it can still improve the penetration of the SCT. Finally, analytic results have been used to estimate the effect of plasma pressure on penetration, and the effect of plasma pressure was found to be small in comparison with the magnetic field pressure. The penetration condition for a vacuum field only is reported. To study the diamagnetic effect in a high beta plasma, additional experiments need to be carried out on a high beta spherical tokamak. (author)

  10. A study on 80 fuel assemblies core for HFETR

    International Nuclear Information System (INIS)

    Sun Shouhua; Wu Yinghua; Bu Yongxi; Liu Shuiqing; Duan Tianyuan; Zhang Liangwan; Lin Jisen

    1996-12-01

    The performance of 80 and 60 fuel assemblies cores for High Flux Engineering Test Reactor (HFETR) has been compared with theoretical analysis and operating results. These results show that the core performance of 80 fuel assemblies is the same as that of 60 fuel assemblies in the following aspects: the permission power of core, the irradiation test of materials, the transmutation doping of single crystalline silicon, the production of Mo-Tc isotopes, etc. The core of 80 fuel assemblies is more convenient in operation after 500 kw test loop installed, and in greatly raising the production of 60 Co source with high specific radioactivity and the usage of fuel. As compared to the production of 60 Co source of 60 fuel assemblies core, the benefit of 80 fuel assemblies core can increase more than 3.8 millions RMB yuan per year. (2 refs., 2 tabs.)

  11. Global calculation of PWR reactor core using the two group energy solution by the response matrix method

    International Nuclear Information System (INIS)

    Conti, C.F.S.; Watson, F.V.

    1991-01-01

    A computational code to solve a two energy group neutron diffusion problem has been developed base d on the Response Matrix Method. That method solves the global problem of PWR core, without using the cross sections homogenization process, thus it is equivalent to a pontwise core calculation. The present version of the code calculates the response matrices by the first order perturbative method and considers developments on arbitrary order Fourier series for the boundary fluxes and interior fluxes. (author)

  12. Flux mapping algorithm (FMA) for 700 MWe PHWR

    International Nuclear Information System (INIS)

    Sonavani, Manoj; Ingle, V.J.; Singhvi, P.K.; Raj, Manish; Fernando, M.P.S.; Kumar, A.N.

    2012-01-01

    For large reactor like 700 MWe PHWR effective spatial control is essential and is provided by RRS. For spatial control purpose reactor core is divided into 14 power zones. Corresponding to each zone is a light water zonal compartment. The 14 ZCCs are located in two radial planes, each containing 7 ZCCs. For each zone, power measurement is carried out using inconel (3 pitch long) self powered neutron detector (SPND) at appropriate location close to the respective ZCC. Since the zone power as obtained by the healthy zone control detector (ZCD) reading belonging to a particular zone may not correspond to its actual power because the detector per zone, measure only average fluxes but the zone extends over a large core region. Therefore accurate estimation of zone power calibration factors is required to estimate the zone powers and also to provide effective spatial power control to avoid the xenon induced spatial power oscillations in large PHWRs like 700 and 540 MWe Reactors. This accurate calculation of zone power is carried out by FMS which uses λ modes in its algorithm. Flux at any point inside the reactor can be represented in terms of the linear combination of these modes. Coefficients used in the expansion are called combining coefficient. If the readings of the detectors are known, then combining coefficients can be estimated by simple matrix operations. Once these combining coefficients are known, flux at any point inside the reactor can be found. (author)

  13. Improved core-edge tokamak transport simulations with the CORSICA 2 code

    International Nuclear Information System (INIS)

    Tarditi, A.; Cohen, R.H.; Crotinger, J.A.

    1996-01-01

    The CORSICA 2 code models the nonlinear transport between the core and the edge of a tokamak plasma. The code couples a 2D axisymmetric edge/SOL model (UEDGE) to a 1D model for the radial core transport in toroidal flux coordinates (the transport module from the CORSICA 1 code). The core density and temperature profiles are joined to the flux-surface average profiles from the 2D code sufficiently inside the magnetic separatrix, at a flux surface on which the edge profiles are approximately constant. In the present version of the code, the deuterium density and electron and ion temperatures are coupled. The electron density is determined by imposing quasi-neutrality, both in the core and in the edge. The model allows the core-edge coupling of multiple ion densities while retaining a single temperature (corresponding to the equilibration value) for the all ion species. Applications of CORSICA 2 to modeling the DIII-D tokamak are discussed. This work will focus on the simulation of the L-H transition, coupling a single ion species (deuterium) and the two (electron and ion) temperatures. These simulations will employ a new self-consistent model for the L-H transition that is being implemented in the UEDGE code. Applications to the modeling of ITER ignition scenarios are also discussed. This will involve coupling a second density species (the thermal alphas), bringing the total number of coupled variables up to four. Finally, the progress in evolving the magnetic geometry is discussed. Currently, this geometry is calculated by CORSICA's MHD equilibrium module (TEQ) at the beginning of the run and fixed thereafter. However, CORSICA 1 can evolve this geometry quasistatically, and this quasistatic treatment is being extended to include the edge/SOL geometry. Recent improvements for code speed-up are also presented

  14. Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

    2008-10-29

    The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

  15. Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

    International Nuclear Information System (INIS)

    Tukiran S; Surian Pinem; Tagor MS; Lily S; Jati Susilo

    2012-01-01

    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 newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm 2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm 2 sand length of cycle is 57 days is the most optimal of IRR. (author)

  16. Implicit Unstructured Computational Aerodynamics on Many-Integrated Core Architecture

    KAUST Repository

    Al Farhan, Mohammed A.

    2014-05-04

    This research aims to understand the performance of PETSc-FUN3D, a fully nonlinear implicit unstructured grid incompressible or compressible Euler code with origins at NASA and the U.S. DOE, on many-integrated core architecture and how a hybridprogramming paradigm (MPI+OpenMP) can exploit Intel Xeon Phi hardware with upwards of 60 cores per node and 4 threads per core. For the current contribution, we focus on strong scaling with many-integrated core hardware. In most implicit PDE-based codes, while the linear algebraic kernel is limited by the bottleneck of memory bandwidth, the flux kernel arising in control volume discretization of the conservation law residuals and the preconditioner for the Jacobian exploits the Phi hardware well.

  17. Core surface flow modelling from high-resolution secular variation

    DEFF Research Database (Denmark)

    Holme, R.; Olsen, Nils

    2006-01-01

    -flux hypothesis, but the spectrum of the SV implies that a conclusive test of frozen-flux is not possible. We parametrize the effects of diffusion as an expected misfit in the flow prediction due to departure from the frozen-flux hypothesis; at low spherical harmonic degrees, this contribution dominates...... the expected departure of the SV predictions from flow to the observed SV, while at high degrees the SV model uncertainty is dominant. We construct fine-scale core surface flows to model the SV. Flow non-uniqueness is a serious problem because the flows are sufficiently small scale to allow flow around non......-series of magnetic data and better parametrization of the external magnetic field....

  18. Compact toroids with Alfvenic flows

    International Nuclear Information System (INIS)

    Wang Zhehui; Tang, X.Z.

    2004-01-01

    The Chandrasekhar equilibria form a class of stationary ideal magnetohydrodynamics equilibria stabilized by magnetic-field-aligned Alfvenic flows. Analytic solutions of the Chandrasekhar equilibria are explicitly constructed for both field-reversed configurations and spheromaks. Favorable confinement property of nested closed flux surfaces and the ideal magnetohydrodynamic stability of the compact toroids are of interest for both magnetic trapping of high energy electrons in astrophysics and confinement of high temperature plasmas in laboratory

  19. Prediction of Flow and Temperature Distributions in a High Flux Research Reactor Using the Porous Media Approach

    Directory of Open Access Journals (Sweden)

    Shanfang Huang

    2017-01-01

    Full Text Available High thermal neutron fluxes are needed in some research reactors and for irradiation tests of materials. A High Flux Research Reactor (HFRR with an inverse flux trap-converter target structure is being developed by the Reactor Engineering Analysis Lab (REAL at Tsinghua University. This paper studies the safety of the HFRR core by full core flow and temperature calculations using the porous media approach. The thermal nonequilibrium model is used in the porous media energy equation to calculate coolant and fuel assembly temperatures separately. The calculation results show that the coolant temperature keeps increasing along the flow direction, while the fuel temperature increases first and decreases afterwards. As long as the inlet coolant mass flow rate is greater than 450 kg/s, the peak cladding temperatures in the fuel assemblies are lower than the local saturation temperatures and no boiling exists. The flow distribution in the core is homogeneous with a small flow rate variation less than 5% for different assemblies. A large recirculation zone is observed in the outlet region. Moreover, the porous media model is compared with the exact model and found to be much more efficient than a detailed simulation of all the core components.

  20. Development of computational technique for labeling magnetic flux-surfaces

    International Nuclear Information System (INIS)

    Nunami, Masanori; Kanno, Ryutaro; Satake, Shinsuke; Hayashi, Takaya; Takamaru, Hisanori

    2006-03-01

    In recent Large Helical Device (LHD) experiments, radial profiles of ion temperature, electric field, etc. are measured in the m/n=1/1 magnetic island produced by island control coils, where m is the poloidal mode number and n the toroidal mode number. When the transport of the plasma in the radial profiles is numerically analyzed, an average over a magnetic flux-surface in the island is a very useful concept to understand the transport. On averaging, a proper labeling of the flux-surfaces is necessary. In general, it is not easy to label the flux-surfaces in the magnetic field with the island, compared with the case of a magnetic field configuration having nested flux-surfaces. In the present paper, we have developed a new computational technique to label the magnetic flux-surfaces. This technique is constructed by using an optimization algorithm, which is known as an optimization method called the simulated annealing method. The flux-surfaces are discerned by using two labels: one is classification of the magnetic field structure, i.e., core, island, ergodic, and outside regions, and the other is a value of the toroidal magnetic flux. We have applied the technique to an LHD configuration with the m/n=1/1 island, and successfully obtained the discrimination of the magnetic field structure. (author)