WorldWideScience

Sample records for cores nuclear

  1. Nuclear reactor core catcher

    International Nuclear Information System (INIS)

    1977-01-01

    A nuclear reactor core catcher is described for containing debris resulting from an accident causing core meltdown and which incorporates a method of cooling the debris by the circulation of a liquid coolant. (U.K.)

  2. Nuclear core baffling apparatus

    International Nuclear Information System (INIS)

    Cooper, F.W. Jr.; Silverblatt, B.L.; Knight, C.B.; Berringer, R.T.

    1979-01-01

    An apparatus for baffling the flow of reactor coolant fluid into and about the core of a nuclear reactor is described. The apparatus includes a plurality of longitudinally aligned baffle plates with mating surfaces that allow longitudinal growth with temperature increases while alleviating both leakage through the aligned plates and stresses on the components supporting the plates

  3. Nuclear core catchers

    International Nuclear Information System (INIS)

    Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

    1976-01-01

    A receptacle is described for taking the molten fragments of a nuclear reactor during a reactor core fusion accident. The receptacle is placed under the reactor. It includes at least one receptacle for the reactor core fragments, with a dome shaped part to distribute the molten fragments and at least one outside layer of alumina bricks around the dome. The characteristic of this receptacle is that the outer layer of bricks contains neutron poison rods which pass through the bricks and protrude in relation to them [fr

  4. Nuclear reactor core flow baffling

    International Nuclear Information System (INIS)

    Berringer, R.T.

    1979-01-01

    A flow baffling arrangement is disclosed for the core of a nuclear reactor. A plurality of core formers are aligned with the grids of the core fuel assemblies such that the high pressure drop areas in the core are at the same elevations as the high pressure drop areas about the core periphery. The arrangement minimizes core bypass flow, maintains cooling of the structure surrounding the core, and allows the utilization of alternative beneficial components such as neutron reflectors positioned near the core

  5. Nuclear reactor core stabilizing arrangement

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1976-01-01

    A nuclear reactor core stabilizing arrangement is described wherein a plurality of actuators, disposed in a pattern laterally surrounding a group of elongated fuel assemblies, press against respective contiguous fuel assemblies on the periphery of the group to reduce the clearance between adjacent fuel assemblies thereby forming a more compacted, vibration resistant core structure. 7 claims, 4 drawing figures

  6. Nuclear reactor core assembly

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1978-01-01

    The object of the present invention is to provide a fast reactor core assembly design for use with a fluid coolant such as liquid sodium or carbon monoxide incorporating a method of increasing the percentage of coolant flow though the blanket elements relative to the total coolant flow through the blanket and fuel elements during shutdown conditions without using moving parts. It is claimed that deterioration due to reactor radiation or temperature conditions is avoided and ready modification or replacement is possible. (U.K.)

  7. Nuclear reactor core

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo; Ishibashi, Yoko; Mochida, Takaaki; Haikawa, Katsumasa; Yamanaka, Akihiro.

    1995-01-01

    A reactor core is radially divided into an inner region, an outer region and an outermost region. As a fuel, three kinds of fuels, namely, a high enrichment degree fuel at 3.4%, a middle enrichment degree fuel at 2.3% and a low enrichment degree at 1.1% of a fuel average enrichment degree of fission product are used. Each of the fuels is bisected to upper and lower portions at an axial center thereof. The difference of average enrichment degrees between upper and lower portions is 0.1% for the high enrichment degree fuel, 0.3% for the middle enrichment degree fuel and 0.2% for the low enrichment degree fuel. In addition, the composition of fuels in each of radial regions comprises 100% of the low enrichment degree fuels in the outermost region, 91% of the higher enrichment degree fuels and 9% of the middle enrichment degree fuels in the outer region, and 34% of the high enrichment degree fuels and 30% of the middle enrichment degree fuels in the inner region. With such a constitution, fuel economy can be improved while maintaining the thermal margin in an initially loaded reactor core of a BWR type reactor. (I.N.)

  8. Nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F

    1974-07-11

    The core of the fast neutron reactor consisting, among other components, of fuel elements enriched in plutonium is divided into modules. Each module contains a bundle of four or six elongated components (fuel elements and control rods). In the arrangement with four components, one is kept rigid while the other three are elastically yielding inclined towards the center and lean against the rigid component. In the modules with six pieces, each component is elastically yielding inclined towards a central cavity. In this way, they form a circular arc. A control rod may be placed in the cavity. In order to counteract a relative lateral movement, the outer surfaces of the components which have hexagonal cross-sections have interlocking bearing cushions. The bearing cushions consist of keyway-type ribs or grooves with the wedges or ribs gripping in the grooves of the neighbouring components. In addition, the ribs have oblique entering surfaces.

  9. Nuclear reactor core safety device

    International Nuclear Information System (INIS)

    Colgate, S.A.

    1977-01-01

    The danger of a steam explosion from a nuclear reactor core melt-down can be greatly reduced by adding a gasifying agent to the fuel that releases a large amount of gas at a predetermined pre-melt-down temperature that ruptures the bottom end of the fuel rod and blows the finely divided fuel into a residual coolant bath at the bottom of the reactor. This residual bath should be equipped with a secondary cooling loop

  10. Nuclear reactor core servicing apparatus

    International Nuclear Information System (INIS)

    Andrea, C.

    1977-01-01

    Disclosed is an improved core servicing apparatus for a nuclear reactor of the type having a reactor vessel, a vessel head having a head penetration therethrough, a removable plug adapted to fit in the head penetration, and a core of the type having an array of elongated assemblies. The improved core servicing apparatus comprises a plurality of support columns suspended from the removable plug and extending downward toward the nuclear core, rigid support means carried by each of the support columns, and a plurality of servicing means for each of the support columns for servicing a plurality of assemblies. Each of the plurality of servicing means for each of the support columns is fixedly supported in a fixed array from the rigid support means. Means are provided for rotating the rigid support means and servicing means between condensed and expanded positions. When in the condensed position, the rigid support means and servicing means lie completely within the coextensive boundaries of the plug, and when in the expanded position, some of the rigid support means and servicing means lie without the coextensive boundaries of the plug

  11. 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.)

  12. Core catcher for nuclear reactor core meltdown containment

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Bowman, F.L.

    1978-01-01

    A bed of graphite particles is placed beneath a nuclear reactor core outside the pressure vessel but within the containment building to catch the core debris in the event of failure of the emergency core cooling system. Spray cooling of the debris and graphite particles together with draining and flooding of coolant fluid of the graphite bed is provided to prevent debris slump-through to the bottom of the bed

  13. Nuclear reactor with several cores

    International Nuclear Information System (INIS)

    Swars, H.

    1977-01-01

    Several sodium-cooled cores in separate vessels with removable closures are placed in a common reactor tank. Each individual vessel is protected against the consequences of an accident in the relevant core. Maintenance devices and inlet and outlet pipes for the coolant are also arranged within the reactor tank. The individual vessels are all enclosed by coolant in a way that in case of emergency cooling or refuelling each core can be continued to be cooled by means of the coolant loops of the other cores. (HP) [de

  14. Sensors for use in nuclear reactor cores

    International Nuclear Information System (INIS)

    Brown, W.L.; Geronime, R.L.

    1978-01-01

    Sensors including radiation detectors and the like for use within the core of nuclear reactors and which are constructed in a manner to provide optimum reliability of the sensor during use are described

  15. Method for refuelling a nuclear reactor core

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This invention relates to an improved method for refuelling a nuclear reactor core inside a reactor vessel. The technique allows a substantial reduction in the refuelling time as compared with previously known methods and permits fewer out of core operations and smaller temporary storage space. (U.K.)

  16. Core baffle for nuclear reactors

    International Nuclear Information System (INIS)

    Machado, O.J.; Berringer, R.T.

    1977-01-01

    The invention concerns the design of the core of a LWR with a large number of fuel assemblies formed by fuel rods and kept in position by spacer grids. According to the invention, at the level of the spacer grids match plates are mounted with openings so the flow of coolant directed upwards will not be obstructed and a parallel bypass will be obtained in the space between the core barrel and the baffle plates. In case of an accident, this configuration reduces or avoids damage from overpressure reactions. (HP) [de

  17. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1979-01-01

    A lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  18. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1979-01-01

    A lateral restraint and control systemm for a nuclear reactor core provides an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit is composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased by an amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  19. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1978-01-01

    Disclosed is a lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  20. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  1. Core construction for nuclear reactors

    International Nuclear Information System (INIS)

    Pettinger, D.S.

    1977-01-01

    HTR core construction with prismatic graphite blocks piled into columns. The front of the blocks is concavely curved. The lines of contact of two blocks are always not vertical, i.e. the blocks of one column are supported by the blocks of neighbouring columns so that ducts are formed. Groups of three or four of these columns may additionally be arranged around a central column which has recesses in order to lock the blocks of one group together. With this arrangement, dimensional changes of the graphite blocks under operating conditions can be taken up. (DG) [de

  2. Nuclear clustering - a cluster core model study

    International Nuclear Information System (INIS)

    Paul Selvi, G.; Nandhini, N.; Balasubramaniam, M.

    2015-01-01

    Nuclear clustering, similar to other clustering phenomenon in nature is a much warranted study, since it would help us in understanding the nature of binding of the nucleons inside the nucleus, closed shell behaviour when the system is highly deformed, dynamics and structure at extremes. Several models account for the clustering phenomenon of nuclei. We present in this work, a cluster core model study of nuclear clustering in light mass nuclei

  3. Nuclear reactor core modelling in multifunctional simulators

    International Nuclear Information System (INIS)

    Puska, E.K.

    1999-01-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  4. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  5. Apparatus for controlling nuclear core debris

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    Nuclear reactor apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling.

  6. Modular core component support for nuclear reactor

    International Nuclear Information System (INIS)

    Finch, L.M.; Anthony, A.J.

    1975-01-01

    The core of a nuclear reactor is made up of a plurality of support modules for containing components such as fuel elements, reflectors and control rods. Each module includes a component support portion located above a grid plate in a low-pressure coolant zone and a coolant inlet portion disposed within a module receptacle which depends from the grid plate into a zone of high-pressure coolant. Coolant enters the module through aligned openings within the receptacle and module inlet portion and flows upward into contact with the core components. The modules are hydraulically balanced within the receptacles to prevent expulsion by the upward coolant forces. (U.S.)

  7. Core support structure for nuclear power plants

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  8. Fast breeder physics and nuclear core design

    International Nuclear Information System (INIS)

    Marth, W.; Schroeder, R.

    1983-07-01

    This report gathers the papers that have been presented on January 18/19, 1983 at a seminar ''Fast breeder physics and nuclear core design'' held at KfK. These papers cover the results obtained within about the last five years in the r+d program and give some indication, what still has to be done. To begin with, the ''tools'' of the core designer, i.e. nuclear data and neutronics codes are covered in a comprehensive way, the seminar emphasized the applications, however. First of all the accuracies obtained for the most important parameters are presented for the design of homogeneous and heterogeneous cores of about 1000 MWe, they are based on the results of critical experiments. This is followed by a survey on activities related to the KNK II reactor, i.e. calculations concerning a modification of the core as well as critical experiments done with respect to re-loads. Finally, work concerning reactivity worths of accident configurations is presented: the generation of reactivity worths for the input of safety-related calculations of a SNR 2 design, and critical experiments to investigate the requirements for the codes to be used for these calculations. These papers are accompanied by two contributions from the industrial partners. The first one deals with the requirements to nuclear design methods as seen by the reactor designer and then shows what has been achieved. The latter one presents state, trends, and methods of the SNR 2 design. The concluding remarks compare the state of the art reached within DeBeNe with international achievements. (orig.) [de

  9. Gas core nuclear rocket feasibility project

    International Nuclear Information System (INIS)

    Howe, S.D.; DeVolder, B.; Thode, L.; Zerkle, D.

    1997-09-01

    The next giant leap for mankind will be the human exploration of Mars. Almost certainly within the next thirty years, a human crew will brave the isolation, the radiation, and the lack of gravity to walk on and explore the Red planet. However, because the mission distances and duration will be hundreds of times greater than the lunar missions, a human crew will face much greater obstacles and a higher risk than those experienced during the Apollo program. A single solution to many of these obstacles is to dramatically decrease the mission duration by developing a high performance propulsion system. The gas core nuclear rocket (GCNR) has the potential to be such a system. The gas core concept relies on the use of fluid dynamic forces to create and maintain a vortex. The vortex is composed of a fissile material which will achieve criticality and produce high power levels. By radiatively coupling to the surrounding fluids, extremely high temperatures in the propellant and, thus, high specific impulses can be generated. The ship velocities enabled by such performance may allow a 9 month round trip, manned Mars mission to be considered. Alternatively, one might consider slightly longer missions in ships that are heavily shielded against the intense Galactic Cosmic Ray flux to further reduce the radiation dose to the crew. The current status of the research program at the Los Alamos National Laboratory into the gas core nuclear rocket feasibility will be discussed

  10. Nuclear reactor core and fuel element therefor

    International Nuclear Information System (INIS)

    Fortescue, P.

    1986-01-01

    This patent describes a nuclear reactor core. This core consists of vertical columns of disengageable fuel elements stacked one atop another. These columns are arranged in side-by-side relationship to form a substantially continuous horizontal array. Each of the fuel elements include a block of refractory material having relatively good thermal conductivity and neutron moderating characteristics. The block has a pair of parallel flat top and bottom end faces and sides which are substantially prependicular to the end faces. The sides of each block is aligned vertically within a vertical column, with the sides of vertically adjacent blocks. Each of the blocks contains fuel chambers, including outer rows containing only fuel chambers along the sides of the block have nuclear fuel material disposed in them. The blocks also contain vertical coolant holes which are located inside the fuel chambers in the outer rows and the fuel chambers which are not located in the outer rows with the fuel chambers and which extend axially completely through from end face to end face and form continuous vertical intracolumn coolant passageways in the reactor core. The blocks have vertical grooves extending along the sides of the blocks form interblock channels which align in groups to form continuous vertical intercolumn coolant passsageways in the reactor core. The blocks are in the form of a regular hexagonal prism with each side of the block having vertical gooves defining one half of one of the coolant interblock channels, six corner edges on the blocks have vertical groves defining one-third of an interblock channel, the vertical sides of the blocks defining planar vertical surfaces

  11. Sensors for use in nuclear reactor cores

    International Nuclear Information System (INIS)

    1980-01-01

    A neutron sensor is described for use in nuclear reactor cores which does not require external power but merely an emitter, a collector and an insulator material between the two to generate an electric current that is indicative of the intensity of the radiation. The sensor is manufactured in such a way that brazed joints or spices are avoided and the insulation material used may be of relatively low density of compaction and will center the emitter and the lead wire with respect to the outer sheath or tube without deformation or varying geometry of the center wire or emitter. (UK)

  12. Sensors for use in nuclear reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-21

    A neutron sensor is described for use in nuclear reactor cores which does not require external power but merely an emitter, a collector and an insulator material between the two to generate an electric current that is indicative of the intensity of the radiation. The sensor is manufactured in such a way that brazed joints or spices are avoided and the insulation material used may be of relatively low density of compaction and will center the emitter and the lead wire with respect to the outer sheath or tube without deformation or varying geometry of the center wire or emitter.

  13. Core of a fast neutron nuclear reactor

    International Nuclear Information System (INIS)

    Giacometti, Christian; Mougniot, J.-C.; Ravier, Jean.

    1974-01-01

    The fast neutron nuclear reactor described includes an internal area in fissile material completely enclosed in an area of fertile material forming the outside blanket. The internal fissile area is provided with housings exclusively filled with fertile material forming one or more inside blankets. In this core the internal blankets are shaped like rings vertically separating superimposed rings of fissile material. The blanket of material nearest to the periphery is circumscribed externally by a contour having an indented shape on its straight section so as to increase the contact area between this blanket and the external blanket [fr

  14. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

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

    1977-04-01

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

  15. Core Technology Development of Nuclear spin polarization

    International Nuclear Information System (INIS)

    Yoo, Byung Duk; Gwon, Sung Ok; Kwon, Duck Hee; Lee, Sung Man

    2009-12-01

    In order to study nuclear spin polarization, we need several core technologies such as laser beam source to polarize the nuclear spin, low pressured helium cell development whose surface is essential to maintain polarization otherwise most of the polarized helium relaxed in short time, development of uniform magnetic field system which is essential for reducing relaxation, efficient vacuum system, development of polarization measuring system, and development of pressure raising system about 1000 times. The purpose of this study is to develop resonable power of laser system, that is at least 5 watt, 1083 nm, 4GHz tuneable. But the limitation of this research fund enforce to develop amplifying system into 5 watt with 1 watt system utilizing laser-diod which is already we have in stock. We succeeded in getting excellent specification of fiber laser system with power of 5 watts, 2 GHz linewidth, more than 80 GHz tuneable

  16. Core access system for nuclear reactor

    International Nuclear Information System (INIS)

    Andrea, C.

    1977-01-01

    Disclosed is an improved nuclear reactor arrangement to facilitate both through-the-head instrumentation and insertion and removal of assemblies from the nuclear core. The arrangement is of the type including a reactor vessel head comprising a large rotatable cover having a plurality of circular openings therethrough, a plurality of upwardly extending nozzles mounted on the upper surface of a large cover, and a plurality of upwardly extending skirts mounted on a large cover about the periphery or boundary of the circular openings; a plurality of small plugs for each of the openings in the large cover, the plugs also having nozzles mounted on the upper surface thereof, and drive mechanisms mounted on top of some of the nozzles and having means extending therethrough into the reactor vessel, the drive mechanisms and nozzles extending above the elevation of the upwardly extending skirts

  17. Fuel assembly and nuclear reactor core

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Aoyama, Motoo; Yamashita, Jun-ichi.

    1995-01-01

    The present invention concerns a fuel assembly and a nuclear reactor core capable of improving a transmutation rate of transuranium elements while improving a residual rate of fission products. In a reactor core of a BWR type reactor to which fuel rods with transuranium elements (TRU) enriched are loaded, the enrichment degree of transuranium elements occupying in fuel materials is determined not less than 2wt%, as well as a ratio of number of atoms between hydrogen and fuel heavy metals in an average reactor core under usual operation state (H/HM) is determined not more than 3 times. In addition, a ratio of the volumes between coolant regions and fuel material regions is determined not more than 2 times. A T ratio (TRU/Pu) is lowered as the TRU enrichment degree is higher and the H/HM ratio is lower. In order to reduce the T ratio not more than 1, the TRU enrichment degree is determined as not less than 2wt%, and the H/HM ratio is determined to not more than 3 times. Accordingly, since the H/HM ratio is reduced to not more than 1, and TRU is transmuted while recycling it with plutonium, the transmutation ratio of transuranium elements can be improved while improving the residual rate of fission products. (N.H.)

  18. Design of radiation shields in nuclear reactor core

    International Nuclear Information System (INIS)

    Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

    2008-01-01

    This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

  19. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

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

    International Nuclear Information System (INIS)

    Blanco, Anibal

    2007-01-01

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

  1. Core clamping device for a nuclear reactor

    International Nuclear Information System (INIS)

    Guenther, R.W.

    1974-01-01

    The core clamping device for a fast neutron reactor includes clamps to support the fuel zone against the pressure vessel. The clamps are arranged around the circumference of the core. They consist of torsion bars arranged parallel at some distance around the core with lever arms attached to the ends whose force is directed in the opposite direction, pressing against the wall of the pressure vessel. The lever arms and pressure plates also actuated by the ends of the torsion bars transfer the stress, the pressure plates acting upon the fuel elements or fuel assemblies. Coupling between the ends of the torsion bars and the pressure plates is achieved by end carrier plates directly attached to the torsion bars and radially movable. This clamping device follows the thermal expansions of the core, allows specific elements to be disengaged in sections and saves space between the core and the neutron reflectors. (DG) [de

  2. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  3. Neutronic analysis of the ford nuclear reactor leu core

    International Nuclear Information System (INIS)

    Raza, S.S.; Hayat, T.

    1989-08-01

    Neutronic analysis of the ford nuclear reactor low enriched uranium core has been carried out to gain confidence in the com puting methodology being used for Pakistan Research Reactor-1 core conversion calculations. The computed value of the effective multiplication factor (Keff) is found to be in good agreement with that quoted by others. (author). 6 figs

  4. Supporting system for the core restraint of nuclear reactors

    International Nuclear Information System (INIS)

    Kaser, A.

    1973-01-01

    The core restraint of water cooled nuclear reactors which is needed to direct the flow of the coolant through the core can be manufactured only in a moderate wall thickness. Thus, the majority of the loads have to be transmitted to the core barrel which is more rigid. The patent refers to a system of circumferential and vertical support members most of which are free to move relatively to each other, thus reducing thermal stresses during operation. (P.K.)

  5. Emergency core cooling systems in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1981-12-01

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

  6. Nuclear waste disposal utilizing a gaseous core reactor

    Science.gov (United States)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  7. Station blackout core damage frequency in an advanced nuclear reactor

    International Nuclear Information System (INIS)

    Carvalho, Luiz Sergio de

    2004-01-01

    Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

  8. Evaluating nuclear physics inputs in core-collapse supernova models

    Science.gov (United States)

    Lentz, E.; Hix, W. R.; Baird, M. L.; Messer, O. E. B.; Mezzacappa, A.

    Core-collapse supernova models depend on the details of the nuclear and weak interaction physics inputs just as they depend on the details of the macroscopic physics (transport, hydrodynamics, etc.), numerical methods, and progenitors. We present preliminary results from our ongoing comparison studies of nuclear and weak interaction physics inputs to core collapse supernova models using the spherically-symmetric, general relativistic, neutrino radiation hydrodynamics code Agile-Boltztran. We focus on comparisons of the effects of the nuclear EoS and the effects of improving the opacities, particularly neutrino--nucleon interactions.

  9. Apparatus for controlling nuclear core debris

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    Disclosed is an apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling

  10. Device for supporting a nuclear reactor core

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The core of a light-water reactor which is enclosed in a prestressed concrete pressure vessel and held within a diffuser basket is supported by a device consisting of a cylindrical shell which surrounds the basket and is rigidly fixed to a plurality of frusto-conical skirts having concurrent axes and located substantially at right angles to the axis of the reactor core. The small base of each skirt is rigidly fixed to the shell and the large base is anchored in openings formed in the reactor vessel for the penetration of coolant inlet and outlet pipes. The top portion of the shell is secured to the top portion of the diffuser basket, a flat surface being formed on the shell at the point of connection with each frusto-conical skirt so as to ensure rigid suspension while permitting thermal expansion

  11. SMART core preliminary nuclear design-II

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong Chan; Ji, Seong Kyun; Chang, Moon Hee

    1997-06-01

    Three loading patterns for 330 MWth SMART core are constructed for 25, 33 and 29 CRDMs, and one loading pattern for larger 69-FA core with 45 CRDMs is also constructed for comparison purpose. In this study, the core consists of 57 reduced height Korean Optimized Fuel Assemblies (KOFAs) developed by KAERI. The enrichment of fuel is 4.95 w/o. As a main burnable poison, 35% B-10 enriched B{sub 4}C-Al{sub 2}O{sub 3} shim is used. To control stuck rod worth, some gadolinia bearing fuel rods are used. The U-235 enrichment of the gadolinia bearing fuel rods is 1.8 w/o as used in KOFA. All patterns return cycle length of about 3 years. Three loading patterns except 25-CRDM pattern satisfy cold shutdown condition of keff {<=} 0.99 without soluble boron. These three patterns also satisfy the refueling condition of keff {<=} 0.95. In addition to the construction of loading pattern, an editing module of MASTER PPI files for rod power history generation is developed and rod power histories are generated for 29-CRDM loading pattern. Preliminary Fq design limit is suggested as 3.71 based on KOFA design experience. (author). 9 tabs., 45 figs., 16 refs.

  12. Thermal barrier and support for nuclear reactor fuel core

    International Nuclear Information System (INIS)

    Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

    1987-01-01

    A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

  13. Spring unit especially intended for a nuclear reactor core

    International Nuclear Information System (INIS)

    Brown, S.J.; Gorholt, Wilhelm.

    1977-01-01

    This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr

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

    International Nuclear Information System (INIS)

    Zhao Qiang

    2005-01-01

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

  15. Methods and techniques of nuclear in-core fuel management

    International Nuclear Information System (INIS)

    Jong, A.J. de.

    1992-04-01

    Review of methods of nuclear in-core fuel management (the minimal critical mass problem, minimal power peaking) and calculational techniques: reactorphysical calculations (point reactivity models, continuous refueling, empirical methods, depletion perturbation theory, nodal computer programs); optimization techniques (stochastic search, linear programming, heuristic parameter optimization). (orig./HP)

  16. Temperature measurements inside nuclear reactor cores

    International Nuclear Information System (INIS)

    Tarassenko, Serge

    1969-11-01

    Non negligible errors may happen in nuclear reactor temperature measurements using magnesium oxide insulated and stainless steel sheathed micro-wire thermocouples, when these thermometric lines are placed under operational conditions typical of electrical power stations. The present work shows that this error is principally due to electrical hysteresis and polarization phenomena in the insulator subjected to the strong fields generated by common-mode voltages. These phenomena favour the unsymmetrical common-mode current flow and thus lead to the differential-mode voltage generation which is superposing on the thermoelectric hot junction potential. A calculation and an experimental approach make possible the importance of the magnesium oxide insulating characteristics, the hot junction insulation, the choice of the main circuits in the data processing equipment as well as the galvanic isolation performances and the common-mode rejection features of all the measurement circuits. A justification is thereby given for the severe conditions imposed for the acceptance of thermoelectric materials; some particular precautions to be taken are described, as well as the high performance characteristics which have to be taken into account in choosing measurement systems linked to thermometric circuits with sheathed micro-wire thermocouples. (author) [fr

  17. Status of core nuclear design technology for future fuel

    International Nuclear Information System (INIS)

    Joo, Hyung Kook; Jung, Hyung Guk; Noh, Jae Man; Kim, Yeong Il; Kim, Taek Kyum; Gil, Choong Sup; Kim, Jung Do; Kim, Young Jin; Sohn, Dong Seong

    1997-01-01

    The effective utilization of nuclear resource is more important factor to be considered in the design of next generation PWR in addition to the epochal consideration on economics and safety. Assuming that MOX fuel can be considered as one of the future fuel corresponding to the above request, the establishment of basic technology for the MOX core design has been performed : : the specification of the technical problem through the preliminary core design and nuclear characteristic analysis of MOX, the development and verification of the neutron library for lattice code, and the acquisition of data to be used for verification of lattice and core analysis codes. The following further studies will be done in future: detailed verification of library E63LIB/A, development of the spectral history effect treatment module, extension of decay chain, development of new homogenization for the MOX fuel assembly. (author). 6 refs., 7 tabs., 2 figs

  18. Thermal radiation in gas core nuclear reactors for space propulsion

    International Nuclear Information System (INIS)

    Slutz, S.A.; Gauntt, R.O.; Harms, G.A.; Latham, T.; Roman, W.; Rodgers, R.J.

    1994-01-01

    A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept. 15 refs

  19. Preliminary concept of a zero power nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  20. Preliminary concept of a zero power nuclear reactor core

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  1. Improving the calculated core stability by the core nuclear design optimization

    International Nuclear Information System (INIS)

    Partanen, P.

    1995-01-01

    Three different equilibrium core loadings for TVO II reactor have been generated in order to improve the core stability properties at uprated power level. The reactor thermal power is assumed to be uprated from 2160 MW th to 2500 MW th , which moves the operating point after a rapid pump rundown where the core stability has been calculated from 1340 MW th and 3200 kg/s to 1675 MW th and 4000 kg/s. The core has been refuelled with ABB Atom Svea-100 -fuel, which has 3,64% w/o U-235 average enrichment in the highly enriched zone. PHOENIX lattice code has been used to provide the homogenized nuclear constants. POLCA4 static core simulator has been used for core loadings and cycle simulations and RAMONA-3B program for simulating the dynamic response to the disturbance for which the stability behaviour has been evaluated. The core decay ratio has been successfully reduced from 0,83 to 0,55 mainly by reducing the power peaking factors. (orig.) (7 figs., 1 tab.)

  2. Nuclear Human Resources Development Program using Educational Core Simulator

    International Nuclear Information System (INIS)

    Choi, Yu Sun; Hong, Soon Kwan

    2015-01-01

    KHNP-CRI(Korea Hydro and Nuclear Power Co.-Central Research Institute) has redesigned the existing Core Simulator(CoSi) used as a sort of training tools for reactor engineers in operating nuclear power plant to support Nuclear Human Resources Development (NHRD) Program focusing on the nuclear department of Dalat university in Vietnam. This program has been supported by MOTIE in Korea and cooperated with KNA(Korea Nuclear Association for International Cooperation) and HYU(Hanyang University) for enhancing the nuclear human resources of potential country in consideration with Korean Nuclear Power Plant as a next candidate energy sources. KHNP-CRI has provided Edu-CoSi to Dalat University in Vietnam in order to support Nuclear Human Resources Development Program in Vietnam. Job Qualification Certificates Program in KHNP is utilized to design a training course for Vietnamese faculty and student of Dalat University. Successfully, knowhow on lecturing the ZPPT performance, training and maintaining Edu-CoSi hardware are transferred by several training courses which KHNP-CRI provides

  3. Nuclear Human Resources Development Program using Educational Core Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yu Sun; Hong, Soon Kwan [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    KHNP-CRI(Korea Hydro and Nuclear Power Co.-Central Research Institute) has redesigned the existing Core Simulator(CoSi) used as a sort of training tools for reactor engineers in operating nuclear power plant to support Nuclear Human Resources Development (NHRD) Program focusing on the nuclear department of Dalat university in Vietnam. This program has been supported by MOTIE in Korea and cooperated with KNA(Korea Nuclear Association for International Cooperation) and HYU(Hanyang University) for enhancing the nuclear human resources of potential country in consideration with Korean Nuclear Power Plant as a next candidate energy sources. KHNP-CRI has provided Edu-CoSi to Dalat University in Vietnam in order to support Nuclear Human Resources Development Program in Vietnam. Job Qualification Certificates Program in KHNP is utilized to design a training course for Vietnamese faculty and student of Dalat University. Successfully, knowhow on lecturing the ZPPT performance, training and maintaining Edu-CoSi hardware are transferred by several training courses which KHNP-CRI provides.

  4. Enhanced core monitoring system for Browns Ferry Nuclear Plant

    International Nuclear Information System (INIS)

    Lindsey, R.S.

    1980-01-01

    A system of computer hardware and software is being developed to supplement the process computers at Browns Ferry Nuclear Plant in the area of reactor core monitoring. All data stored in the process computers will be made available through a data link to an onsite minicomputer which will store and edit the data for engineering and operations personnel. Important core parameters will be effectively displayed on color graphic CRT terminals using techniques such as blinking, shading, and color coding to emphasize significant values. This data will also be made available to Tennessee Valley Authority's Chattanooga central office support groups through a data network between existing computers

  5. Hyper-heuristic applied to nuclear reactor core design

    International Nuclear Information System (INIS)

    Domingos, R P; Platt, G M

    2013-01-01

    The design of nuclear reactors gives rises to a series of optimization problems because of the need for high efficiency, availability and maintenance of security levels. Gradient-based techniques and linear programming have been applied, as well as genetic algorithms and particle swarm optimization. The nonlinearity, multimodality and lack of knowledge about the problem domain makes de choice of suitable meta-heuristic models particularly challenging. In this work we solve the optimization problem of a nuclear reactor core design through the application of an optimal sequence of meta-heuritics created automatically. This combinatorial optimization model is known as hyper-heuristic.

  6. Shock absorber in combination with a nuclear reactor core structure

    International Nuclear Information System (INIS)

    Housman, J.J.

    1976-01-01

    This invention relates to the provision of shock absorbers for use in blind control rod passages of a nuclear reactor core structure which are not subject to degradation. The shock absorber elements are made of a porous brittle carbonaceous material, a porous brittle ceramic material, or a porous brittle refractory oxide and have a void volume of between 30% and 70% of the total volume of the element for energy absorption by fracturing due to impact loading by a control rod. (UK)

  7. Real-time advanced nuclear reactor core model

    International Nuclear Information System (INIS)

    Koclas, J.; Friedman, F.; Paquette, C.; Vivier, P.

    1990-01-01

    The paper describes a multi-nodal advanced nuclear reactor core model. The model is based on application of modern equivalence theory to the solution of neutron diffusion equation in real time employing the finite differences method. The use of equivalence theory allows the application of the finite differences method to cores divided into hundreds of nodes, as opposed to the much finer divisions (in the order of ten thousands of nodes) where the unmodified method is currently applied. As a result the model can be used for modelling of the core kinetics for real time full scope training simulators. Results of benchmarks, validate the basic assumptions of the model and its applicability to real-time simulation. (orig./HP)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  9. In-core fuel management for nuclear reactor

    International Nuclear Information System (INIS)

    Ross, M.F.; Visner, S.

    1986-01-01

    This patent describes in-core fuel management for nuclear reactor in which the first cycle of a pressurized water nuclear power reactor has a multiplicity of elongated, square fuel assemblies supported side-by-side to form a generally cylindrical, stationary core consisting entirely of fresh fuel assemblies. Each assembly of the first type has a substantially similar low average fissile enrichment of at least about 1.8 weight percent U-235, each assembly of the second type having a substantially similar intermediate average fissile enrichment at least about 0.4 weight percent greater than that of the first type, and each assembly of the third type having a substantially similar high average fissile enrichment at least about 0.4 weight percent greater than that of the intermediate type, the arrangement of the low, intermediate, and high enrichment assembly types which consists of: a generally cylindrical inner core region consisting of approximately two-thirds the total assemblies in the core and forming a figurative checkerboard array having a first checkerboard component at least two-thirds of which consists of high enrichment and intermediate enrichment assemblies, at least some of the high enrichment assemblies containing fixed burnable poison shims, and a second checkerboard component consisting of assemblies other than the high enrichment type; and a generally annular outer region consisting of the remaining assemblies and including at least some but less than two-thirds of the high enrichment type assemblies

  10. The development of direct core monitoring in Nuclear Electric plc

    International Nuclear Information System (INIS)

    Curtis, R.F.; Jones, S. Reed, J.; Wickham, A.J.

    1996-01-01

    Monitoring of graphite behaviour in Nuclear Electric Magnox and AGR reactors is necessary to support operating safety cases and to ensure that reactor operation is optimized to sustain the necessary core integrity for the economic life of the reactors. The monitoring programme combines studies for pre-characterized ''installed'' samples with studies on samples trepanned from within the cores and also with studies of core and channel geometry using specially designed equipment. Nuclear Electric has two trepanning machines originally designed for Magnox-reactor work which have been used for a substantial programme over many years. They have recently been upgraded to improve sampling speed, safety and versatility - the last being demonstrated by their adaptation for a recently-won contract associated with decommissioning the Windscale piles. Radiological hazards perceived when the AGR trepanning system was designed resulted in very cumbersome equipment. This has worked well but has been inconvenient in operation. The development of a smaller and improved system for deploying the equipment is now reported. Channel dimension monitoring equipment is discussed in detail with examples of data recovered from both Magnox and AGR cores. A resolution of ± 2 of arc (tilt) and ± 0.01 mm change in diameter in attainable. It is also theoretically possible to establish brick stresses by measuring geometry changes which result from trepanning. Current development work on a revolving scanning laser rangefinder which will enable the measurement of diameters to a resolution of 0.001 mm will also be discussed. This paper also discusses non-destructive techniques for crack detection employing ultrasound or resistance networks, the use of special manipulators to deliver inspection and repair equipment and recent developments to install displacement monitors in peripheral regions of the cores, to aid the understanding of the interaction of the restraint system with the core - the region

  11. Westinghouse Nuclear Core Design Training Center - a design simulator

    International Nuclear Information System (INIS)

    Altomare, S.; Pritchett, J.; Altman, D.

    1992-01-01

    The emergence of more powerful computing technology enables nuclear design calculations to be done on workstations. This shift to workstation usage has already had a profound effect in the training area. In 1991, the Westinghouse Electric Corporation's Commercial Nuclear Fuel Division (CNFD) developed and implemented a Nuclear Core Design Training Center (CDTC), a new concept in on-the-job training. The CDTC provides controlled on-the-job training in a structured classroom environment. It alllows one trainer, with the use of a specially prepared training facility, to provide full-scope, hands-on training to many trainees at one time. Also, the CDTC system reduces the overall cycle time required to complete the total training experience while also providing the flexibility of individual training in selected modules of interest. This paper provides descriptions of the CDTC and the respective experience gained in the application of this new concept

  12. Nuclear physics: the core of matter, the fuel of stars

    International Nuclear Information System (INIS)

    Schiffer, J.P.

    1999-01-01

    Dramatic progress has been made in all branches of physics since the National Research Council's 1986 decadal survey of the field. The Physics in a New Era series explores these advances and looks ahead to future goals. The series includes assessments of the major subfields and reports on several smaller subfields, and preparation has begun on an overview volume on the unity of physics, its relationships to other fields, and its contributions to national needs. Nuclear Physics is the latest volume of the series. The book describes current activity in understanding nuclear structure and symmetries, the behavior of matter at extreme densities, the role of nuclear physics in astrophysics and cosmology, and the instrumentation and facilities used by the field. It makes recommendations on the resources needed for experimental and theoretical advances in the coming decade. Nuclear physics addresses the nature of matter making up 99.9 percent of the mass of our everyday world. It explores the nuclear reactions that fuel the stars, including our Sun, which provides the energy for all life on Earth. The field of nuclear physics encompasses some 3,000 experimental and theoretical researchers who work at universities and national laboratories across the United States, as well as the experimental facilities and infrastructure that allow these researchers to address the outstanding scientific questions facing us. This report provides an overview of the frontiers of nuclear physics as we enter the next millennium, with special attention to the state of the science in the United States.The current frontiers of nuclear physics involve fundamental and rapidly evolving issues. One is understanding the structure and behavior of strongly interacting matter in terms of its basic constituents, quarks and gluons, over a wide range of conditions - from normal nuclear matter to the dense cores of neutron stars, and to the Big Bang that was the birth of the universe. Another is to describe

  13. Device for measuring flow rate in a nuclear reactor core

    International Nuclear Information System (INIS)

    Hamano, Jiro.

    1980-01-01

    Purpose: To always calculate core flow rate automatically and accurately in BWR type nuclear power plants. Constitution: Jet pumps are provided to the recycling pump and to the inside of the pressure vessel of a nuclear reactor. The jet pumps comprise a plurality of calibrated jet pumps for forcively convecting the coolants and a plurality of not calibrated jet pumps in order to cool the heat generated in the reactor core. The difference in the pressures between the upper and the lower portions in both of the jet pumps is measured by difference pressure transducers. Further, a thermo-sensitive element is provided to measure the temperature of recycling water at the inlet of the recycling pump. The output signal from the difference pressure transducer is inputted to a process computer, calculated periodically based on predetermined calculation equations, compensated for the temperature by a recycling water temperature signal and outputted as a core flow rate signal to a recoder. The signal is also used for the power distribution calculation in the process computer and the minimum limit power ratio as the thermal limit value for the fuels is outputted. (Furukawa, Y.)

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

    International Nuclear Information System (INIS)

    Dugan, E.T.

    1976-01-01

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

  15. Nuclear detectors for in-core power-reactors

    International Nuclear Information System (INIS)

    Duchene, Jean; Verdant, Robert.

    1979-12-01

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

  16. Review of coaxial flow gas core nuclear rocket fluid mechanics

    International Nuclear Information System (INIS)

    Weinstein, H.

    1976-01-01

    In a prematurely aborted attempt to demonstrate the feasibility of using a gas core nuclear reactor as a rocket engine, NASA initiated a number of studies on the relevant fluid mechanics problems. These studies were carried out at NASA laboratories, universities and industrial research laboratories. Because of the relatively sudden termination of most of this work, a unified overview was never presented which demonstrated the accomplishments of the program and pointed out the areas where additional work was required for a full understanding of the cavity flow. This review attempts to fulfill a part of this need in two important areas

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-04-01

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

  18. Automated software analysis of nuclear core discharge data

    International Nuclear Information System (INIS)

    Larson, T.W.; Halbig, J.K.; Howell, J.A.; Eccleston, G.W.; Klosterbuer, S.F.

    1993-03-01

    Monitoring the fueling process of an on-load nuclear reactor is a full-time job for nuclear safeguarding agencies. Nuclear core discharge monitors (CDMS) can provide continuous, unattended recording of the reactor's fueling activity for later, qualitative review by a safeguards inspector. A quantitative analysis of this collected data could prove to be a great asset to inspectors because more information can be extracted from the data and the analysis time can be reduced considerably. This paper presents a prototype for an automated software analysis system capable of identifying when fuel bundle pushes occurred and monitoring the power level of the reactor. Neural network models were developed for calculating the region on the reactor face from which the fuel was discharged and predicting the burnup. These models were created and tested using actual data collected from a CDM system at an on-load reactor facility. Collectively, these automated quantitative analysis programs could help safeguarding agencies to gain a better perspective on the complete picture of the fueling activity of an on-load nuclear reactor. This type of system can provide a cost-effective solution for automated monitoring of on-load reactors significantly reducing time and effort

  19. Global physical and numerical stability of a nuclear reactor core

    International Nuclear Information System (INIS)

    Morales-Sandoval, Jaime; Hernandez-Solis, Augusto

    2005-01-01

    Low order models are used to investigate the influence of integration methods on observed power oscillations of some nuclear reactor simulators. The zero-power point reactor kinetics with six-delayed neutron precursor groups are time discretized using explicit, implicit and Crank-Nicholson methods, and the stability limit of the time mesh spacing is exactly obtained by locating their characteristic poles in the z-transform plane. These poles are the s to z mappings of the inhour equation roots and, except for one of them, they show little or no dependence on the integration method. Conditions for stable power oscillations can be also obtained by tracking when steady state output signals resulting from reactivity oscillations in the s-Laplace plane cross the imaginary axis. The dynamics of a BWR core operating at power conditions is represented by a reduced order model obtained by adding three ordinary differential equations, which can model void and Doppler reactivity feedback effects on power, and collapsing all delayed neutron precursors in one group. Void dynamics are modeled as a second order system and fuel heat transfer as a first order system. This model shows rich characteristics in terms of indicating the relative importance of different core parameters and conditions on both numerical and physical oscillations observed by large computer code simulations. A brief discussion of the influence of actual core and coolant conditions on the reduced order model is presented

  20. Core fusion accidents in nuclear power reactors. Knowledge review

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

  1. Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

    International Nuclear Information System (INIS)

    Ehud Greenspan

    2008-01-01

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

  2. Nuclear power reactor core melt accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

    2015-11-01

    For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus-FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day. Following the introduction, which describes the structure of this book and highlights the objectives of R and D on core melt accidents, this book briefly presents the design and operating principles (Chapter 2) and safety principles (Chapter 3) of the reactors currently in operation in France, as well as the main accident scenarios envisaged and studied (Chapter 4). The objective of these chapters is not to provide exhaustive information on these subjects (the reader should refer to the general reference documents listed in the corresponding chapters), but instead to provide the information needed in order to understand, firstly, the general approach adopted in France for preventing and mitigating the consequences of core melt

  3. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes.

  4. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    International Nuclear Information System (INIS)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun

    2016-01-01

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes

  5. Minaret, a deterministic neutron transport solver for nuclear core calculations

    International Nuclear Information System (INIS)

    Moller, J-Y.; Lautard, J-J.

    2011-01-01

    We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

  6. Minaret, a deterministic neutron transport solver for nuclear core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Moller, J-Y.; Lautard, J-J., E-mail: jean-yves.moller@cea.fr, E-mail: jean-jacques.lautard@cea.fr [CEA - Centre de Saclay , Gif sur Yvette (France)

    2011-07-01

    We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

  7. Reducing the risk to Mars: The gas core nuclear rocket

    International Nuclear Information System (INIS)

    Howe, S.D.; DeVolder, B.; Thode, L.; Zerkle, D.

    1998-01-01

    The next giant leap for mankind will be the human exploration of Mars. Almost certainly within the next thirty years, a human crew will brave the isolation, the radiation, and the lack of gravity to walk on and explore the Red planet. However, because the mission distances and duration will be hundreds of times greater than the lunar missions, a human crew will face much greater obstacles and a higher risk than those experienced during the Apollo program. A single solution to many of these obstacles is to dramatically decrease the mission duration by developing a high performance propulsion system. The gas-core nuclear rocket (GCNR) has the potential to be such a system. The authors have completed a comparative study of the potential impact that a GCNR could have on a manned Mars mission. The total IMLEO, transit times, and accumulated radiation dose to the crew will be compared with the NASA Design Reference Missions

  8. Nuclear many-body problem with repulsive hard core interactions

    Energy Technology Data Exchange (ETDEWEB)

    Haddad, L M

    1965-07-01

    The nuclear many-body problem is considered using the perturbation-theoretic approach of Brueckner and collaborators. This approach is outlined with particular attention paid to the graphical representation of the terms in the perturbation expansion. The problem is transformed to centre-of-mass coordinates in configuration space and difficulties involved in ordinary methods of solution of the resulting equation are discussed. A new technique, the 'reference spectrum method', devised by Bethe, Brandow and Petschek in an attempt to simplify the numerical work in presented. The basic equations are derived in this approximation and considering the repulsive hard core part of the interaction only, the effective mass is calculated at high momentum (using the same energy spectrum for both 'particle' and 'hole' states). The result of 0.87m is in agreement with that of Bethe et al. A more complete treatment using the reference spectrum method in introduced and a self-consistent set of equations is established for the reference spectrum parameters again for the case of hard core repulsions. (author)

  9. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

    The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

  10. Development of an automated core model for nuclear reactors

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1998-01-01

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input

  11. Nuclear equation of state for core-collapse supernova simulations with realistic nuclear forces

    Energy Technology Data Exchange (ETDEWEB)

    Togashi, H., E-mail: hajime.togashi@riken.jp [Nishina Center for Accelerator-Based Science, Institute of Physical and Chemical Research (RIKEN), 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Nakazato, K. [Faculty of Arts and Science, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Takehara, Y.; Yamamuro, S.; Suzuki, H. [Department of Physics, Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan); Takano, M. [Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Department of Pure and Applied Physics, Graduate School of Advanced Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan)

    2017-05-15

    A new table of the nuclear equation of state (EOS) based on realistic nuclear potentials is constructed for core-collapse supernova numerical simulations. Adopting the EOS of uniform nuclear matter constructed by two of the present authors with the cluster variational method starting from the Argonne v18 and Urbana IX nuclear potentials, the Thomas–Fermi calculation is performed to obtain the minimized free energy of a Wigner–Seitz cell in non-uniform nuclear matter. As a preparation for the Thomas–Fermi calculation, the EOS of uniform nuclear matter is modified so as to remove the effects of deuteron cluster formation in uniform matter at low densities. Mixing of alpha particles is also taken into account following the procedure used by Shen et al. (1998, 2011). The critical densities with respect to the phase transition from non-uniform to uniform phase with the present EOS are slightly higher than those with the Shen EOS at small proton fractions. The critical temperature with respect to the liquid–gas phase transition decreases with the proton fraction in a more gradual manner than in the Shen EOS. Furthermore, the mass and proton numbers of nuclides appearing in non-uniform nuclear matter with small proton fractions are larger than those of the Shen EOS. These results are consequences of the fact that the density derivative coefficient of the symmetry energy of our EOS is smaller than that of the Shen EOS.

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

    Science.gov (United States)

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

    1991-01-01

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

  13. Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

    Directory of Open Access Journals (Sweden)

    Andrea Cerutti

    Full Text Available Hepatitis C virus (HCV infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS, but no nuclear export signal (NES has yet been identified.We show here that the aa(109-133 region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126 in the identified NES or in the sequence encoding the mature core aa(1-173 significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

  14. Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

    Science.gov (United States)

    Cerutti, Andrea; Maillard, Patrick; Minisini, Rosalba; Vidalain, Pierre-Olivier; Roohvand, Farzin; Pecheur, Eve-Isabelle; Pirisi, Mario; Budkowska, Agata

    2011-01-01

    Hepatitis C virus (HCV) infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS), but no nuclear export signal (NES) has yet been identified.We show here that the aa(109-133) region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126) in the identified NES or in the sequence encoding the mature core aa(1-173) significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

  15. Core of a liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Wright, J.R.; McFall, A.

    1975-01-01

    The core of a liquid-cooled nuclear reactor, e.g. of a sodium-cooled fast reactor, is protected in such a way that the recoil wave resulting from loss of coolant in a cooling channel and caused by released gas is limited to a coolant inlet chamber of this cooling channel. The channels essentially consist of the coolant inlet chamber and a fuel chamber - with a fission gas storage plenum - through which the coolant flows. Between the two chambers, a locking device within a tube is provided offering a much larger flow resistance to the backflow of gas or coolant than in flow direction. The locking device may be a hydraulic countertorque control system, e.g. a valvular line. Other locking devices have got radially helical vanes running around an annular flow space. Furthermore, the locking device may consist of a number of needles running parallel to each other and forming a circular grid. Though it can be expanded by the forward flow - the needles are spreading - , it acts as a solid barrier for backflows. (TK) [de

  16. Nuclear Power Reactor Core Melt Accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus- FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day

  17. Localization of Vibrating Noise Sources in Nuclear Reactor Cores

    International Nuclear Information System (INIS)

    Hultqvist, Pontus

    2004-09-01

    In this thesis the possibility of locating vibrating noise sources in a nuclear reactor core from the neutron noise has been investigated using different localization methods. The influence of the vibrating noise source has been considered to be a small perturbation of the neutron flux inside the reactor. Linear perturbation theory has been used to construct the theoretical framework upon which the localization methods are based. Two different cases have been considered: one where a one-dimensional one-group model has been used and another where a two-dimensional two-energy group noise simulator has been used. In the first case only one localization method is able to determine the position with good accuracy. This localization method is based on finding roots of an equation and is sensitive to other perturbations of the neutron flux. It will therefore work better with the assistance of approximative methods that reconstruct the noise source to determine if the results are reliable or not. In the two-dimensional case the results are more promising. There are several different localization techniques that reproduce both the vibrating noise source position and the direction of vibration with enough precision. The approximate methods that reconstruct the noise source are substantially better and are able to support the root finding method in a more constructive way. By combining the methods, the results will be more reliable

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  19. Seismic response of a block-type nuclear reactor core

    International Nuclear Information System (INIS)

    Dove, R.C.; Bennett, J.G.; Merson, J.L.

    1976-05-01

    An analytical model is developed to predict seismic response of large gas-cooled reactor cores. The model is used to investigate scaling laws involved in the design of physical models of such cores, and to make parameter studies

  20. Molten core material holding device in a nuclear reactor

    International Nuclear Information System (INIS)

    Nakamura, Hisashi; Tanaka, Nobuo; Takahashi, Katsuro.

    1985-01-01

    Purpose: To improve the function of cooling to hold molten core materials in a molten core material holding device. Constitution: Plenum structures are formed into a pan-like configuration, in which liners made of metal having high melting point and relatively high heat conductivity such as tantalum, tungsten, rhenium or alloys thereof are integrally appended to hold and directly cool the molten reactor core materials. Further, a plurality of heat pipes, passing through the plenum structures, facing the cooling portion thereof to the coolants at the outer side and immersing the heating portion into the molten core materials fallen to deposit in the inner liners are disposed radially. Furthermore, heat pipes embodded in the plenum structure are disposed in the same manner below the liners. Thus, the plenum structures and the molten reactor core materials can be cooled at a high efficiency. (Seki, T.)

  1. Nuclear-fuel-cycle education: Module 5. In-core fuel management

    International Nuclear Information System (INIS)

    Levine, S.H.

    1980-07-01

    The purpose of this project was to develop a series of educational modules for use in nuclear-fuel-cycle education. These modules are designed for use in a traditional classroom setting by lectures or in a self-paced, personalized system of instruction. This module on in-core fuel management contains information on computational methods and theory; in-core fuel management using the Virginia Polytechnic Institute and State University computer modules; pressurized water reactor in-core fuel management; boiling water reactor in-core fuel management; and in-core fuel management for gas-cooled and fast reactors

  2. The effects of radiation on aluminium alloys in the core of energy nuclear reactors

    International Nuclear Information System (INIS)

    Petrossian, V.G.

    1995-01-01

    One of the attractive directions in the worldwide practice of nuclear installations is the replacement of expensive zirconium alloy with more cheap materials, particularly aluminium allo. For Heat Supply Nuclear Plants (HSNP) with approximately 473 K core temperatures, the use of heat-resistant aluminium alloys seems to be reasonable. The present work is concerned with the studies on radiation effects on aluminium alloy, and interaction between the alloy and coolant in the reactor core. (author). 2 refs., 3 figs., 1 tab

  3. 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

  4. Human Cytomegalovirus Nuclear Capsids Associate with the Core Nuclear Egress Complex and the Viral Protein Kinase pUL97.

    Science.gov (United States)

    Milbradt, Jens; Sonntag, Eric; Wagner, Sabrina; Strojan, Hanife; Wangen, Christina; Lenac Rovis, Tihana; Lisnic, Berislav; Jonjic, Stipan; Sticht, Heinrich; Britt, William J; Schlötzer-Schrehardt, Ursula; Marschall, Manfred

    2018-01-13

    The nuclear phase of herpesvirus replication is regulated through the formation of regulatory multi-component protein complexes. Viral genomic replication is followed by nuclear capsid assembly, DNA encapsidation and nuclear egress. The latter has been studied intensely pointing to the formation of a viral core nuclear egress complex (NEC) that recruits a multimeric assembly of viral and cellular factors for the reorganization of the nuclear envelope. To date, the mechanism of the association of human cytomegalovirus (HCMV) capsids with the NEC, which in turn initiates the specific steps of nuclear capsid budding, remains undefined. Here, we provide electron microscopy-based data demonstrating the association of both nuclear capsids and NEC proteins at nuclear lamina budding sites. Specifically, immunogold labelling of the core NEC constituent pUL53 and NEC-associated viral kinase pUL97 suggested an intranuclear NEC-capsid interaction. Staining patterns with phospho-specific lamin A/C antibodies are compatible with earlier postulates of targeted capsid egress at lamina-depleted areas. Important data were provided by co-immunoprecipitation and in vitro kinase analyses using lysates from HCMV-infected cells, nuclear fractions, or infectious virions. Data strongly suggest that nuclear capsids interact with pUL53 and pUL97. Combined, the findings support a refined concept of HCMV nuclear trafficking and NEC-capsid interaction.

  5. Human Cytomegalovirus Nuclear Capsids Associate with the Core Nuclear Egress Complex and the Viral Protein Kinase pUL97

    Directory of Open Access Journals (Sweden)

    Jens Milbradt

    2018-01-01

    Full Text Available The nuclear phase of herpesvirus replication is regulated through the formation of regulatory multi-component protein complexes. Viral genomic replication is followed by nuclear capsid assembly, DNA encapsidation and nuclear egress. The latter has been studied intensely pointing to the formation of a viral core nuclear egress complex (NEC that recruits a multimeric assembly of viral and cellular factors for the reorganization of the nuclear envelope. To date, the mechanism of the association of human cytomegalovirus (HCMV capsids with the NEC, which in turn initiates the specific steps of nuclear capsid budding, remains undefined. Here, we provide electron microscopy-based data demonstrating the association of both nuclear capsids and NEC proteins at nuclear lamina budding sites. Specifically, immunogold labelling of the core NEC constituent pUL53 and NEC-associated viral kinase pUL97 suggested an intranuclear NEC-capsid interaction. Staining patterns with phospho-specific lamin A/C antibodies are compatible with earlier postulates of targeted capsid egress at lamina-depleted areas. Important data were provided by co-immunoprecipitation and in vitro kinase analyses using lysates from HCMV-infected cells, nuclear fractions, or infectious virions. Data strongly suggest that nuclear capsids interact with pUL53 and pUL97. Combined, the findings support a refined concept of HCMV nuclear trafficking and NEC-capsid interaction.

  6. Analysis of ex-core detector response measured during nuclear ship Mutsu land-loaded core critical experiment

    International Nuclear Information System (INIS)

    Itagaki, M.; Abe, J.I.; Kuribayashi, K.

    1987-01-01

    There are some cases where the ex-core neutron detector response is dependent not only on the fission source distribution in a core but also on neutron absorption in the borated water reflector. For example, an unexpectedly large response variation was measured during the nuclear ship Mutsu land-loaded core critical experiment. This large response variation is caused largely by the boron concentration change associated with the change in control rod positioning during the experiment. The conventional Crump-Lee response calculation method has been modified to take into account this boron effect. The correction factor in regard to this effect has been estimated using the one-dimensional transport code ANISN. The detector response variations obtained by means of this new calculation procedure agree well with the measured values recorded during the experiment

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

    International Nuclear Information System (INIS)

    Zhang Hong

    1997-01-01

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

  8. Real-time simulation of ex-core nuclear instrumentation system

    International Nuclear Information System (INIS)

    Zhao Qiang; Zhang Zhijian; Cao Xinrong

    2005-01-01

    Real-time simulation of ex-core nuclear instrumentation system is an indispensable part of nuclear power plant (NPP) full-scope training simulator. The simulation method, which is based upon the theory of measurement, is introduced in the paper. The fitting formula between the measured data and the three-dimensional neutron flux distribution in the core is established. The fitting parameter is adjusted according to the reactor physical calculation or the experiment of power calibration. The simulation result shows that the method can simulate the ex-core neutron instrumentation system accurately in real-time and meets the needs of NPP full-scope training simulator. (authors)

  9. Gas core nuclear thermal rocket engine research and development in the former USSR

    International Nuclear Information System (INIS)

    Koehlinger, M.W.; Bennett, R.G.; Motloch, C.G.; Gurfink, M.M.

    1992-09-01

    Beginning in 1957 and continuing into the mid 1970s, the USSR conducted an extensive investigation into the use of both solid and gas core nuclear thermal rocket engines for space missions. During this time the scientific and engineering. problems associated with the development of a solid core engine were resolved. At the same time research was undertaken on a gas core engine, and some of the basic engineering problems associated with the concept were investigated. At the conclusion of the program, the basic principles of the solid core concept were established. However, a prototype solid core engine was not built because no established mission required such an engine. For the gas core concept, some of the basic physical processes involved were studied both theoretically and experimentally. However, no simple method of conducting proof-of-principle tests in a neutron flux was devised. This report focuses primarily on the development of the. gas core concept in the former USSR. A variety of gas core engine system parameters and designs are presented, along with a summary discussion of the basic physical principles and limitations involved in their design. The parallel development of the solid core concept is briefly described to provide an overall perspective of the magnitude of the nuclear thermal propulsion program and a technical comparison with the gas core concept

  10. Nuclear characteristics evaluation for Kyoto University Research Reactor with low-enriched uranium core

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Ken; Unesaki, Hironobu [Kyoto University Research Reactor Institute, Kumatori-cho Sennan-gun Osaka (Japan)

    2008-07-01

    A project to convert the fuel of Kyoto University Research Reactor (KUR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) is in progress as a part of RERTR program. Prior to the operation of LEU core, the nuclear characteristics of the core have been evaluated to confirm the safety operation. In the evaluation, nuclear parameters, such as the excess reactivity, shut down margin control rod worth, reactivity coefficients, were calculated, and they were compared with the safety limits. The results of evaluation show that the LEU core is able to satisfy the safety requirements for operation, i.e. all the parameters satisfy the safety limits. Consequently, it was confirmed that the LEU fuel core has the proper nuclear characteristics for the safety operation. (authors)

  11. A nuclear reactor core fuel reload optimization using Artificial-Ant-Colony Connective Networks

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Schirru, Roberto

    2005-01-01

    A Pressurized Water Reactor core must be reloaded every time the fuel burnup reaches a level when it is not possible to sustain nominal power operation. The nuclear core fuel reload optimization consists in finding a burned-up and fresh-fuel-assembly pattern that maximizes the number of full operational days. This problem is NP-hard, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Besides that, the problem is non-linear and its search space is highly discontinual and multimodal. In this work a parallel computational system based on Ant Colony System (ACS) called Artificial-Ant-Colony Networks is introduced to solve the nuclear reactor core fuel reload optimization problem. ACS is a system based on artificial agents that uses the reinforcement learning technique and was originally developed to solve the Traveling Salesman Problem, which is conceptually similar to the nuclear fuel reload problem. (author)

  12. Nuclear cardiology core syllabus of the European Association of Cardiovascular Imaging (EACVI).

    Science.gov (United States)

    Gimelli, Alessia; Neglia, Danilo; Schindler, Thomas H; Cosyns, Bernard; Lancellotti, Patrizio; Kitsiou, Anastasia

    2015-04-01

    The European Association of Cardiovascular Imaging (EACVI) Core Syllabus for Nuclear Cardiology is now available online. The syllabus lists key elements of knowledge in nuclear cardiology. It represents a framework for the development of training curricula and provides expected knowledge-based learning outcomes to the nuclear cardiology trainees. Published on behalf of the European Society of Cardiology. All rights reserved. © The Author 2015. For permissions please email: journals.permissions@oup.com.

  13. Improvements to the sodium supply system of a nuclear reactor core

    International Nuclear Information System (INIS)

    Chevallier, Rene; Marchais, Christian.

    1981-01-01

    This invention concerns an improvement to the sodium supply system of a nuclear reactor core and, in particular, concerns the area included between the outlet of the primary circulation pumps and the core proper. A simplified structure and a lightening of all this linking area between the circulation pumps and the distribution tank under the core is achieved and this results in a very significant reduction in the risks of deterioration and in a definite increase in the reliability of the reactor. The invention is therefore an improvement to the sodium supply system of the nuclear reactor core vessel with incorporated exchangers, in which the cool sodium, after passing through the primary exchangers, is collected in a ring compartment from whence it is taken up by the pumps and moved to at least one pipe reaching a distribution tank located under the reactor core [fr

  14. Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2011-01-01

    Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

  15. Solving the uncommon nuclear reactor core neutronics problems

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1983-01-01

    Calculational procedures have been implemented for solving importance and higher harmonic neutronics problems. Solutions are obtained routinely to support analysis of reactor core performance, treating up to three space coordinates with the multigroup diffusion theory approximation to neutron transport. The techniques used and some of the calculational difficulties are discussed

  16. 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.

  17. Nucleoporins as components of the nuclear pore complex core structure and Tpr as the architectural element of the nuclear basket.

    Science.gov (United States)

    Krull, Sandra; Thyberg, Johan; Björkroth, Birgitta; Rackwitz, Hans-Richard; Cordes, Volker C

    2004-09-01

    The vertebrate nuclear pore complex (NPC) is a macromolecular assembly of protein subcomplexes forming a structure of eightfold radial symmetry. The NPC core consists of globular subunits sandwiched between two coaxial ring-like structures of which the ring facing the nuclear interior is capped by a fibrous structure called the nuclear basket. By postembedding immunoelectron microscopy, we have mapped the positions of several human NPC proteins relative to the NPC core and its associated basket, including Nup93, Nup96, Nup98, Nup107, Nup153, Nup205, and the coiled coil-dominated 267-kDa protein Tpr. To further assess their contributions to NPC and basket architecture, the genes encoding Nup93, Nup96, Nup107, and Nup205 were posttranscriptionally silenced by RNA interference (RNAi) in HeLa cells, complementing recent RNAi experiments on Nup153 and Tpr. We show that Nup96 and Nup107 are core elements of the NPC proper that are essential for NPC assembly and docking of Nup153 and Tpr to the NPC. Nup93 and Nup205 are other NPC core elements that are important for long-term maintenance of NPCs but initially dispensable for the anchoring of Nup153 and Tpr. Immunogold-labeling for Nup98 also results in preferential labeling of NPC core regions, whereas Nup153 is shown to bind via its amino-terminal domain to the nuclear coaxial ring linking the NPC core structures and Tpr. The position of Tpr in turn is shown to coincide with that of the nuclear basket, with different Tpr protein domains corresponding to distinct basket segments. We propose a model in which Tpr constitutes the central architectural element that forms the scaffold of the nuclear basket.

  18. Analysis of LWR Full MOX Core Physics Experiments with Major Nuclear Data Libraries

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toru [Japan Nuclear Energy Safety Organization, Tokyo (Japan)

    2007-07-01

    Nuclear Power Engineering Corporation (NUPEC) studied high moderation full MOX cores as a part of advanced LWR core concept studies from 1994 to 2003 supported by the Ministry of Economy, Trade and Industry. In order to obtain the major physics characteristics of such advanced MOX cores, NUPEC carried out core physics experimental programs called MISTRAL and BASALA from 1996 to 2002 in the EOLE critical facility of the Cadarache Center in collaboration with CEA. NUPEC also obtained a part of experimental data of the EPICURE program that CEA had conducted for 30 % Pu recycling in French PWRs. Japan Nuclear Energy Safety Organization(JNES) established in 2003 as an incorporated administrative agency took over the NUPEC's projects for nuclear regulation and has been implementing FUBILA program that is for high burn up BWR full MOX cores. This paper presents an outline of the programs and a summary of the analysis results of the criticality of those experimental cores with major nuclear data libraries.

  19. In-core fuel management programs for nuclear power reactors

    International Nuclear Information System (INIS)

    1984-10-01

    In response to the interest shown by Member States, the IAEA organized a co-ordinated research programme to develop and make available in the open domain a set of programs to perform in-core fuel management calculations. This report summarizes the work performed in the context of the CRP. As a result of this programme, complete in-core fuel management packages for three types of reactors, namely PWR's, BWR's and PHWR are now available from the NEA Data Bank. For some reactor types, these program packages are available with three levels of sophistication ranging from simple methods for educational purposes to more comprehensive methods that can be used for reactor design and operation. In addition some operating data have been compiled to allow code validation. (author)

  20. Feasibility study on nuclear core design for soluble boron free small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Hah, Chang Joo; Ju, Cho Sung [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  1. Core safety of Indian nuclear power plants (NPPs) under extreme ...

    Indian Academy of Sciences (India)

    Of course, the accidents led to release of radioactivity due to probable melt down of reactor ... advances in technology and a better understanding of the nuclear power, the ..... PHWR system offers certain intrinsic advantages (Narora Atomic Power ...... system, Non Active Process Water System (NAPWS), Service Water Sys-.

  2. Pinning down nuclear. To the core of the matter

    International Nuclear Information System (INIS)

    Boeck, Helmut; Gerstmayr, Michael; Radde, Eileen

    2014-01-01

    The nuclear disaster in Fukushima shocked the world tremendously. The call to pull out of nuclear energy is getting louder - and more often than not by politicians trying to lure the favour of voters. Through the media there are half-truths and false information floating about the global consequences of the disaster and sensational prognoses for the future, all of which are in turn unsettling for the general public. Are the opposers to nuclear energy playing with the fear of the public or is the threat real? This book tells, in a captivating manner - authenticated with examples and incidents not known by many - what the threat for the area actually looks like. They confront the level of truth in the frightening scenarios and inform about the situation in case of emergency. Furthermore, they examine factors that preceded the disaster and broach the subject of the incredible hunger for energy, which dominates the world and continues to drive the commercial use of nuclear energy. Also the ghost of Chernobyl and its aftermath, which has been dismissed from our minds, is re-examined based on current knowledge. The book impresses with insider know-how, latest detailed knowledge, amazing facts and an entertaining narrative style.

  3. Pinning down nuclear. To the core of the matter

    Energy Technology Data Exchange (ETDEWEB)

    Boeck, Helmut; Gerstmayr, Michael [Technische Univ., Vienna (Austria); International Atomic Energy Agency, Vienna (Austria); Radde, Eileen [Nuclear Engineering Seibersdorf GmbH (Austria); International Atomic Energy Agency, Vienna (Austria)

    2014-07-01

    The nuclear disaster in Fukushima shocked the world tremendously. The call to pull out of nuclear energy is getting louder - and more often than not by politicians trying to lure the favour of voters. Through the media there are half-truths and false information floating about the global consequences of the disaster and sensational prognoses for the future, all of which are in turn unsettling for the general public. Are the opposers to nuclear energy playing with the fear of the public or is the threat real? This book tells, in a captivating manner - authenticated with examples and incidents not known by many - what the threat for the area actually looks like. They confront the level of truth in the frightening scenarios and inform about the situation in case of emergency. Furthermore, they examine factors that preceded the disaster and broach the subject of the incredible hunger for energy, which dominates the world and continues to drive the commercial use of nuclear energy. Also the ghost of Chernobyl and its aftermath, which has been dismissed from our minds, is re-examined based on current knowledge. The book impresses with insider know-how, latest detailed knowledge, amazing facts and an entertaining narrative style.

  4. Nuclear design and analysis report for KALIMER breakeven core conceptual design

    International Nuclear Information System (INIS)

    Kim, Sang Ji; Song, Hoon; Lee, Ki Bog; Chang, Jin Wook; Hong, Ser Gi; Kim, Young Gyun; Kim, Yeong Il

    2002-04-01

    During the phase 2 of LMR design technology development project, the breakeven core configuration was developed with the aim of the KALIMER self-sustaining with regard to the fissile material. The excess fissile material production is limited only to the extent of its own requirement for sustaining its planned power operation. The average breeding ratio is estimated to be 1.05 for the equilibrium core and the fissile plutonium gain per cycle is 13.9 kg. The nuclear performance characteristics as well as the reactivity coefficients have been analyzed so that the design evaluation in other activity areas can be made. In order to find out a realistic heavy metal flow evolution and investigate cycle-dependent nuclear performance parameter behaviors, the startup and transition cycle loading strategies are developed, followed by the startup core physics analysis. Driver fuel and blankets are assumed to be shuffled at the time of each reload. The startup core physics analysis has shown that the burnup reactivity swing, effective delayed neutron fraction, conversion ratio and peak linear heat generation rate at the startup core lead to an extreme of bounding physics data for safety analysis. As an outcome of this study, a whole spectrum of reactor life is first analyzed in detail for the KALIMER core. It is experienced that the startup core analysis deserves more attention than the current design practice, before the core configuration is finalized based on the equilibrium cycle analysis alone.

  5. Notes on nuclear reactor core analysis code: CITATION

    International Nuclear Information System (INIS)

    Cepraga, D.G.

    1980-01-01

    The method which has evolved over the years for making power reactor calculations is the multigroup diffusion method. The CITATION code is designed to solve multigroup neutronics problems with application of the finite-difference diffusion theory approximation to neutron transport in up to three-dimensional geometry. The first part of this paper presents information about the mathematical equations programmed along with background material and certain displays to convey the nature of some of the formulations. The results obtained with the CITATION code regarding the neutron and burnup core analysis for a typical PWR reactor are presented in the second part of this paper. (author)

  6. Nuclear reactor ex-core startup neutron detector

    International Nuclear Information System (INIS)

    Wyvill, J.R.

    1980-01-01

    A sensitive ex-core neutron detector is needed to monitor the power level of reactors during startup. The neutron detector of this invention has a photomultiplier with window resistant to radiation darkening at the input end and an electrical connector at the output end. The photomultiplier receives light signals from a neutron-responsive scintillator medium, typically a cerium-doped lithium silicate glass, that responds to neutrons after they have been thermalized by a silicone resin moderator. Enclosing and shielding the photmultiplier, the scintillator medium and the moderator is a combined lead and borated silicone resin housing

  7. Nuclear determination of saturation profiles in core plugs

    International Nuclear Information System (INIS)

    Sletsgaard, J.; Oelgaard, P.L.

    1997-01-01

    A method to determine liquid saturations in core plugs during flooding is of importance when the relative permeability and capillary pressure function are to be determined. This part of the EFP-95 project uses transmission of γ-radiation to determine these saturations. In γ-transmission measurements, the electron density of the given substance is measured. This is an advantage as compared to methods that use electric conductivity, since neither oil nor gas conducts electricity. At the moment a single 137 Cs-source is used, but a theoretical investigation of whether it is possible to determine three saturations, using two radioactive sources with different γ-energies, has been performed. Measurements were made on three core plugs. To make sure that the measurements could be reproduced, all the plugs had a point of reference, i.e. a mark so that it was possible to place the plug same way every time. Two computer programs for calculation of saturation and porosity and the experimental setup are listed. (EG)

  8. The density jump at the inner core boundary using underground nuclear explosion records

    International Nuclear Information System (INIS)

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

    2001-01-01

    This paper presents the estimation of the minimum jump value using experimental wave forms reflected from the boundary between the Earth core and mantle (PcP) and the one between the inner and outer core (PKiKP) at a distance of 6 deg. Digital seismic records of underground nuclear tests conducted at the Semipalatinsk test site in 70s by Zerenda-Vostochny-Chkalovo seismic array have been used. (author)

  9. Contribution of Anticipated Transients Without Scram (ATWS) to core melt at United States nuclear power plants

    International Nuclear Information System (INIS)

    Giachetti, R.T.

    1989-09-01

    This report looks at WASH-1400 and several other Probabilistic Risk Assessments (PRAs) and Probabilistic Safety Studies (PSSs) to determine the contribution of Anticipated Transients Without Scram (ATWS) events to the total core melt probability at eight nuclear power plants in the United States. After considering each plant individually, the results are compared from plant to plant to see if any generic conclusions regarding ATWS, or core melt in general, can be made. 8 refs., 34 tabs

  10. 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

  11. Impact of nuclear 'pasta' on neutrino transport in collapsing stellar cores

    International Nuclear Information System (INIS)

    Sonoda, Hidetaka; Watanabe, Gentaro; Sato, Katsuhiko; Takiwaki, Tomoya; Yasuoka, Kenji; Ebisuzaki, Toshikazu

    2007-01-01

    Nuclear 'pasta', nonspherical nuclei in dense matter, is predicted to occur in collapsing supernova cores. We show how pasta phases affect the neutrino transport cross section via weak neutral current using several nuclear models. This is the first calculation of the neutrino opacity of the phases with rod-like and slab-like nuclei taking account of finite temperature effects, which are well described by the quantum molecular dynamics. We also show that pasta phases can occupy 10-20% of the mass of supernova cores in the later stage of the collapse

  12. In-core gamma dosimetry by solid state nuclear track detectors

    International Nuclear Information System (INIS)

    Khan, H.A.

    1980-02-01

    Results are reported of a study undertaken to develop Solid State Nuclear Track Detectors (SSNTD) for the measurement of gamma doses in the megarad region such as those existing in and around a nuclear reactor core. The changes brought about in the track etching parameters and in the ultraviolet and infrared transmittances, have been studied for possible use as gamma dose measuring indices. Effects of various parameters in the core such as neutron flux, beta particles, water, temperature, and gamma ray spectrum have been investigated and found to have only small influence on the proposed gamma dose measuring indices

  13. Calculations and selection of a TRIGA core for the Nuclear Reactor IAN-R1

    International Nuclear Information System (INIS)

    Castiblanco, L.A.; Sarta, J.A.

    1997-01-01

    The Reactor Group used the code WIMS reduced to five groups of energy, together with the code CITATION, and evaluated four configurations for a core, according to the grid actually installed. The four configurations were taken from the two proposals presented to the Instituto de Ciencias Nucleares y Energias Alternativas by General Atomics Company. In this paper, the Authors selected the best configuration according to the performance of flux distribution and excess reactivity, for a TRIGA core to be installed in the Nuclear Reactor IAN-R1

  14. Pattern recognition of subcooled boiling in core of nuclear reactors

    International Nuclear Information System (INIS)

    Inyushev, V.V.; Sharaevskij, I.G.

    2003-01-01

    The noise signals at the outlet of main nuclear power plant technological parameter gauges (neutron flux, dynamic pressure etc.) contain important information on technical state of the equipment. In the work efficient algorithms of random process identification that after respective spectral transformation are considered as multidimensional random vectors were developed. Automated classification of these vector in the developed algorithms in realized on the base of probability function, especially of Bayes classifier. Application of classifier is based on construction of multidimensional distribution of probabilities in feature space of corresponding dimension, with the help of which random vectors-realizations of corresponding images that are subjected to automated classification are described

  15. FFTF reload core nuclear design for increased experimental capability

    International Nuclear Information System (INIS)

    Rothrock, R.B.; Nelson, J.V.; Dobbin, K.D.; Bennett, R.A.

    1976-01-01

    In anticipation of continued growth in the FTR experimental irradiations program, the enrichments for the next batches of reload driver fuel to be manufactured have been increased to provide a substantially enlarged experimental reactivity allowance. The enrichments for these fuel assemblies, termed ''Cores 3 and 4,'' were selected to meet the following objectives and constraints: (1) maintain a reactor power capability of 400 MW (based on an evaluation of driver fuel centerline melting probability at 15 percent overpower); (2) provide a peak neutron flux of nominally 7 x 10 15 n/cm 2 -sec, with a minimum acceptable value of 95 percent of this (i.e., 6.65 x 10 15 n/cm 2 -sec); and (3) provide the maximum experimental reactivity allowance that is consistent with the above constraints

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    1999-03-01

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

  18. Application of gaseous core reactors for transmutation of nuclear waste

    Science.gov (United States)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

  19. Sustaining Nuclear Safety: Upholding the Core Regulatory Values

    International Nuclear Information System (INIS)

    Kumar, S.

    2016-01-01

    Nuclear Energy and management of safety therein, has a somewhat distinct streak in that from its early days it has had the privilege of being shaped and supervised by the eminent scientists and engineers, in fact it owes its very origin to them. This unique engagement has resulted in culmination of the several safety elements like defence-in-depth in the form of multiple safety layers, redundancy, diversity and physical separation of components, protection against single failures as well as common cause failures right at the beginning of designing a nuclear reactor. The fundamental principles followed by regulators across the globe have many similarities such as, creation of an organization which has a conflict-free primary responsibility of safety supervision, laying down the safety criteria and requirements for the respective industry and developing and using various tools and regulatory methodology to ensure adherence to the laid down regulatory requirements. Yet the regulatory regimes in different States have evolved differently and therefore, has certain attributes which are unique to these and confer on them their identity.

  20. Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

    International Nuclear Information System (INIS)

    Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

    1980-01-01

    Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

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

    International Nuclear Information System (INIS)

    Jung Kun Lee; Byoung Sung Han

    1997-01-01

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

  2. A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Schirru, Roberto; Carvalho da Silva, Fernando; Medeiros, Jose Antonio Carlos Canedo

    2008-01-01

    The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem

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

    International Nuclear Information System (INIS)

    Li Dongsheng

    2005-10-01

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

  4. A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: alanmmlima@yahoo.com.br; Schirru, Roberto [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: schirru@lmp.ufrj.br; Carvalho da Silva, Fernando [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: fernando@con.ufrj.br; Medeiros, Jose Antonio Carlos Canedo [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: canedo@lmp.ufrj.br

    2008-09-15

    The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem.

  5. NUCORE - A system for nuclear structure calculations with cluster-core models

    International Nuclear Information System (INIS)

    Heras, C.A.; Abecasis, S.M.

    1982-01-01

    Calculation of nuclear energy levels and their electromagnetic properties, modelling the nucleus as a cluster of a few particles and/or holes interacting with a core which in turn is modelled as a quadrupole vibrator (cluster-phonon model). The members of the cluster interact via quadrupole-quadrupole and pairing forces. (orig.)

  6. A new nodal kinetics method for analyzing fast control rod motions in nuclear reactor cores

    International Nuclear Information System (INIS)

    Kaya, S.; Yavuz, H.

    2001-01-01

    A new nodal kinetics approach is developed for analyzing large reactivity accidents in nuclear reactor cores. This method shows promising that it has capability of inspecting promt criticality transients and it gives comparable results with respect to those of other techniques. (orig.)

  7. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Dong Hyuk; Shim, Hyung Jin; Kim, Chang Hyo [Seoul National University, Seoul (Korea, Republic of)

    2014-06-15

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

  9. In-core nuclear fuel management optimization of VVER1000 using perturbation theory

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser

    2011-01-01

    In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain fuel integrity. Because of the numerous possible patterns of the fuel assemblies in the reactor core, finding the best configuration is so important and complex. Different methods for optimization of fuel loading pattern in the core have been introduced so far. In this study, a software is programmed in C ⧣ language to find an order of the fuel loading pattern of the VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process lunches by considering the initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. It shall be noticed that the designed algorithm is performed by just shuffling the fuel assemblies. The obtained results by employing the mentioned method on a typical reactor reveal that this method has a high precision in achieving a pattern with an allowable radial power peaking factor. (author)

  10. Rules for design of nuclear graphite core components - some considerations and approaches

    International Nuclear Information System (INIS)

    Svalbonas, V.; Stilwell, T.C.; Zudans, Z.

    1978-01-01

    The use of graphite as a structural element presents unusual problems both for the designer and stress analysist. When the structure happens to be a nuclear reactor core, these problems are significantly magnified both by the environment and the attendant safety requirements. In the high temperature gas reactor (HTGR) core a large number of elements are constructed of nuclear graphite. This paper discusses the attendant difficulties, and presents some approaches, for ASME code safety-consistent design and analysis. The statistical scatter of material properties, which complicates even the definitions of allowable stress, as well as the brittle, anisotropic, inhomogeneous nature of the graphite was considered. The study of this subject was undertaken under contract to the U.S. Nuclear Regulatory Commission. (Auth.)

  11. Phase diagram of nuclear 'pasta' and its uncertainties in supernova cores

    International Nuclear Information System (INIS)

    Sonoda, Hidetaka; Watanabe, Gentaro; Sato, Katsuhiko; Yasuoka, Kenji; Ebisuzaki, Toshikazu

    2008-01-01

    We examine the model dependence of the phase diagram of inhomogeneous nulcear matter in supernova cores using the quantum molecular dynamics (QMD). Inhomogeneous matter includes crystallized matter with nonspherical nuclei--''pasta'' phases--and the liquid-gas phase-separating nuclear matter. Major differences between the phase diagrams of the QMD models can be explained by the energy of pure neutron matter at low densities and the saturation density of asymmetric nuclear matter. We show the density dependence of the symmetry energy is also useful to understand uncertainties of the phase diagram. We point out that, for typical nuclear models, the mass fraction of the pasta phases in the later stage of the collapsing cores is higher than 10-20%

  12. Protein kinases responsible for the phosphorylation of the nuclear egress core complex of human cytomegalovirus.

    Science.gov (United States)

    Sonntag, Eric; Milbradt, Jens; Svrlanska, Adriana; Strojan, Hanife; Häge, Sigrun; Kraut, Alexandra; Hesse, Anne-Marie; Amin, Bushra; Sonnewald, Uwe; Couté, Yohann; Marschall, Manfred

    2017-10-01

    Nuclear egress of herpesvirus capsids is mediated by a multi-component nuclear egress complex (NEC) assembled by a heterodimer of two essential viral core egress proteins. In the case of human cytomegalovirus (HCMV), this core NEC is defined by the interaction between the membrane-anchored pUL50 and its nuclear cofactor, pUL53. NEC protein phosphorylation is considered to be an important regulatory step, so this study focused on the respective role of viral and cellular protein kinases. Multiply phosphorylated pUL50 varieties were detected by Western blot and Phos-tag analyses as resulting from both viral and cellular kinase activities. In vitro kinase analyses demonstrated that pUL50 is a substrate of both PKCα and CDK1, while pUL53 can also be moderately phosphorylated by CDK1. The use of kinase inhibitors further illustrated the importance of distinct kinases for core NEC phosphorylation. Importantly, mass spectrometry-based proteomic analyses identified five major and nine minor sites of pUL50 phosphorylation. The functional relevance of core NEC phosphorylation was confirmed by various experimental settings, including kinase knock-down/knock-out and confocal imaging, in which it was found that (i) HCMV core NEC proteins are not phosphorylated solely by viral pUL97, but also by cellular kinases; (ii) both PKC and CDK1 phosphorylation are detectable for pUL50; (iii) no impact of PKC phosphorylation on NEC functionality has been identified so far; (iv) nonetheless, CDK1-specific phosphorylation appears to be required for functional core NEC interaction. In summary, our findings provide the first evidence that the HCMV core NEC is phosphorylated by cellular kinases, and that the complex pattern of NEC phosphorylation has functional relevance.

  13. A system for obtaining an optimized pre design of nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, L.A.

    1989-01-01

    This work proposes a method for obtaing a first design of nuclear reactor cores. It takes into consideration the objectives of the project, physical limits, economical limits and the reactor safety. For this purpose, some simplifications were made in the reactor model: one-energy-group, unidimensional and homogeneous core. The adopted model represents a typical PWR core and the optimized parameters are the fuel thickness, refletor thickness, enrichement and moderating ratio. The objective is to gain a larger residual reactivity at the end of the cycle. This work also presents results for a PWR core. From the results, many conclusions are established: system efficiency, limitations and problems. Also some suggestions are proposed to improve the system performance for futures works. (author) [pt

  14. Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

    International Nuclear Information System (INIS)

    Tanabe, Fumiya

    2011-01-01

    In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)

  15. A system to obtain an optimized first design of a nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, L.A.

    1988-01-01

    This work proposes a method for obtaining a first design of nuclear reactor cores. It takes into consideration the objectives of the project, physical limits, economical limits and the reactor safety. For this purpose, some simplifications were made in the reactor model: one energy-group, one-dimensional and homogeneous core. The adopted model represents a typical PWR core and the optimized parameters are the fuel thickness, reflector thickness, enrichment and moderating ratio. The objective is to gain a larger residual reactivity at the end of the cycle. This work also presents results for a PWR core. From the results, many conclusions are established: system efficiency, limitations and problems. Also some suggestions are proposed to improve the system performance for future works. (autor)

  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. Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

    International Nuclear Information System (INIS)

    Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

    2001-01-01

    The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

  18. Conversion of the core of the TRIGA Mark III reactor at the Mexican Nuclear Centre

    International Nuclear Information System (INIS)

    Moran Lopez, J.M.; Lucatero, M.A.; Reyes Andrade, B.; Rivero Gutierrez, T.; Sainz Mejia, E.

    1990-01-01

    It was decided to convert the core of the TRIGA MARK III reactor at the Mexican Nuclear Centre run by the National Nuclear Institute because of problems detected during the operation, such as a lack of excess reactivity for operation at nominal power over long periods and difficulties in the maintenance and calibration of the control panel. In order to compensate for the lack of excess reactivity the fuel elements taken to the highest burnup were replaced by fresh elements acquired for this purpose. The latter, however, had a different enrichment, and this necessitated a detailed analysis of the neutronic and thermohydraulic behaviour of the reactor with a view to determining a mixed core configuration which would meet safe operation requirements. In conducting the thermohydraulic analysis, a natural convection coolant flow model was developed to determine coolant velocity and pressure drop patterns within the core. The heat transfer equations were solved and it was found that the hottest fuel element did not attain critical heat flux conditions. In loading this core it was also necessary to analyse procedures and to consider the possible effects of reaching criticality with fuel elements having different enrichments. The loading procedure is described, as is the measurement system and the results obtained. In order to resolve the calibration and maintenance problems, a new, more advanced control panel was designed with conventional and nuclear detection systems and modern components

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

    International Nuclear Information System (INIS)

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    2002-01-01

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

  1. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    Nagano, M.; Sakurai, S.; Yamaguchi, H.

    1997-01-01

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  2. Sensitivity of control times in function of core parameters and oscillations control in thermal nuclear systems

    International Nuclear Information System (INIS)

    Amorim, E.S. do; D'Oliveira, A.B.; Galvao, O.B.; Oyama, K.

    1981-03-01

    Sensitivity of control times to variation of a thermal reactor core parameters is defined by suitable changes in the power coefficient, core size and fuel enrichment. A control strategy is developed based on control theory concepts and on considerations of the physics of the problem. Digital diffusion theory simulation is described which tends to verify the control concepts considered, face dumped oscillations introduced in one thermal nuclear power system. The effectivity of the control actions, in terms of eliminating oscillations, provided guidelines for the working-group engaged in the analysis of the control rods and its optimal performance. (Author) [pt

  3. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt

    International Nuclear Information System (INIS)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects

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

    International Nuclear Information System (INIS)

    Jelinek, Tomas

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  6. 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

  7. In-core assembly configuration having a dual-wall pressure boundary for nuclear reactor

    International Nuclear Information System (INIS)

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

    1988-01-01

    This patent describes an in-core detector assembly of the type having an in-core part and an out-of-core part and having an elongated outer hollow housing tube with a wall thickness, an inner hollow calibration tube with a wall thickness and disposed concentrically within the outer tube to define an annular space therewith, and a plurality of discrete, circular, rod-like elements extending through the annular space, the improvement comprising: the elements having outer diameters and being of a number to substantially occupy the entire annular space of both the incore and out-of-core parts without significant voids between elements; each of the elements including at least an outer sheath and interior highly compacted mineral insulation for the entire length of the element; a first number of the elements also including center lead means connected to condition responsive element means in the in-core part of the length of the assembly and a second, remaining number of the elements being non-operating elements. The wall thickness of the housing tube and the wall thickness of the calibration tube, taken together with the diameter of the elements, provide a thickness dimension adequate to meet code primary pressure requirements for normal nuclear reactor in-core conditions, while the wall thickness of the calibration tube alone provides a thickness dimension less than adequate to meet such requirements

  8. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  9. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y H; Cha, K H; Lee, S H [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  10. Design and test of the borosilicate glass burnable poison rod for Qinshan nuclear power plant core

    International Nuclear Information System (INIS)

    Huang Jinhua; Sun Hanhong

    1988-08-01

    Material for the burnable poison of Qinshan Nuclear Power Plant core is GG-17 borosilicate glass. The chemical composition and physico-chemical properties of GG-17 is very close to Pyrex-7740 glass used by Westinghouse. It is expected from the results of the experiments that the borosilicate glass burnable poison rod can be successfully used in Qinshan Nuclear Power Plant due to good physical, mechanical, corrosion-resistant and irradiaton properties for both GG-17 glass and cold-worked stainless steel cladding. Change of material for burnable poison from boron-bearing stainless steel to borosilicate glass will bring about much more economic benefit to Qinshan Naclear Power Plant

  11. An assessment of coupling algorithms for nuclear reactor core physics simulations

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, Steven, E-mail: hamiltonsp@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Berrill, Mark, E-mail: berrillma@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Clarno, Kevin, E-mail: clarnokt@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Pawlowski, Roger, E-mail: rppawlo@sandia.gov [Sandia National Laboratories, MS 0316, P.O. Box 5800, Albuquerque, NM 87185 (United States); Toth, Alex, E-mail: artoth@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Kelley, C.T., E-mail: tim_kelley@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Evans, Thomas, E-mail: evanstm@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States)

    2016-04-15

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

  12. A new equation of state Based on Nuclear Statistical Equilibrium for Core-Collapse Simulations

    Science.gov (United States)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2012-09-01

    We calculate a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ~ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect on the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores.

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  14. Gaseous core nuclear-driven engines featuring a self-shutoff mechanism to provide nuclear safety

    International Nuclear Information System (INIS)

    Heidrich, J.; Pettibone, J.; Chow, Tze-Show; Condit, R.; Zimmerman, G.

    1991-11-01

    Nuclear driven engines are described that could be run in either pulsed or steady state modes. In the pulsed mode nuclear energy is released by fissioning of uranium or plutonium in a supercritical assembly of fuel and working gas. In a steady state mode a fuel-gas mixture is injected into a magnetic nozzle where it is compressed into a critical state and produces energy. Engine performance is modeled using a code that calculates hydrodynamics, fission energy production, and neutron transport self-consistently. Results are given demonstrating a large negative temperature coefficient that produces self-shutoff or control of energy production. Reduced fission product inventory and the self-shutoff provide inherent nuclear safety. It is expected that nuclear engine reactor units could be scaled up from about 100 MW e

  15. Optimization analysis of the nuclear fuel cycle transition to the last core

    International Nuclear Information System (INIS)

    Rebollo, L.; Blanco, J.

    2001-01-01

    The Zorita NPP was the first Spanish commercial nuclear reactor connected to the grid. It is a 160 MW one loop PWR, Westinghouse design, owned by UFG, in operation since 1968. The configuration of the reactor core is based on 69 fuel elements type 14 x 14, the standard reload of the present equilibrium cycle being based on 16 fuel elements with 3.6% enrichment in 235 U. In order to properly plan the nuclear fuel management of the transition cycles to its end of life, presently foreseen by 2008, an based on the non-reprocessing option required by the policy of the Spanish Administration, a technical-economical optimization analysis has been performed. As a result, a fuel management strategy has been defined looking for getting simultaneously the minimum integral fuel cost of the transition from the present equilibrium cycle to the last core, as well as the minimum residual worth of the fuel remaining in the core after the final outage. Based on the ''lessons learned'' derived from the study, the time margin for the decision making has been determined, and a planning of the nuclear fuel supply for the transition reloads, specifying both the number of fuel elements and their enrichment in 235 U, as been prepared. Finally, based on the calculated economical worth of the partially burned fuel of the last core, after the end of its operation cycle, a financial cover for yearly compensation from now on of the foreseen final lost has been elaborated. Most of the conceptual conclusions obtained are applicable to the other commercial nuclear reactors in operation owned by UFG, so that they are understood to be of general interest and broad application to commercial PWR. (author)

  16. Scale model study of the seismic response of a nuclear reactor core

    International Nuclear Information System (INIS)

    Dove, R.C.; Dunwoody, W.E.; Rhorer, R.L.

    1983-01-01

    The use of scale models to study the dynamics of a system of graphite core blocks used in certain nuclear reactor designs is described. Scaling laws, material selecton, model instrumentation to measure collision forces, and the response of several models to simulated seismic excitation are covered. The effects of Coulomb friction between the blocks and the clearance gaps between the blocks on the system response to seismic excitation are emphasized

  17. About the application of MCNP4 code in nuclear reactor core design calculations

    International Nuclear Information System (INIS)

    Svarny, J.

    2000-01-01

    This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)

  18. In core reload design for cycle 4 of Daya Bay nuclear power station both units

    International Nuclear Information System (INIS)

    Zhang Zongyao; Liu Xudong; Xian Chunyu; Li Dongsheng; Zhang Hong; Liu Changwen; Rui Min; Wang Yingming; Zhao Ke; Zhang Hong; Xiao Min

    1998-01-01

    The basic principles and the contents of the reload design for Daya Bay nuclear power station are briefly introduced. The in core reload design results, and the comparison between the calculated values and the measured values of both units the fourth cycle are also given. The reload design results of the two units satisfy all the economic requirements and safety criteria. The experimented results shown that the predicated values are tally good with all the measurement values

  19. Measurements of the HEU and LEU in-core spectra at the Ford Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wehe, D K [Oak Ridge National Laboratory, Oak Ridge, TN (United States); King, J S; Lee, J C; Martin, W R [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

    1985-07-01

    The Ford Nuclear Reactor (FNR) at the University of Michigan has been serving as the test site for a low-enriched uranium (LEU) fuel whole-core demonstration. As part of the experimental program, the differential neutron spectrum has been measured in a high-enriched uranium (HEU) core and an LEU core. The HEU and LEU spectra were determined by unfolding the measured activities of foils that were irradiated in the reactor. When the HEU and LEU spectra are compared from meV to 10 MeV, significant differences between the two spectra are apparent below 10 eV. These are probably caused by the additional {sup 238}U resonance absorption in the LEU fuel. No measurable difference occurs in the shape of the spectra above MeV. (author)

  20. Rift Valley fever phlebovirus NSs protein core domain structure suggests molecular basis for nuclear filaments.

    Science.gov (United States)

    Barski, Michal; Brennan, Benjamin; Miller, Ona K; Potter, Jane A; Vijayakrishnan, Swetha; Bhella, David; Naismith, James H; Elliott, Richard M; Schwarz-Linek, Ulrich

    2017-09-15

    Rift Valley fever phlebovirus (RVFV) is a clinically and economically important pathogen increasingly likely to cause widespread epidemics. RVFV virulence depends on the interferon antagonist non-structural protein (NSs), which remains poorly characterized. We identified a stable core domain of RVFV NSs (residues 83-248), and solved its crystal structure, a novel all-helical fold organized into highly ordered fibrils. A hallmark of RVFV pathology is NSs filament formation in infected cell nuclei. Recombinant virus encoding the NSs core domain induced intranuclear filaments, suggesting it contains all essential determinants for nuclear translocation and filament formation. Mutations of key crystal fibril interface residues in viruses encoding full-length NSs completely abrogated intranuclear filament formation in infected cells. We propose the fibrillar arrangement of the NSs core domain in crystals reveals the molecular basis of assembly of this key virulence factor in cell nuclei. Our findings have important implications for fundamental understanding of RVFV virulence.

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

    International Nuclear Information System (INIS)

    Hecker, H.C.

    1984-04-01

    This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U 233 -Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred

  2. Comment on the in-core measurement in the WWER nuclear power plant

    International Nuclear Information System (INIS)

    Krett, V.; Dach, K.; Erben, O.

    1985-01-01

    The activity of the Nuclear Research Institute (NRI) Rez in the field of in-core measurement sensors is described in the paper. The results of comparison and calibration experiments realized on the WWR-S research reactor at the NRI are presented. Measurements with fission calorimeters and SPN detectors carried out in the framework of diagnostic fuel assembly program of WWER NPP reactors are described. Noise measurements with detectors of in-core instrumentation of diagnostic fuel assemblies are also mentioned. Comparison experiments on the WWER-440 NPP reactor are described and the method of function verification of neutron sensors of the in-core control system of these reactors is given. (author)

  3. 3D Core Model for simulation of nuclear power plants: Simulation requirements, model features, and validation

    International Nuclear Information System (INIS)

    Zerbino, H.

    1999-01-01

    In 1994-1996, Thomson Training and Simulation (TT and S) earned out the D50 Project, which involved the design and construction of optimized replica simulators for one Dutch and three German Nuclear Power Plants. It was recognized early on that the faithful reproduction of the Siemens reactor control and protection systems would impose extremely stringent demands on the simulation models, particularly the Core physics and the RCS thermohydraulics. The quality of the models, and their thorough validation, were thus essential. The present paper describes the main features of the fully 3D Core model implemented by TT and S, and its extensive validation campaign, which was defined in extremely positive collaboration with the Customer and the Core Data suppliers. (author)

  4. Compact sodium cooled nuclear power plant with fast core (KNK II- Karlsruhe), Safety Report

    International Nuclear Information System (INIS)

    1977-09-01

    After the operation of the KNK plant with a thermal core (KNK I), the installation of a fast core (KNK II) had been realized. The planning of the core and the necessary reconstruction work was done by INTERATOM. Owner and customer was the Nuclear Research Center Karlsruhe (KfK), while the operating company was the Kernkraftwerk-Betriebsgesellschaft mbH (KBG) Karlsruhe. The main goals of the KNK II project and its special experimental test program were to gather experience for the construction, the licensing and operation of future larger plants, to develop and to test fuel and absorber assemblies and to further develop the sodium technology and the associated components. The present safety report consists of three parts. Part 1 contains the description of the nuclear plant. Hereby, the reactor and its components, the handling facilities, the instrumentation with the plant protection, the design of the plant including the reactor core and the nominal operation processes are described. Part 2 contains the safety related investigation and measures. This concerns the reactivity accidents, local cooling perturbations, radiological consequences with the surveillance measures and the justification of the choice of structural materials. Part three finally is the appendix with the figures, showing the different buildings, the reactor and its components, the heat transfer systems and the different auxiliary facilities [de

  5. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S., E-mail: alexandrossilva@ifba.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia da Bahia (IFBA), Vitoria da Conquista, BA (Brazil); Mazaira, Leorlen Y.R., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Dominguez, Dany S.; Hernandez, Carlos R.G., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Programa de Pos-Graduacao em Modelagem Computacional; Lira, Carlos A.B.O., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  6. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    International Nuclear Information System (INIS)

    Silva, Alexandro S.; Mazaira, Leorlen Y.R.; Dominguez, Dany S.; Hernandez, Carlos R.G.

    2015-01-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  7. The Impact of the Nuclear Equation of State in Core Collapse Supernovae

    Science.gov (United States)

    Baird, M. L.; Lentz, E. J.; Hix, W. R.; Mezzacappa, A.; Messer, O. E. B.; Liebendoerfer, M.; TeraScale Supernova Initiative Collaboration

    2005-12-01

    One of the key ingredients to the core collapse supernova mechanism is the physics of matter at or near nuclear density. Included in simulations as part of the Equation of State (EOS), nuclear repulsion experienced at high densities are responsible for the bounce shock, which initially causes the outer envelope of the supernova to expand, as well as determining the structure of the newly formed proto-neutron star. Recent years have seen renewed interest in this fundamental piece of supernova physics, resulting in several promising candidate EOS parameterizations. We will present the impact of these variations in the nuclear EOS using spherically symmetric, Newtonian and General Relativistic neutrino transport simulations of stellar core collapse and bounce. This work is supported in part by SciDAC grants to the TeraScale Supernovae Initiative from the DOE Office of Science High Energy, Nuclear, and Advanced Scientific Computing Research Programs. Oak Ridge National Laboratory is managed by UT-Battelle, LLC, for U.S. Department of Energy under contract DEAC05-00OR22725

  8. Comparison of serpent and triton generated FEW group constants for APR1400 nuclear reactor core

    International Nuclear Information System (INIS)

    Elsawi, Mohamed A.; Alnoamani, Zainab

    2015-01-01

    The accuracy of full-core reactor power calculations using diffusion codes is strongly dependent on the quality of the homogenized cross sections and other few-group constants generated by lattice codes. For many years, deterministic lattice codes have been used to generate these constants using different techniques: the discrete ordinates, collision probability or the method of characteristics, just to name a few. These codes, however, show some limitations, for example, on complex geometries or near heavy absorbers as in modern pressurized water reactor (PWR) designs like the Korean Advanced Power Reactor 1400 (APR1400) core. The use of continuous-energy Monte Carlo (MC) codes to produce nuclear constants can be seen as an attractive option when dealing with fuel or reactor types that lie beyond the capabilities of conventional deterministic lattice transport codes. In this paper, the few-group constants generated by two of the state-of-the-art reactor physics codes, SERPENT and SCALE/TRITON, will be critically studied and their reliability for being used in subsequent diffusion calculations will be evaluated. SERPENT is a 3D, continuous-energy, Monte Carlo reactor physics code which has a built-in burn-up calculation capability. It has been developed at the Technical Research Center of Finland (VTT) since 2004. SCALE/TRITON, on the other hand, is a control module developed within the framework of SCALE package that enables performing deterministic 2-D transport calculations on nuclear reactor core lattices. The approach followed in this paper is as follows. First, the few-group nuclear constants for the APR1400 reactor core were generated using SERPENT (version 2.1.22) and NEWT (in SCALE version 6.1.2) codes. For both codes, the critical spectrum, calculated using the B1 method, was used as a weighting function. Second, 2-D diffusion calculations were performed using the US NRC core simulator PARCS employing the two few-group constant sets generated in the first

  9. Quantification of cost of margin associated with in-core nuclear fuel management for a PWR

    International Nuclear Information System (INIS)

    Kropaczek, D.J.; Turinsky, P.J.

    1989-01-01

    The problem of in-core nuclear fuel management optimization is discussed. The problem is to determine the location of core material, such as the fuel and burnable poisons, so as to minimize (maximize) a stated objective within engineering constraints. Typical objectives include maximization of cycle energy production or discharged fuel exposure, and minimization of power peaking factor or reactor vessel fluence. Constraints include discharge burnup limits and one or more of the possible objectives if not selected as the objective. The optimization problem can be characterized as a large combinatorial problem with nonlinear objective function and constraints, which are likely to be active. The authors have elected to employ the integer Monte Carlo programming method to address this optimization problem because of the just-noted problem characteristics. To evaluate the core physics characteristics as a function of fuel loading pattern, second-order accurate perturbation theory is employed with successive application to improve estimates of the optimum loading pattern. No constraints on fuel movement other than requiring quarter-core symmetry were imposed. In this paper the authors employed this methodology to address a related problem. The problem being addressed can be stated as What is the cost associated with margin? Specifically, they wish to assign some financial value in terms of increased levelized fuel cycle cost associated with an increase in core margin of some type, such as power peaking factor

  10. Burst shield for a pressurized nuclear-reactor core and method of operating same

    International Nuclear Information System (INIS)

    Beine, B.; Schilling, F.

    1976-01-01

    A pressurized nuclear-reactor core stands on a base up from which extends a cylindrical side wall formed of a plurality of hollow iron castings held together by circumferential and longitudinal prestressed elements. A cylindrical space between this shield and the core serves for inspection of the core and is normally filled with cast-iron segmental slabs so that if the core bursts pieces thrown out do not acquire any dangerous kinetic energy before engaging the burst shield. The top of the shield is removably secured to the side so that it can be moved out of the way periodically for removal of the filler slabs and inspection of the core. An anchor on the upper end of each longitudinal prestressing element bears against a sleeve pressing against the uppermost side element, and a nut engageable with this anchor is engageable down over the top to hold it in place, removal of this nut leaving the element prestressed in the side wall. 11 claims, 16 drawing figures

  11. Neutrino-pair emission from nuclear de-excitation in core-collapse supernova simulations

    Science.gov (United States)

    Fischer, T.; Langanke, K.; Martínez-Pinedo, G.

    2013-12-01

    We study the impact of neutrino-pair production from the de-excitation of highly excited heavy nuclei on core-collapse supernova simulations, following the evolution up to several 100 ms after core bounce. Our study is based on the agile-boltztransupernova code, which features general relativistic radiation hydrodynamics and accurate three-flavor Boltzmann neutrino transport in spherical symmetry. In our simulations the nuclear de-excitation process is described in two different ways. At first we follow the approach proposed by Fuller and Meyer [Astrophys. J.AJLEEY0004-637X10.1086/170317 376, 701 (1991)], which is based on strength functions derived in the framework of the nuclear Fermi-gas model of noninteracting nucleons. Second, we parametrize the allowed and forbidden strength distributions in accordance with measurements for selected nuclear ground states. We determine the de-excitation strength by applying the Brink hypothesis and detailed balance. For both approaches, we find that nuclear de-excitation has no effect on the supernova dynamics. However, we find that nuclear de-excitation is the leading source for the production of electron antineutrinos as well as heavy-lepton-flavor (anti)neutrinos during the collapse phase. At sufficiently high densities, the associated neutrino spectra are influenced by interactions with the surrounding matter, making proper simulations of neutrino transport important for the determination of the neutrino-energy loss rate. We find that, even including nuclear de-excitations, the energy loss during the collapse phase is overwhelmingly dominated by electron neutrinos produced by electron capture.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  13. An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

    International Nuclear Information System (INIS)

    Clark, J.S.; Walton, J.T.; Mcguire, M.L.

    1992-07-01

    Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines. 11 refs

  14. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    International Nuclear Information System (INIS)

    Silva, Alexandro S.; Dominguez, Dany S.; Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G.; Lira, Carlos Alberto Brayner de Oliveira

    2015-01-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  15. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M J; Oyinloye, J O; Chambers, I [Electrowatt Consulting Engineers and Scientists, Warrington, Cheshire (United Kingdom); Scott, C K [Atlantic Nuclear Services, Fredericton, NB (Canada); Omar, A M [Atomic Energy Control Board, Ottawa, ON (Canada)

    1991-12-31

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs.

  16. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    International Nuclear Information System (INIS)

    Lewis, M.J.; Oyinloye, J.O.; Chambers, I.; Scott, C.K.; Omar, A.M.

    1990-01-01

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs

  17. In core instrumentation for online nuclear heating measurements of material testing reactor

    International Nuclear Information System (INIS)

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

    2010-01-01

    The present work focuses on nuclear heating. This work belongs to a new advanced research program called IN-CORE which means 'Instrumentation for Nuclear radiations and Calorimetry Online in REactor' between the LCP (University of Provence-CNRS) and the CEA (French Atomic Energy Commission) - Jules Horowitz Reactor (JHR) program. This program started in September 2009 and is dedicated to the conception and the design of an innovative mobile experimental device coupling several sensors and ray detectors for on line measurements of relevant physical parameters (photonic heating, neutronic flux ...) and for an accurate parametric mapping of experimental channels in the JHR Core. The work presented below is the first step of this program and concerns a brief state of the art related to measurement methods of nuclear heating phenomena in research reactor in general and MTR in particular. A special care is given to gamma heating measurements. A first part deals with numerical codes and models. The second one presents instrumentation divided into various kinds of sensor such as calorimeter measurements and gamma ionization chamber measurements. Their basic principles, characteristics such as metrological parameters, operating mode, disadvantages/advantages, ... are discussed. (author)

  18. A genetic algorithm solution for combinatorial problems - the nuclear core reload example

    Energy Technology Data Exchange (ETDEWEB)

    Schirru, R.; Silva, F.C. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia; Pereira, C.M.N.A. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil); Chapot, J.L.C. [FURNAS, Rio de Janeiro, RJ (Brazil)

    1997-12-01

    This paper presents a solution to Traveling Salesman Problem based upon genetic algorithms (GA), using the classic crossover, but avoiding the feasibility problem in offspring individuals, allowing the natural evolution of the GA without introduction of heuristics in the genetic crossover operator. The genetic model presented, that we call the List Model (LM) is based on the encoding and decoding genotype in the way to always generate a phenotype that has a valid structure, over which will be applied the fitness, represented by the total distance. The main purpose of this work was to develop the basis for a new genetic model to be used in the reload of nuclear core of a PWR. In a generic way, this problem can be interpreted as a a search of the optimal combination of N different fuel elements in N nuclear core `holes`, where each combination or load pattern, determines the neutron flux shape and its associate peak factor. The goal is to find out the load pattern that minimizes the peak factor and consequently maximize the useful life of the nuclear fuel. The GA with the List Model was applied to the Angra-1 PWR reload problem and the results are remarkably better than the ones used in the last fuel cycle. (author). 12 refs., 3 figs., 2 tabs.

  19. Review of the Shoreham Nuclear Power Station Probabilistic Risk Assessment: internal events and core damage frequency

    International Nuclear Information System (INIS)

    Ilberg, D.; Shiu, K.; Hanan, N.; Anavim, E.

    1985-11-01

    A review of the Probabilistic Risk Assessment of the Shoreham Nuclear Power Station was conducted with the broad objective of evaluating its risks in relation to those identified in the Reactor Safety Study (WASH-1400). The scope of the review was limited to the ''front end'' part, i.e., to the evaluation of the frequencies of states in which core damage may occur. Furthermore, the review considered only internally generated accidents, consistent with the scope of the PRA. The review included an assessment of the assumptions and methods used in the Shoreham study. It also encompassed a reevaluation of the main results within the scope and general methodological framework of the Shoreham PRA, including both qualitative and quantitative analyses of accident initiators, data bases, and accident sequences which result in initiation of core damage. Specific comparisons are given between the Shoreham study, the results of the present review, and the WASH-1400 BWR, for the core damage frequency. The effect of modeling uncertainties was considered by a limited sensitivity study so as to show how the results would change if other assumptions were made. This review provides an independently assessed point value estimate of core damage frequency and describes the major contributors, by frontline systems and by accident sequences. 17 figs., 81 tabs

  20. Out-of-core nuclear fuel cycle optimization utilizing an engineering workstation

    International Nuclear Information System (INIS)

    Turinsky, P.J.; Comes, S.A.

    1986-01-01

    Within the past several years, rapid advances in computer technology have resulted in substantial increases in their performance. The net effect is that problems that could previously only be executed on mainframe computers can now be executed on micro- and minicomputers. The authors are interested in developing an engineering workstation for nuclear fuel management applications. An engineering workstation is defined as a microcomputer with enhanced graphics and communication capabilities. Current fuel management applications range from using workstations as front-end/back-end processors for mainframe computers to completing fuel management scoping calculations. More recently, interest in using workstations for final in-core design calculations has appeared. The authors have used the VAX 11/750 minicomputer, which is not truly an engineering workstation but has comparable performance, to complete both in-core and out-of-core fuel management scoping studies. In this paper, the authors concentrate on our out-of-core research. While much previous work in this area has dealt with decisions concerned with equilibrium cycles, the current project addresses the more realistic situation of nonequilibrium cycles

  1. Spent nuclear fuel application of CORE reg-sign systems engineering software

    International Nuclear Information System (INIS)

    Grimm, R.J.

    1996-01-01

    The Department of Energy (DOE) has adopted a systems engineering approach for the successful completion of the Spent Nuclear Fuel (SNF) Program mission. The DOE has utilized systems engineering principles to develop the SNF Program guidance documents and has held several systems engineering workshops to develop the functional hierarchies of both the programmatic and technical side of the SNF Program. The sheer size and complexity of the SNF Program, however, has led to problems that the Westinghouse Savannah River Company (WSRC) is working to manage through the use of systems engineering software. WSRC began using CORE reg-sign, an off-the-shelf PC based software package, to assist the DOE in management of the SNF program. This paper details the successful use of the CORE reg-sign systems engineering software to date and the proposed future activities

  2. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    International Nuclear Information System (INIS)

    Prabhu Gaunkar, N.; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C.; Bulu, I.; Ganesan, K.; Song, Y. Q.

    2015-01-01

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors

  3. A nuclear heuristic for application to metaheuristics in-core fuel management optimization

    Energy Technology Data Exchange (ETDEWEB)

    Meneses, Anderson Alvarenga de Moura, E-mail: ameneses@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program; Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Gambardella, Luca Maria, E-mail: luca@idsia.c [Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Schirru, Roberto, E-mail: schirru@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program

    2009-07-01

    The In-Core Fuel Management Optimization (ICFMO) is a well-known problem of nuclear engineering whose features are complexity, high number of feasible solutions, and a complex evaluation process with high computational cost, thus it is prohibitive to have a great number of evaluations during an optimization process. Heuristics are criteria or principles for deciding which among several alternative courses of action are more effective with respect to some goal. In this paper, we propose a new approach for the use of relational heuristics for the search in the ICFMO. The Heuristic is based on the reactivity of the fuel assemblies and their position into the reactor core. It was applied to random search, resulting in less computational effort concerning the number of evaluations of loading patterns during the search. The experiments demonstrate that it is possible to achieve results comparable to results in the literature, for future application to metaheuristics in the ICFMO. (author)

  4. Spent nuclear fuel application of CORE reg-sign systems engineering software

    International Nuclear Information System (INIS)

    Grimm, R.J.

    1996-01-01

    The DOE has adopted a systems engineering approach for the successful completion of the Spent Nuclear Fuel (SNF) Program mission. The DOE has utilized systems engineering principles to develop the SNF program guidance documents and has held several systems engineering workshops to develop the functional hierarchies of both the programmatic and technical side of the SNF program. The sheer size and complexity of the SNF program has led to problems that the Westinghouse Savannah River Company (WSRC) is working to manage through the use of systems engineering software. WSRC began using CORE reg-sign, an off the shelf PC based software package, to assist DOE in management of the SNF program. This paper details the successful use of the CORE reg-sign systems engineering software to date and the proposed future activities

  5. Control and balance of nuclear matters used for core fabrication of Super Phenix

    International Nuclear Information System (INIS)

    Beche, M.; Guillet, H.; Heyraud, H.; Levrard, J.; Pajot, J.

    1987-05-01

    The fabrication of the core of the fast breeder reactor set up at Creys Malville ended in March 1984. It started in 1978 and it required, for the fabrication of the 410 assemblies, the utilization of 7438 kg of plutonium. To satisfy national and international regulations, DPFER/SFER has used a methodology to follow and to control the movements of the nuclear materials. These controls are achieved by physical methods, chemical methods and empiric methods. Euratom has conducted a succession of inspections during the 5.5 years of that campaign. The inventory difference, in the fabrication of that core, represents about 0.1% of the total mass of the plutonium handled [fr

  6. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases

  7. A nuclear heuristic for application to metaheuristics in-core fuel management optimization

    International Nuclear Information System (INIS)

    Meneses, Anderson Alvarenga de Moura; Gambardella, Luca Maria; Schirru, Roberto

    2009-01-01

    The In-Core Fuel Management Optimization (ICFMO) is a well-known problem of nuclear engineering whose features are complexity, high number of feasible solutions, and a complex evaluation process with high computational cost, thus it is prohibitive to have a great number of evaluations during an optimization process. Heuristics are criteria or principles for deciding which among several alternative courses of action are more effective with respect to some goal. In this paper, we propose a new approach for the use of relational heuristics for the search in the ICFMO. The Heuristic is based on the reactivity of the fuel assemblies and their position into the reactor core. It was applied to random search, resulting in less computational effort concerning the number of evaluations of loading patterns during the search. The experiments demonstrate that it is possible to achieve results comparable to results in the literature, for future application to metaheuristics in the ICFMO. (author)

  8. Impact of correlations between core configurations for the evaluation of nuclear data uncertainty propagation for reactivity

    International Nuclear Information System (INIS)

    Frosio, T.; Bonaccorsi, T.; Blaise, P.

    2017-01-01

    The precise estimation of Pearson correlations, also called 'representativity' coefficients, between core configurations is a fundamental quantity for properly assessing the nuclear data (ND) uncertainties propagation on integral parameters such as k-eff, power distributions, or reactivity coefficients. In this paper, a traditional adjoint method is used to propagate ND uncertainty on reactivity and reactivity coefficients and estimate correlations between different states of the core. We show that neglecting those correlations induces a loss of information in the final uncertainty. We also show that using approximate values of Pearson does not lead to an important error of the model. This calculation is made for reactivity at the beginning of life and can be extended to other parameters during depletion calculations. (authors)

  9. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    Energy Technology Data Exchange (ETDEWEB)

    Prabhu Gaunkar, N., E-mail: neelampg@iastate.edu; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C. [Department of Electrical and Computer Engineering, Iowa State University, Ames, Iowa 50011 (United States); Bulu, I.; Ganesan, K.; Song, Y. Q. [Schlumberger-Doll Research, Cambridge, Massachusetts 02139 (United States)

    2015-05-07

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors.

  10. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given

  11. The effect of uncertainties in nuclear reactor plant-specific failure data on core damage frequency

    International Nuclear Information System (INIS)

    Martz, H.F.

    1995-05-01

    It is sometimes the case in PRA applications that reported plant-specific failure data are, in fact, only estimates which are uncertain. Even for detailed plant-specific data, the reported exposure time or number of demands is often only an estimate of the actual exposure time or number of demands. Likewise the reported number of failure events or incidents is sometimes also uncertain because incident or malfunction reports may be ambiguous. In this report we determine the corresponding uncertainty in core damage frequency which can b attributed to such uncertainties in plant-specific data using a simple but typical nuclear power reactor example

  12. Interaction between core analysis methodology and nuclear design: some PWR examples

    International Nuclear Information System (INIS)

    Rothleder, B.M.; Eich, W.J.

    1982-01-01

    The interaction between core analysis methodology and nuclear design is exemplified by PSEUDAX, a major improvement related to the Advanced Recycle methodology program (ARMP) computer code system, still undergoing development by the Electric Power Research Institute. The mechanism of this interaction is explored by relating several specific nulcear design changes to the demands placed by these changes on the ARMP system, and by examining the meeting of these demands, first within the standard ARMP methodology and then through augmentation of the standard methodology by development of PSEUDAX

  13. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  14. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant

    International Nuclear Information System (INIS)

    Zamora R, L.; Medina F, A.

    1999-01-01

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

  15. A nuclear reactor core fuel reload optimization using Artificial-Ant-Colony Connective Networks; Recarga de reatores nucleares utilizando redes conectivas de colonias de formigas artificiais

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de; Schirru, Roberto [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: alan@lmp.ufrj.br; schirru@lmp.ufrj.br

    2005-07-01

    A Pressurized Water Reactor core must be reloaded every time the fuel burnup reaches a level when it is not possible to sustain nominal power operation. The nuclear core fuel reload optimization consists in finding a burned-up and fresh-fuel-assembly pattern that maximizes the number of full operational days. This problem is NP-hard, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Besides that, the problem is non-linear and its search space is highly discontinual and multimodal. In this work a parallel computational system based on Ant Colony System (ACS) called Artificial-Ant-Colony Networks is introduced to solve the nuclear reactor core fuel reload optimization problem. ACS is a system based on artificial agents that uses the reinforcement learning technique and was originally developed to solve the Traveling Salesman Problem, which is conceptually similar to the nuclear fuel reload problem. (author)

  16. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Directory of Open Access Journals (Sweden)

    Gholam Reza Ansarifar

    2015-12-01

    Full Text Available One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC, which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  17. Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

    International Nuclear Information System (INIS)

    Friess, Friederike Renate

    2017-01-01

    According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

  18. Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Friess, Friederike Renate

    2017-07-12

    According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

  19. Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 1. ZPPR analysis results

    International Nuclear Information System (INIS)

    Mantourov, Guennadi

    2001-05-01

    This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess of Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. The data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS and ZPPR fast reactor cores applying JNC core calculation code CITATION. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the first flight collision probability method and subgroup approach. Especially a converting program was written to transmit the prepared effective cross sections to JNC standard PDS files. Then the CITATION code was applied for 3-D XYZ neutronics calculations of BFS and ZPPR JUPITER experiments series cores. The effects of nuclear data library have been studied by comparing the former results based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for ZPPR-9, ZPPR-13A and ZPPR-17A cores are presented. The calculated correction factor in all cases was less than 1.0%. So the uncertainty in C value caused by possible errors in calculation of these corrections is expected to be less than 0.3% in case of ZPPR-13A and ZPPR-17A cores, and rather less for ZPPR-9 core. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC calculation route revealed a large enough discrepancy in k-eff for ZPPR-9 (about 0.6%) and ZPPR-17A (about 0.5%) cores. For BFS-62-1 and BFS-62-2 cores such analysis is in progress. Stretch cell models for both BFS cores were formed and cell calculations using FFCP code have started. Some results of cell calculations are presented. (author)

  20. Development of a parallel genetic algorithm using MPI and its application in a nuclear reactor core. Design optimization

    International Nuclear Information System (INIS)

    Waintraub, Marcel; Pereira, Claudio M.N.A.; Baptista, Rafael P.

    2005-01-01

    This work presents the development of a distributed parallel genetic algorithm applied to a nuclear reactor core design optimization. In the implementation of the parallelism, a 'Message Passing Interface' (MPI) library, standard for parallel computation in distributed memory platforms, has been used. Another important characteristic of MPI is its portability for various architectures. The main objectives of this paper are: validation of the results obtained by the application of this algorithm in a nuclear reactor core optimization problem, through comparisons with previous results presented by Pereira et al.; and performance test of the Brazilian Nuclear Engineering Institute (IEN) cluster in reactors physics optimization problems. The experiments demonstrated that the developed parallel genetic algorithm using the MPI library presented significant gains in the obtained results and an accentuated reduction of the processing time. Such results ratify the use of the parallel genetic algorithms for the solution of nuclear reactor core optimization problems. (author)

  1. The Use of Hidden Markov Models for Anomaly Detection in Nuclear Core Condition Monitoring

    Science.gov (United States)

    Stephen, Bruce; West, Graeme M.; Galloway, Stuart; McArthur, Stephen D. J.; McDonald, James R.; Towle, Dave

    2009-04-01

    Unplanned outages can be especially costly for generation companies operating nuclear facilities. Early detection of deviations from expected performance through condition monitoring can allow a more proactive and managed approach to dealing with ageing plant. This paper proposes an anomaly detection framework incorporating the use of the Hidden Markov Model (HMM) to support the analysis of nuclear reactor core condition monitoring data. Fuel Grab Load Trace (FGLT) data gathered within the UK during routine refueling operations has been seen to provide information relating to the condition of the graphite bricks that comprise the core. Although manual analysis of this data is time consuming and requires considerable expertise, this paper demonstrates how techniques such as the HMM can provide analysis support by providing a benchmark model of expected behavior against which future refueling events may be compared. The presence of anomalous behavior in candidate traces is inferred through the underlying statistical foundation of the HMM which gives an observation likelihood averaged along the length of the input sequence. Using this likelihood measure, the engineer can be alerted to anomalous behaviour, indicating data which might require further detailed examination. It is proposed that this data analysis technique is used in conjunction with other intelligent analysis techniques currently employed to analyse FGLT to provide a greater confidence measure in detecting anomalous behaviour from FGLT data.

  2. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    Energy Technology Data Exchange (ETDEWEB)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki [Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan) and Advanced Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Numazu College of Technology, Ooka 3600, Numazu, Shizuoka 410-8501 (Japan); Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan)

    2012-11-12

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to {approx} 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  3. Roles of nuclear weak rates on the evolution of degenerate cores in stars

    Directory of Open Access Journals (Sweden)

    Suzuki Toshio

    2017-01-01

    Full Text Available Electron-capture and β-decay rates in stellar environments are evaluated with the use of new shell-model Hamiltonians for sd-shell and pf-shell nuclei as well as for nuclei belonging to the island of inversion. Important role of the nuclear weak rates on the final evolution of stellar degenerate cores is presented. The weak interaction rates for sd-shell nuclei are calculated to study nuclear Urca processes in O-Ne-Mg cores of stars with 8-10 M⊙ (solar mass and their effects on the final fate of the stars. Nucleosynthesis of iron-group elements in Type Ia supernova explosions are studied with the weak rates for pf-shell nuclei. The problem of the neutron-rich iron-group isotope over-production compared to the solar abundances is shown to be nearly solved with the use of the new rates and explosion model of slow defraglation with delayed detonation. Evaluation of the weak rates is extended to the island of inversion and the region of neutron-rich nuclei near 78Ni, where two major shells contribute to their configurations.

  4. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    International Nuclear Information System (INIS)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2012-01-01

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ∼ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  5. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    Science.gov (United States)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2012-11-01

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ~ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  6. The Great Deluge Algorithm applied to a nuclear reactor core design optimization problem

    International Nuclear Information System (INIS)

    Sacco, Wagner F.; Oliveira, Cassiano R.E. de

    2005-01-01

    The Great Deluge Algorithm (GDA) is a local search algorithm introduced by Dueck. It is an analogy with a flood: the 'water level' rises continuously and the proposed solution must lie above the 'surface' in order to survive. The crucial parameter is the 'rain speed', which controls convergence of the algorithm similarly to Simulated Annealing's annealing schedule. This algorithm is applied to the reactor core design optimization problem, which consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a 3-enrichment-zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. This problem was previously attacked by the canonical genetic algorithm (GA) and by a Niching Genetic Algorithm (NGA). NGAs were designed to force the genetic algorithm to maintain a heterogeneous population throughout the evolutionary process, avoiding the phenomenon known as genetic drift, where all the individuals converge to a single solution. The results obtained by the Great Deluge Algorithm are compared to those obtained by both algorithms mentioned above. The three algorithms are submitted to the same computational effort and GDA reaches the best results, showing its potential for other applications in the nuclear engineering field as, for instance, the nuclear core reload optimization problem. One of the great advantages of this algorithm over the GA is that it does not require special operators for discrete optimization. (author)

  7. Diffusion induced nuclear reactions in metals: a possible source of heat in the core

    International Nuclear Information System (INIS)

    Hamza, V.M.; Iyer, S.S.S.

    1989-01-01

    It has recently been proposed that diffusion of light nuclei in metals can give rise to unusual electrical charge distributions in their lattice structures, inducing thereby certain nuclear reactions that are otherwise uncommon. In the light of these results we advance the hypothesis that such nuclear reactions take place in the metal rich core of the earth, based on following observations: 1 - The solubility of hydrogen in metals is relatively high compared to that in silicates. 2 - Studies of rare gas samples in intraplate volcanos and diamonds show that 3 He/ He ratio increases with depth in the mantle. 3 - There are indications that He is positively correlated with enrichment of metals in lavas. We propose that hydrogen incorporated into metallic phases at the time of planetary accretion was carried to the core by downward migration of metal rich melts during the early states of proto-earth. Preliminary estimates suggest that cold fusion reactions can give rise to an average rate of heat generation of 8.2x10 12 W and may thus serve as a supplementary source of energy for the geomagnetic dynamo. (author)

  8. The gravitational attraction algorithm: a new metaheuristic applied to a nuclear reactor core design optimization problem

    International Nuclear Information System (INIS)

    Sacco, Wagner F.; Oliveira, Cassiano R.E. de

    2005-01-01

    A new metaheuristic called 'Gravitational Attraction Algorithm' (GAA) is introduced in this article. It is an analogy with the gravitational force field, where a body attracts another proportionally to both masses and inversely to their distances. The GAA is a populational algorithm where, first of all, the solutions are clustered using the Fuzzy Clustering Means (FCM) algorithm. Following that, the gravitational forces of the individuals in relation to each cluster are evaluated and this individual or solution is displaced to the cluster with the greatest attractive force. Once it is inside this cluster, the solution receives small stochastic variations, performing a local exploration. Then the solutions are crossed over and the process starts all over again. The parameters required by the GAA are the 'diversity factor', which is used to create a random diversity in a fashion similar to genetic algorithm's mutation, and the number of clusters for the FCM. GAA is applied to the reactor core design optimization problem which consists in adjusting several reactor cell parameters in order to minimize the average peak-factor in a 3-enrichment-zone reactor, considering operational restrictions. This problem was previously attacked using the canonical genetic algorithm (GA) and a Niching Genetic Algorithm (NGA). The new metaheuristic is then compared to those two algorithms. The three algorithms are submitted to the same computational effort and GAA reaches the best results, showing its potential for other applications in the nuclear engineering field as, for instance, the nuclear core reload optimization problem. (author)

  9. Possible generation of heat from nuclear fusion in Earth's inner core.

    Science.gov (United States)

    Fukuhara, Mikio

    2016-11-23

    The cause and source of the heat released from Earth's interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: 2 D +  2 D +  2 D → 2 1 H +  4 He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 10 12  J/m 3 , based on the assumption that Earth's primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth's interior to the universe, and pass through Earth, respectively.

  10. Nuclear reactor core having nuclear fuel and composite burnable absorber arranged for power peaking and moderator temperature coefficient control

    International Nuclear Information System (INIS)

    Kapil, S.K.

    1991-01-01

    This patent describes a nuclear reactor core. It comprises a first group of fuel rods containing fissionable material and being free of burnable absorber material; and a second group of fuel rods containing fissionable material and first and second burnable absorber material; the first burnable absorber material being a boron-bearing material which does not contain erbium and the second burnable absorber material being an erbium material; the first and second burnable absorber materials being in the form of an outer coating on the fissionable material, the outer coating being composed of an inner layer of one of the boron-bearing material which does not contain erbium and the erbium material and an outer layer of the other of the boron-bearing material which does not contain erbium and the erbium material

  11. Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

    International Nuclear Information System (INIS)

    Mitchell, Jonathan; Fordham, Edmund J.

    2014-01-01

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general

  12. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  13. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

  14. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

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

  15. Molten Core - Concrete interactions in nuclear accidents. Theory and design of an experimental facility

    International Nuclear Information System (INIS)

    Sevon, T.

    2005-11-01

    In a hypothetical severe accident in a nuclear power plant, the molten core of the reactor may flow onto the concrete floor of containment building. This would cause a molten core . concrete interaction (MCCI), in which the heat transfer from the hot melt to the concrete would cause melting of the concrete. In assessing the safety of nuclear reactors, it is important to know the consequences of such an interaction. As background to the subject, this publication includes a description of the core melt stabilization concept of the European Pressurized water Reactor (EPR), which is being built in Olkiluoto in Finland. The publication includes a description of the basic theory of the interaction and the process of spalling or cracking of concrete when it is heated rapidly. A literature survey and some calculations of the physical properties of concrete and corium. concrete mixtures at high temperatures have been conducted. In addition, an equation is derived for conservative calculation of the maximum possible concrete ablation depth. The publication also includes a literature survey of experimental research on the subject of the MCCI and discussion of the results and deficiencies of the experiments. The main result of this work is the general design of an experimental facility to examine the interaction of molten metals and concrete. The main objective of the experiments is to assess the probability of spalling, or cracking, of concrete under pouring of molten material. A program of five experiments has been designed, and pre-test calculations of the experiments have been conducted with MELCOR 1.8.5 accident analysis program and conservative analytic calculations. (orig.)

  16. Introduction of virtual detectors for core monitoring system of korean standard nuclear power plant

    International Nuclear Information System (INIS)

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

    2000-01-01

    A novel algorithm known as the virtual detector method (VDM) is introduced to reconstruct the axial power shape (APS) for the on-line core monitoring system of the Korean Standard Nuclear Power Plant (KSNP). A pure statistical method (SM) is also introduced and the results are compared with the currently implemented five-mode Fourier fitting method (FFM). VDM adopts nine virtual detector informations coupled with a regression model based on the Alternating Conditional Expectation (ACE) algorithm. VDM uses Fourier fitting with the information of nine virtual detectors expanded from the currently implemented FFM, which uses five-level detector information. By introducing virtual detectors, we can increase the number of axial detectors, and thus expect the computational errors of APS to be reduced. The two methods (SM and VDM) are applied to in-core mapping data from six cycles of Yong Gwang nuclear power plant Units 3 and 4. For ∼ 3500 cases of APSs extracted from a cycle of operation which is simulated by a three-dimensional nodal code, the accuracy of the three methods (SM, VDM, FFM) is compared. The average root mean square (RMS) error and average of axial peaking error of SM and VDM resulted in reduction of more than 50 % and 70 %, respectively, relative to FFM. VDM and SM also show more realistic axial profiles and predict more accurate axial peaking than FFM. These improvements can contribute to a larger thermal margin. SM shows the most accurate results for all cases. VDM can almost obtain the same results as SM, and using far fewer computation steps. VDM can be a useful tool for precisely reconstructing axial power shapes in a core monitoring system. (authors)

  17. Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 2. BFS-62 analysis results

    International Nuclear Information System (INIS)

    Mantourov, Guennadi

    2001-11-01

    This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. Data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS-62 and ZPPR JUPIER series fast reactor cores, applying JNC core calculation code CITATION-FBR. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the First Flight Collision Probability method and subgroup approach. Especially, two converting programs were written to transmit the prepared effective cross sections to JNC standard PDS files to let then the CITATION code be applied for 3-D HEXZ neutronics calculations of the investigated cores. The effects of nuclear data library have been studied by comparing the results calculated using ABBN-93 nuclear data library with the former ones obtained in JNC based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for 4 BFS-62 cores as well as for 3 ZPPR JUPITER experiment series cores ZPPR-9, ZPPR-13A and ZPPR-17A are presented. The comparison results for reaction rates distributions for 2 BFS-62 uranium loaded cores are included too. The calculated correction factors applied in all cases were less than 1.0%. The estimated uncertainty in k-effective C values caused by possible errors in calculation of the applied corrections is about 0.3% in case of BFS-62 and ZPPR MOX cores, and is about 0.2% for BFS-62 uranium-loaded cores. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC's calculation route for k-effective results is about 0.3% for ZPPR and BFS-62 cores with plutonium. As for BFS uranium-loaded cores (BFS-62-1 and BFS-62-2) the nuclear data library effect is about 0.1%. Next the sensitivity analysis was applied. It shown that the main contributors to the nuclear data library effect

  18. Third-generation site characterization: Cryogenic core collection, nuclear magnetic resonance, and electrical resistivity

    Science.gov (United States)

    Kiaalhosseini, Saeed

    In modern contaminant hydrology, management of contaminated sites requires a holistic characterization of subsurface conditions. Delineation of contaminant distribution in all phases (i.e., aqueous, non-aqueous liquid, sorbed, and gas), as well as associated biogeochemical processes in a complex heterogeneous subsurface, is central to selecting effective remedies. Arguably, a factor contributing to the lack of success of managing contaminated sites effectively has been the limitations of site characterization methods that rely on monitoring wells and grab sediment samples. The overarching objective of this research is to advance a set of third-generation (3G) site characterization methods to overcome shortcomings of current site characterization techniques. 3G methods include 1) cryogenic core collection (C3) from unconsolidated geological subsurface to improve recovery of sediments and preserving key attributes, 2) high-throughput analysis (HTA) of frozen core in the laboratory to provide high-resolution, depth discrete data of subsurface conditions and processes, 3) resolution of non-aqueous phase liquid (NAPL) distribution within the porous media using a nuclear magnetic resonance (NMR) method, and 4) application of a complex resistivity method to track NAPL depletion in shallow geological formation over time. A series of controlled experiments were conducted to develop the C 3 tools and methods. The critical aspects of C3 are downhole circulation of liquid nitrogen via a cooling system, the strategic use of thermal insulation to focus cooling into the core, and the use of back pressure to optimize cooling. The C3 methods were applied at two contaminated sites: 1) F.E. Warren (FEW) Air Force Base near Cheyenne, WY and 2) a former refinery in the western U.S. The results indicated that the rate of core collection using the C3 methods is on the order of 30 foot/day. The C3 methods also improve core recovery and limits potential biases associated with flowing sands

  19. Utilization of cross-section covariance data in FBR core nuclear design and cross-section adjustment

    International Nuclear Information System (INIS)

    Ishikawa, Makoto

    1994-01-01

    In the core design of large fast breeder reactors (FBRs), it is essentially important to improve the prediction accuracy of nuclear characteristics from the viewpoint of both reducing cost and insuring reliability of the plant. The cross-section errors, that is, covariance data are one of the most dominant sources for the prediction uncertainty of the core parameters, therefore, quantitative evaluation of covariance data is indispensable for FBR core design. The first objective of the present paper is to introduce how the cross-section covariance data are utilized in the FBR core nuclear design works. The second is to delineate the cross-section adjustment study and its application to an FBR design, because this improved design method markedly enhances the needs and importance of the cross-section covariance data. (author)

  20. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    International Nuclear Information System (INIS)

    Park, Hae Kyun; Kim, Sang Nyung

    2013-01-01

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was conservatively calculated

  1. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hae Kyun; Kim, Sang Nyung [Kyung Hee Univ., Yongin (Korea, Republic of)

    2013-10-15

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was

  2. Fuel/propellant mixing in an open-cycle gas core nuclear rocket engine

    International Nuclear Information System (INIS)

    Guo, X.; Wehrmeyer, J.A.

    1997-01-01

    A numerical investigation of the mixing of gaseous uranium and hydrogen inside an open-cycle gas core nuclear rocket engine (spherical geometry) is presented. The gaseous uranium fuel is injected near the centerline of the spherical engine cavity at a constant mass flow rate, and the hydrogen propellant is injected around the periphery of the engine at a five degree angle to the wall, at a constant mass flow rate. The main objective is to seek ways to minimize the mixing of uranium and hydrogen by choosing a suitable injector geometry for the mixing of light and heavy gas streams. Three different uranium inlet areas are presented, and also three different turbulent models (k-var-epsilon model, RNG k-var-epsilon model, and RSM model) are investigated. The commercial CFD code, FLUENT, is used to model the flow field. Uranium mole fraction, axial mass flux, and radial mass flux contours are obtained. copyright 1997 American Institute of Physics

  3. FLICA-4 (version 1). A computer code for three dimensional thermal analysis of nuclear reactor cores

    International Nuclear Information System (INIS)

    Raymond, P.; Allaire, G.; Boudsocq, G.; Caruge, D.; Gramont, T. de; Toumi, I.

    1995-01-01

    FLICA-4 is a thermal-hydraulic computer code, developed at the French Atomic Energy Commission (CEA) for three-dimensional steady-state or transient two-phase flow, and aimed at design and safety thermal analysis of nuclear reactor cores. It is available for various UNIX workstations and CRAY computers under UNICOS.It is based on four balance equations which include three balance equations for the mixture and a mass balance equation for the less concentrated phase which allows for the calculation of non equilibrium flows such as sub-cooled boiling and superheated steam. A drift velocity model takes into account the velocity unbalance between phases. The equations are solved using a finite volume numerical scheme. Typical running time, specific features (coupling with other codes) and auxiliary programs are presented. 1 tab., 9 refs

  4. Reproducing cultural identity in negotiating nuclear power: the Union of Concerned Scientists and emergency core cooling

    International Nuclear Information System (INIS)

    Downey, G.L.

    1988-01-01

    This paper advances the concept of 'cultural identity' to account for the nexus between structure and practice in technological negotiations. It describes how the formation of the Union of Concerned Scientists (UCS), and that group's subsequent discourse and nonverbal actions, both reproduced the established identities of group members and contributed to negotiations that reconstituted those identities. In particular, UCS claims about emergency core-cooling systems in nuclear plants were congruent with the combination of a shared ideology, the social interests of Massachusetts Institute of Technology faculty, and established principles of engineering design. The cultural analysis of identity reproduction shows the opposition between cognitive and social phenomena to be a significant distinction framing action in Western culture. The analysis also suggests that new attention be given to the relationship between the constitutive and reproductive functions of discourse and nonverbal action. (author)

  5. How did Fukushima-Dai-ichi core meltdown change the probability of nuclear accidents?

    International Nuclear Information System (INIS)

    Escobar Rangel, Lina; Leveque, Francois

    2012-10-01

    How to predict the probability of a nuclear accident using past observations? What increase in probability the Fukushima Dai-ichi event does entail? Many models and approaches can be used to answer these questions. Poisson regression as well as Bayesian updating are good candidates. However, they fail to address these issues properly because the independence assumption in which they are based on is violated. We propose a Poisson Exponentially Weighted Moving Average (PEWMA) based in a state-space time series approach to overcome this critical drawback. We find an increase in the risk of a core meltdown accident for the next year in the world by a factor of ten owing to the new major accident that took place in Japan in 2011. (authors)

  6. Reproducing cultural identity in negotiating nuclear power: the Union of Concerned Scientists and emergency core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Downey, G L

    1988-05-01

    This paper advances the concept of 'cultural identity' to account for the nexus between structure and practice in technological negotiations. It describes how the formation of the Union of Concerned Scientists (UCS), and that group's subsequent discourse and nonverbal actions, both reproduced the established identities of group members and contributed to negotiations that reconstituted those identities. In particular, UCS claims about emergency core-cooling systems in nuclear plants were congruent with the combination of a shared ideology, the social interests of Massachusetts Institute of Technology faculty, and established principles of engineering design. The cultural analysis of identity reproduction shows the opposition between cognitive and social phenomena to be a significant distinction framing action in Western culture. The analysis also suggests that new attention be given to the relationship between the constitutive and reproductive functions of discourse and nonverbal action.

  7. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    International Nuclear Information System (INIS)

    Griffiths, M.; Coleman, C.E.; Holt, R.A.; Sagat, S.; Urbanic, V.F.; Chow, C.K.

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and β-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the α-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of 'breakaway' growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs

  8. Hualong One's nuclear reactor core design and relative safety issues research

    Energy Technology Data Exchange (ETDEWEB)

    Yu, H., E-mail: yuhong_xing@126.com [Nuclear Power Inst. of China, Design and Research Sub-Inst., Chengdu, Sichuan (China)

    2015-07-01

    'Full text:' Hualong One, a third generation 1000MWe-class pressurized water reactor, is developed by China National Nuclear Cooperation (CNNC), based on the self-reliant technologies and experiences from China 40 years designing, construction, operation and maintenance of NPPs. In China, it has been approved to construct at Fuqing 5&6 and Fangchenggang 3&4. The Hualong One adopts advanced design features to dramatically enhance plant safety, economic efficiency and convenience of operation and maintenance. It consists of three loops with nominal thermal power output 3060 MWt and a 60-year design life. Its reactor core has 177 fuel assemblies, 18 month refueling interval (after initial cycle), and more than 15% thermal margin. It adopts low leakage loading pattern which can achieve better economy of the neutron, higher reactivity and lower radiation damage of pressure vessel. For the safety design, incorporating the feedback of Fukushima accident, the Hualong One has a combination of active and passive safety systems, a single station layout, double containment structure, and comprehensive implementation of defence-in-depth design principles. The new design features has been successfully evaluated to ensure that they enhance the performance and safety of Hualong One. Several experimental activates have been conducted, such as cavity injection and cooling system testing, passive containment heat removal system testing, and passive residual heat removal system of secondary side testing. The future improvements of Hualong reactor will focus on better economic core design and more reliable safety system. (author)

  9. Evaluation of the need for stochastic optimization of out-of-core nuclear fuel management decisions

    International Nuclear Information System (INIS)

    Thomas, R.L. Jr.

    1989-01-01

    Work has been completed on utilizing mathematical optimization techniques to optimize out-of-core nuclear fuel management decisions. The objective of such optimization is to minimize the levelized fuel cycle cost over some planning horizon. Typical decision variables include feed enrichments and number of assemblies, burnable poison requirements, and burned fuel to reinsert for every cycle in the planning horizon. Engineering constraints imposed consist of such items as discharge burnup limits, maximum enrichment limit, and target cycle energy productions. Earlier the authors reported on the development of the OCEON code, which employs the integer Monte Carlo Programming method as the mathematical optimization method. The discharge burnpups, and feed enrichment and burnable poison requirements are evaluated, initially employing a linear reactivity core physics model and refined using a coarse mesh nodal model. The economic evaluation is completed using a modification of the CINCAS methodology. Interest now is to assess the need for stochastic optimization, which will account for cost components and cycle energy production uncertainties. The implication of the present studies is that stochastic optimization in regard to cost component uncertainties need not be completed since deterministic optimization will identify nearly the same family of near-optimum cycling schemes

  10. Development of advanced nuclear core analysis system applicable to various reactor types

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2002-03-01

    This fiscal year, aiming at development of an advanced detailed analysis system applicable to nuclear core performance analysis of various fast reactors currently considered, the concept of cross section library set was examined and the specification of library set was determined. That is to say, referring the world most advanced reactor physics analysis system ERANOS (European Reactor Analysis Optimized System) and the result of preceding research 'preparation of next generation cross section library', 900 energy groups structure, concrete cross section data to be included and the format of cross section library were defined. And we performed elaborate work revising the group cross section production system which was prepared in the preceding research. After that the revision work was completed, to confirm the capability of revised cross section production system, we produced a prototype 450 groups cross section library. And we carried out a series of bench mark tests including analysis of small fast reactors utilizing this prototype cross section library and confirmed that the prototype cross section library has sufficient accuracy for predicting core performance. Furthermore, we estimated the computer resource information such as memory size, hard disk capacity and calculation time, etc. necessary for producing 900 groups detailed cross section library. In addition, we identified problems to be solved for developing a cell calculation code installed in our detailed analysis system. (author)

  11. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, M; Coleman, C E; Holt, R A; Sagat, S; Urbanic, V F [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.; Chow, C K [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Nuclear Research Establishment

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and {beta}-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the {alpha}-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of `breakaway` growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs.

  12. Assessment of water hammer effects on boiling water nuclear reactor core dynamics

    Directory of Open Access Journals (Sweden)

    Bousbia-Salah Anis

    2007-01-01

    Full Text Available Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

  13. Development of advanced nuclear core analysis system applicable to various reactor types (II)

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2003-03-01

    A 900 group cross section library based on the specification determined last year was produced for 27 nuclei of the fast reactor benchmark problem evaluated in nuclear data file JENDL-3.2. In addition, the new SLAROM code, which has been developed as an advanced detail analysis system, was revised so as to make cell calculations effectively with the above 900 group library. Furthermore, new functions were added to the SLAROM so that the SLAROM evaluates assembly parameters using effective cross sections derived by the SLAROM and produces any condensed effective cross section set for core performance analysis. With the 900 group cross section library and the revised SALROM, three cell calculations for fast and medium neutron speed reactors having different neutron spectrum were performed, and the results were compared with those calculated by the continuos energy Monte Carlo code MVP. By the comparisons, it is concluded that the newly revised SLAROM and a 900 group cross section library give accuracy comparable to MVP for predicting core performances. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-15

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

  16. Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

    Science.gov (United States)

    Huddar, Lakshana Ravindranath

    With electricity demand predicted to rise by more than 50% within the next 20 years and a burgeoning world population requiring reliable emissions-free base-load electricity, can we design advanced nuclear reactors to help meet this challenge? At the University of California, Berkeley (UCB) Fluoride-salt-cooled High Temperature Reactors (FHR) are currently being investigated. FHRs are designed with better safety and economic characteristics than conventional light water reactors (LWR) currently in operation. These reactors operate at high temperature and low pressure making them more efficient and safer than LWRs. The pebble-bed FHR (PB-FHR) variant includes an annular nuclear reactor core that is filled with randomly packed pebble fuel. It is crucial to characterize the heat transfer within this unique geometry as this informs the safety limits of the reactor. The work presented in this dissertation focused on furthering the understanding of heat transfer in pebble-bed nuclear reactor cores using fluoride salts as a coolant. This was done through experimental, analytical and computational techniques. A complex nuclear system with a coolant that has never previously been in commercial use requires experimental data that can directly inform aspects of its design. It is important to isolate heat transfer phenomena in order to understand the underlying physics in the context of the PB-FHR, as well as to make decisions about further experimental work that needs to be done in support of developing the PB-FHR. Certain organic oils can simulate the heat transfer behaviour of the fluoride salt if relevant non-dimensional parameters are matched. The advantage of this method is that experiments can be done at a much lower temperature and at a smaller geometric scale compared to FHRs, thereby lowering costs. In this dissertation, experiments were designed and performed to collect data demonstrating similitude. The limitations of these experiments were also elucidated by

  17. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  18. CORE DESIGNS OF ABWR FOR PROPOSED OF THE FIRST NUCLEAR POWER PLANT IN INDONESIA

    Directory of Open Access Journals (Sweden)

    Yohannes Sardjono

    2015-04-01

    Full Text Available Indonesia as an archipelago has been experiencing high growth industry and energy demand due to high population growth, dynamic economic activities. The total population is around 230 million people and 75 % to the total population is living in Java. The introduction of Nuclear Power Plant on Java Bali electricity grid will be possible in 2022 for 2 GWe, using proven technology reactor like ABWR or others light water reactor with nominal power 1000 MWe. In this case, the rated thermal power for the equilibrium cycles is 3926 MWt, the cycle length is 18 month and overall capacity factor is 87 %. The designs were performed for an 872-fuel bundles ABWR core using GE-11 fuel type in an 9×9 fuel rod arrays with 2 Large Central Water Rods (LCWR. The calculations were divided into two steps; the first is to generate bundle library and the other is to make the thermal and reactivity limits satisfied for the core designs. Toshiba General Electric Bundle lattice Analysis (TGBLA and PANACEA computer codes were used as designs tools. TGBLA is a General Electric proprietary computer code which is used to generate bundle lattice library for fuel designs. PANACEA is General Electric proprietary computer code which is used as thermal hydraulic and neutronic coupled BWR core simulator. This result of core designs describes reactivity and thermal margins i.e.; Maximum Linear Heat Generation rate (MLHGR is lower than 14.4 kW/ft, Minimum Critical Power Ratio (MCPR is upper than 1.25, Hot Excess Reactivity (HOTXS is upper than 1 %Dk at BOC and 0.8 %Dk at 200 MWD/ST and Cold Shutdown Margin Reactivity (CSDM is upper than 1 %Dk. It is concluded that the equilibrium core design using GE-11 fuel bundle type satisfies the core design objectives for the proposed of the firs Indonesia ABWR Nuclear Power Plant. Keywords: The first NPP in Indonesia, ABWR-1000 MWe, and core designs.   Indonesia adalah sebagai negara kepulauan yang laju pertumbuhan industri, energi, penduduk

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

    International Nuclear Information System (INIS)

    Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono

    2014-01-01

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

  20. Verification Results of Safety-grade Optical Modem for Core Protection Calculator (CPC) in Korea Standard Nuclear Power Plant (KSNP)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jangyeol; Son, Kwangseop; Lee, Youngjun; Cheon, Sewoo; Cha, Kyoungho; Lee, Jangsoo; Kwon, Keechoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    We confirmed that the coverage criteria for a safety-grade optical modem of a Core Protection Calculator is satisfactory using a traceability analysis matrix between high-level requirements and lower-level system test case data set. This paper describes the test environment, test components and items, a traceability analysis, and system tests as a result of system verification and validation based on Software Requirement Specifications (SRS) for a safety-grade optical modem of a Core Protection Calculator (CPC) in a Korea Standard Nuclear Power Plant (KSNP), and Software Design Specifications (SDS) for a safety-grade optical modem of a CPC in a KSNP. All tests were performed according to the test plan and test procedures. Functional testing, performance testing, event testing, and scenario based testing for a safety-grade optical modem of a Core Protection Calculator in a Korea Standard Nuclear Power Plant as a thirty-party verifier were successfully performed.

  1. Development of a multi-objective PBIL evolutionary algorithm applied to a nuclear reactor core reload optimization problem

    International Nuclear Information System (INIS)

    Machado, Marcelo D.; Dchirru, Roberto

    2005-01-01

    The nuclear reactor core reload optimization problem consists in finding a pattern of partially burned-up and fresh fuels that optimizes the plant's next operation cycle. This optimization problem has been traditionally solved using an expert's knowledge, but recently artificial intelligence techniques have also been applied successfully. The artificial intelligence optimization techniques generally have a single objective. However, most real-world engineering problems, including nuclear core reload optimization, have more than one objective (multi-objective) and these objectives are usually conflicting. The aim of this work is to develop a tool to solve multi-objective problems based on the Population-Based Incremental Learning (PBIL) algorithm. The new tool is applied to solve the Angra 1 PWR core reload optimization problem with the purpose of creating a Pareto surface, so that a pattern selected from this surface can be applied for the plant's next operation cycle. (author)

  2. In-medium no-core shell model for ab initio nuclear structure calculations

    International Nuclear Information System (INIS)

    Gebrerufael, Eskendr

    2017-01-01

    In this work, we merge two successful ab initio nuclear-structure methods, the no-core shell model (NCSM) and the multi-reference in-medium similarity renormalization group (IM-SRG), to define a novel many-body approach for the comprehensive description of ground and excited states of closed- and open-shell medium-mass nuclei. Building on the key advantages of the two methods - the decoupling of excitations at the many-body level in the IM-SRG, and the exact diagonalization in the NCSM applicable up to medium-light nuclei - their combination enables fully converged no-core calculations for an unprecedented range of nuclei and observables at moderate computational cost. The efficiency and rapid model-space convergence of the new approach make it ideally suited for ab initio studies of ground and low-lying excited states of nuclei up to the medium-mass regime. Interactions constructed within the framework of chiral effective field theory provide an excellent opportunity to describe properties of nuclei from first principles, i.e., rooted in quantum chromodynamics, they overcome the lack of predictive power of phenomenological potentials. The hard core of these interactions causes strong short-range correlations, which we soften by using the similarity-renormalization-group transformation that accelerates the model-space convergence of many-body calculations. Three-nucleon effects, which are mandatory for the correct description of bulk properties of nuclei, are included in our calculations by using the normal-ordered two-body approximation, which has been shown to be sufficient to capture the main effects of the three-nucleon interaction. Using these interactions, we analyze energies of ground and excited states in the carbon and oxygen isotopic chains, where conventional NCSM calculations are still feasible and provide an important benchmark. Furthermore, we study the Hoyle state in 12 C - a three-alpha cluster state that cannot be converged in standard NCSM

  3. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Boer, Brian

    2007-01-01

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  4. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  5. FLICA-4 (version 1) a computer code for three dimensional thermal analysis of nuclear reactor cores

    International Nuclear Information System (INIS)

    Raymond, P.; Allaire, G.; Boudsocq, G.

    1995-01-01

    FLICA-4 is a thermal-hydraulic computer code developed at the French Energy Atomic Commission (CEA) for three dimensional steady state or transient two phase flow for design and safety thermal analysis of nuclear reactor cores. The two phase flow model of FLICA-4 is based on four balance equations for the fluid which includes: three balance equations for the mixture and a mass balance equation for the less concentrated phase which permits the calculation of non-equilibrium flows as sub cooled boiling and superheated steam. A drift velocity model takes into account the velocity disequilibrium between phases. The thermal behaviour of fuel elements can be computed by a one dimensional heat conduction equation in plane, cylindrical or spherical geometries and coupled to the fluid flow calculation. Convection and diffusion of solution products which are transported either by the liquid or by the gas, can be evaluated by solving specific mass conservation equations. A one dimensional two phase flow model can also be used to compute 1-D flow in pipes, guide tubes, BWR assemblies or RBMK channels. The FLICA-4 computer code uses fast running time steam-water functions. Phasic and saturation physical properties are computed by using bi-cubic spline functions. Polynomial coefficients are tabulated from 0.1 to 22 MPa and 0 to 800 degrees C. Specific modules can be utilised in order to generate the spline coefficients for any other fluid properties

  6. A hybrid niched-island genetic algorithm applied to a nuclear core optimization problem

    International Nuclear Information System (INIS)

    Pereira, Claudio M.N.A.

    2005-01-01

    Diversity maintenance is a key-feature in most genetic-based optimization processes. The quest for such characteristic, has been motivating improvements in the original genetic algorithm (GA). The use of multiple populations (called islands) has demonstrating to increase diversity, delaying the genetic drift. Island Genetic Algorithms (IGA) lead to better results, however, the drift is only delayed, but not avoided. An important advantage of this approach is the simplicity and efficiency for parallel processing. Diversity can also be improved by the use of niching techniques. Niched Genetic Algorithms (NGA) are able to avoid the genetic drift, by containing evolution in niches of a single-population GA, however computational cost is increased. In this work it is investigated the use of a hybrid Niched-Island Genetic Algorithm (NIGA) in a nuclear core optimization problem found in literature. Computational experiments demonstrate that it is possible to take advantage of both, performance enhancement due to the parallelism and drift avoidance due to the use of niches. Comparative results shown that the proposed NIGA demonstrated to be more efficient and robust than an IGA and a NGA for solving the proposed optimization problem. (author)

  7. An intelligent nuclear reactor core controller for load following operations, using recurrent neural networks and fuzzy systems

    International Nuclear Information System (INIS)

    Boroushaki, M.; Ghofrani, M.B.; Lucas, C.; Yazdanpanah, M.J.

    2003-01-01

    In the last decade, the intelligent control community has paid great attention to the topic of intelligent control systems for nuclear plants (core, steam generator...). Papers mostly used approximate and simple mathematical SISO (single-input-single-output) model of nuclear plants for testing and/or tuning of the control systems. They also tried to generalize theses models to a real MIMO (multi-input-multi-output) plant, while nuclear plants are typically of complex nonlinear and multivariable nature with high interactions between their state variables and therefore, many of these proposed intelligent control systems are not appropriate for real cases. In this paper, we designed an on-line intelligent core controller for load following operations, based on a heuristic control algorithm, using a valid and updatable recurrent neural network (RNN). We have used an accurate 3-dimensional core calculation code to represent the real plant and to train the RNN. The results of simulation show that this intelligent controller can control the reactor core during load following operations, using optimum control rod groups manoeuvre and variable overlapping strategy. This methodology represents a simple and reliable procedure for controlling other complex nonlinear MIMO plants, and may improve the responses, comparing to other control systems

  8. Methodology for Identification of the Coolant Thermalhydraulic Regimes in the Core of Nuclear Reactors

    International Nuclear Information System (INIS)

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

    2002-01-01

    The paper deals with one of the acute for the nuclear energy problem of accident regimes of NPPs recognition diagnostics using noise signal diagnostics methodology. The methodology intends transformation of the random noise signals of the main technological parameters at the exit of a nuclear facility (neutron flow, dynamic pressure etc.) which contain the important information about the technical status of the equipment. The effective algorithms for identification of random processes wore developed. After proper transformation its were considered as multidimensional random vectors. Automatic classification of these vectors in the developed algorithms is realized on the basis of the probability function in particular Bayes classifier and decision functions. Till now there no mathematical models for thermalhydraulic regimes of fuel assemblies recognition on the acoustic and neutron noises parameters in the core of nuclear facilities. The two mathematical models for analysis of the random processes submitted to the automatic classification is proposed, i.e. statistical (using Bayes classifier of acoustic spectral density diagnosis signals) and geometrical (on the basis of formation in the featured space of dividing hyper-plane). The theoretical basis of the bubble boiling regimes in the fuel assemblies is formulated as identification of these regimes on the basis of random parameters of auto spectral density of acoustic noise (ASD) measured in the fuel assemblies (dynamic pressure in the upper plenum in the paper). The elaborated algorithms allow recognize realistic status of the fuel assemblies. For verification of the proposed mathematical models the analysis of experimental measurements was carried out. The research of the boiling onset and definition of the local values of the flow parameters in the seven-beam fuel assembly (length of 1.3 m, diameter of 6 mm) have shown the correct identification of the bubble boiling regimes. The experimental measurements on

  9. A new equation of state for core-collapse supernovae based on realistic nuclear forces and including a full nuclear ensemble

    International Nuclear Information System (INIS)

    Furusawa, S; Togashi, H; Nagakura, H; Sumiyoshi, K; Yamada, S; Suzuki, H; Takano, M

    2017-01-01

    We have constructed a nuclear equation of state (EOS) that includes a full nuclear ensemble for use in core-collapse supernova simulations. It is based on the EOS for uniform nuclear matter that two of the authors derived recently, applying a variational method to realistic two- and three-body nuclear forces. We have extended the liquid drop model of heavy nuclei, utilizing the mass formula that accounts for the dependences of bulk, surface, Coulomb and shell energies on density and/or temperature. As for light nuclei, we employ a quantum-theoretical mass evaluation, which incorporates the Pauli- and self-energy shifts. In addition to realistic nuclear forces, the inclusion of in-medium effects on the full ensemble of nuclei makes the new EOS one of the most realistic EOSs, which covers a wide range of density, temperature and proton fraction that supernova simulations normally encounter. We make comparisons with the FYSS EOS, which is based on the same formulation for the nuclear ensemble but adopts the relativistic mean field theory with the TM1 parameter set for uniform nuclear matter. The new EOS is softer than the FYSS EOS around and above nuclear saturation densities. We find that neutron-rich nuclei with small mass numbers are more abundant in the new EOS than in the FYSS EOS because of the larger saturation densities and smaller symmetry energy of nuclei in the former. We apply the two EOSs to 1D supernova simulations and find that the new EOS gives lower electron fractions and higher temperatures in the collapse phase owing to the smaller symmetry energy. As a result, the inner core has smaller masses for the new EOS. It is more compact, on the other hand, due to the softness of the new EOS and bounces at higher densities. It turns out that the shock wave generated by core bounce is a bit stronger initially in the simulation with the new EOS. The ensuing outward propagations of the shock wave in the outer core are very similar in the two simulations, which

  10. A core concept for the self-consistent nuclear energy system based on the promising future technology

    International Nuclear Information System (INIS)

    Arie, K.; Suzuki, M.; Kawashima, M.; Igashira, M.; Shimizu, A.; Fujii-e, Y.

    1995-01-01

    Feasibility of FP burning while maintaining fuel breeding capability for the Self-Consistent Nuclear Energy System is evaluated through neutron balance and a fast reactor core. It is shown that all radioactive FPs produced by itself can be burnt by a fast reactor while maintaining breeding capability, assuming separation of radioactive FP and stable FP isotopes. Assuming that the recovery system of fuel and FPs to be burnt is based on a pyro-chemical process, the major long-lived FPs of I, Pd, Tc, Sn, Se can be burnt with keeping breeding capability by suitability arranging materials in the fast reactor core. (Author)

  11. Observation of the state of the nuclear reactor core by means of non-linear observation algorithms

    International Nuclear Information System (INIS)

    Maciel Palacio, F.E.; Espana, M.D.

    1990-01-01

    A combined, variable-adaptive structure, non-linear observer was designed in order to observe the state of the nuclear reactor core, based on the Absolute Stability Theory. The observer was proved under noise and modelling error conditions. Successful results were obtained in the observation of the states in both cases, showing clear improvement in the observation due to the application of adaptive and variable structure ideas. (Author) [es

  12. In-reactor testing of the closed cycle gas core reactor---the nuclear light bulb concept

    International Nuclear Information System (INIS)

    Gauntt, R.O.; Slutz, S.A.; Harms, G.A.; Latham, T.S.; Roman, W.C.; Rodgers, R.J.

    1993-01-01

    The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>1800 s) and thrust (>445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (∼4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept

  13. Conductive core of radiation-resistant high-pressure electric bushing, especially for nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    Zajic, V

    1981-09-01

    A radiation-resistant high-pressure electric bushing was developed featuring a conductive core consisting of a hollow moulding. At the point of attachment to the bushing insulator the core moulding is widened, thus forming a ring support of a diameter larger by at least 10% than the diameter of the conductive core cylindrical section. On the outer side of the pressure body the core cavity is narrowed and tightly closed with the conductor. On the side facing the medium of higher pressure, the conductive core is provided with a thread. Core manufacture and connection of the conductor to the bushing is very simple. The bushing can be used for an environment with pressures exceeding 10 MPa.

  14. Conductive core of radiation-resistant high-pressure electric bushing, especially for nuclear technology

    International Nuclear Information System (INIS)

    Zajic, V.

    1981-01-01

    A radiation-resistant high-pressure electric bushing was developed featuring a conductive core consisting of a hollow moulding. At the point of attachment to the bushing insulator the core moulding is widened, thus forming a ring support of a diameter larger by at least 10% than the diameter of the conductive core cylindrical section. On the outer side of the pressure body the core cavity is narrowed and tightly closed with the conductor. On the side facing the medium of higher pressure, the conductive core is provided with a thread. Core manufacture and connection of the conductor to the bushing is very simple. The bushing can be used for an environment with pressures exceeding 10 MPa. (J.B.)

  15. Nuclear reactor core having nuclear fuel and composite burnable absorber arranged for power peaking and moderator temperature coefficient control

    International Nuclear Information System (INIS)

    Kapil, S.K.

    1992-01-01

    This patent describes a burnable absorber coated nuclear fuel. It comprises a nuclear fuel substrate containing a fissionable material; and an outer burnable absorber coating applied on an outer surface of the substrate; the outer absorber coating being composed of an inner layer of a boron-bearing material except for erbium boride and an outer layer of an erbium material

  16. ORIGEN2.1 Cycle Specific Calculation of Krsko Nuclear Power Plant Decay Heat and Core Inventory

    International Nuclear Information System (INIS)

    Vukovic, J.; Grgic, D.; Konjarek, D.

    2010-01-01

    This paper presents ORIGEN2.1 computer code calculation of Krsko Nuclear Power Plant core for Cycle 24. The isotopic inventory, core activity and decay heat are calculated in one run for the entire core using explicit depletion and decay of each fuel assembly. Separate pre-ori application which was developed is utilized to prepare corresponding ORIGEN2.1 inputs. This application uses information on core loading pattern to determine fuel assembly specific depletion history using 3D burnup which is obtained from related PARCS computer code calculation. That way both detailed single assembly calculations as well as whole core inventory calculations are possible. Because of the immense output of the ORIGEN2.1, another application called post-ori is used to retrieve and plot any calculated property on the basis of nuclide, element, summary isotope or group of elements for activation products, actinides and fission products segments. As one additional possibility, with the post-ori application it is able to calculate radiotoxicity from calculated ORIGEN2.1 inventory. The results which are obtained using the calculation model of ORIGEN2.1 computer code are successfully compared against corresponding ORIGEN-S computer code results.(author).

  17. Effective height of the core of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Binh, Nguyen Duc; Tuan, Nguyen Minh; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam); Martin, D P; Yip, F G [High Institute of Nuclear Sciences and Technology (Cuba)

    1994-10-01

    Measurements of thermal neutron relative distributions in axial direction at different positions in the reactor core and for various control rod configurations have been carried out, and axial buckling and effective height of the core deduced. (author). 4 refs., 3 figs., 1 tab.

  18. Planning of the development of the MMIS core technology based on nuclear-IT convergence

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Kee Choon; Kim, Chang Hwoi; Hwang, In Koo [KAERI, Daejeon (Korea, Republic of); and others

    2012-01-15

    - Drive nuclear-IT convergence technologies such as middleware applied new concept nuclear instrumentation and control architecture, automated operation of future nuclear power plant, virtual reality/augmented reality, design and verification technology of a nuclear power plant main control room, software dependability, and cyber security technology - Write state-of-the-art report for the nuclear instrumentation and control based on IT convergence - A prototype which implemented related equipment and software subject to nuclear reactor operator that reside in the main control room (Reactor Operator, RO) order to a on-site operator (Local Operator, LO) and confirm the task performance matches the RO's intention - 'IT Convergence intelligent instrumentation and control technology' project planning for the Fourth Nuclear Power Research and Development in the long-term plan.

  19. Planning of the development of the MMIS core technology based on nuclear-IT convergence

    International Nuclear Information System (INIS)

    Kwon, Kee Choon; Kim, Chang Hwoi; Hwang, In Koo

    2012-01-01

    - Drive nuclear-IT convergence technologies such as middleware applied new concept nuclear instrumentation and control architecture, automated operation of future nuclear power plant, virtual reality/augmented reality, design and verification technology of a nuclear power plant main control room, software dependability, and cyber security technology - Write state-of-the-art report for the nuclear instrumentation and control based on IT convergence - A prototype which implemented related equipment and software subject to nuclear reactor operator that reside in the main control room (Reactor Operator, RO) order to a on-site operator (Local Operator, LO) and confirm the task performance matches the RO's intention - 'IT Convergence intelligent instrumentation and control technology' project planning for the Fourth Nuclear Power Research and Development in the long-term plan

  20. Utilization of local area network technology and decentralized structure for nuclear reactor core temperature monitoring

    International Nuclear Information System (INIS)

    Casella, M.; Peirano, F.

    1986-01-01

    The present system concerns Superphenix type reactors. It is a new version of system for monitoring the reactor core temperatures. It has been designed to minimize the cost and the wiring complexity because of the large number of channels (800). For this, equipments are arranged on the roof slab of the reactor with a single link to the control room; from which the name Integrated Treatment of Core Temperatures: TITC 1500 and the natural choice of a distributed system. This system monitors permanently the thermal state of the core a Superphenix type reactor. This monitoring system aims at detecting anomalies of core temperature rise, releasing automatic shutdown (safety), and providing to the monitoring systems not concerned safety the information concerning the core [fr

  1. Theoretical analysis and numerical modelling of heat transfer and fuel migration in underlying soils and constructive elements of nuclear plants during an accident release from the core

    International Nuclear Information System (INIS)

    Arutunjan, R.V.; Bolshov, L.A.; Vitukov, V.V.; Goloviznin, V.M.; Dykhne, A.M.; Kiselev, V.P.; Klementova, S.V.; Krayushkin, I.E.; Moskovchenko, A.V.; Pismennii, V.D.; Popkov, A.G.; Chernov, S.Y.; Chudanov, V.V.; Khoruzhii, O.V.; Yudin, A.I.

    1990-01-01

    Migration of fuel fragments and core fission products during severe accidents on nuclear plants is studied analytically and numerically. The problems of heat transfer and migration of volume heat sources in construction materials and underlying soils are considered

  2. Nuclear data for fission reactor core design and safety analysis: Requirements and status of accuracy of nuclear data

    International Nuclear Information System (INIS)

    Rowlands, J.L.

    1984-01-01

    The types of nuclear data required for fission reactor design and safety analysis, and the ways in which the data are represented and approximated for use in reactor calculations, are summarised first. The relative importance of different items of nuclear data in the prediction of reactor parameters is described and ways of investigating the accuracy of these data by evaluating related integral measurements are discussed. The use of sensitivity analysis, together with estimates of the uncertainties in nuclear data and relevant integral measurements, in assessing the accuracy of prediction of reactor parameters is described. The inverse procedure for deciding nuclear data requirements from the target accuracies for prediction of reactor parameters follows on from this. The need for assessments of the uncertainties in nuclear data evaluations and the form of the uncertainty information is discussed. The status of the accuracies of predictions and nuclear data requirements are then summarised. The reactor parameters considered include: (a) Criticality conditions, conversion and burn-up effects. (b) Energy production and deposition, decay heating, irradiation damage, dosimetry and induced radioactivity. (c) Kinetics characteristics and control, including temperature, power and coolant density coefficients, delayed neutrons and control absorbers. (author)

  3. Evaluation of nuclear power plant component failure probability and core damage probability using simplified PSA model

    International Nuclear Information System (INIS)

    Shimada, Yoshio

    2000-01-01

    It is anticipated that the change of frequency of surveillance tests, preventive maintenance or parts replacement of safety related components may cause the change of component failure probability and result in the change of core damage probability. It is also anticipated that the change is different depending on the initiating event frequency or the component types. This study assessed the change of core damage probability using simplified PSA model capable of calculating core damage probability in a short time period, which is developed by the US NRC to process accident sequence precursors, when various component's failure probability is changed between 0 and 1, or Japanese or American initiating event frequency data are used. As a result of the analysis, (1) It was clarified that frequency of surveillance test, preventive maintenance or parts replacement of motor driven pumps (high pressure injection pumps, residual heat removal pumps, auxiliary feedwater pumps) should be carefully changed, since the core damage probability's change is large, when the base failure probability changes toward increasing direction. (2) Core damage probability change is insensitive to surveillance test frequency change, since the core damage probability change is small, when motor operated valves and turbine driven auxiliary feed water pump failure probability changes around one figure. (3) Core damage probability change is small, when Japanese failure probability data are applied to emergency diesel generator, even if failure probability changes one figure from the base value. On the other hand, when American failure probability data is applied, core damage probability increase is large, even if failure probability changes toward increasing direction. Therefore, when Japanese failure probability data is applied, core damage probability change is insensitive to surveillance tests frequency change etc. (author)

  4. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach; Simulacao termohidraulica do nucleo do reator nuclear HTR-10 com o uso da abordagem realistica CFD

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S.; Dominguez, Dany S., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil); Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas, La Habana (Cuba); Lira, Carlos Alberto Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  5. Advanced Small-Safe Long-Life Lead Cooled Reactor Cores for Future Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jin Hyeong; Hong, Ser Gi [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    One of the reasons for use of the lead or lead-bismuth alloy coolants is the high boiling temperature that avoids the possibility of coolant voiding. Also, these coolants are compatible with air, steam, and water. Therefore, intermediate coolant loop is not required as in the sodium cooled reactors 3. Lead is considered to be more attractive coolant than lead-bismuth alloy because of its higher availability, lower price, and much lower amount of polonium activity by factor of 104 relatively to lead. On the other hand, lead has higher melting temperature of 601K than that of lead-bismuth (398K), which narrows the operating temperature range and also leads to the possibility of freezing and blockage in fresh cores. Neutronically, the lead and lead-bismuth have very similar characteristics to each other. The lead-alloy coolants have lower moderating power and higher scattering without increasing moderation for neutrons below 0.5MeV, which reduces the leakage of the neutrons through the core and provides an excellent reflecting capability for neutrons. Due to the above features of lead or lead-alloy coolants, there have been lots of studies on the small lead cooled core designs. In this paper, small-safe long-life lead cooled reactor cores having high discharge burnup are designed and neutronically analyzed.. The cores considered in this work rates 110MWt (36.7MWe). In this work, the long-life with high discharge burnup was achieved by using thorium or depleted uranium blanket loaded in the central region of the core. Also, we considered a reference core having no blanket for the comparison. This paper provides the detailed neutronic analyses for these small long-life cores and the detailed analyses of the reactivity coefficients and the composition changes in blankets. The results of the core design and analyses show that our small long-life cores can be operated without refueling over their long-lives longer than 45EFPYs (Effective Full Power Year). In this work

  6. JSPS-CAS Core University Program seminar on summary of 10-year collaborations in plasma and nuclear fusion research area

    International Nuclear Information System (INIS)

    Toi, Kazuo; Wang Kongjia

    2011-07-01

    The JSPS-CAS Core University Program (CUP) seminar on “Summary of 10-year Collaborations in Plasma and Nuclear Fusion Research Area” was held from March 9 to March 11, 2011 in the Okinawa Prefectural Art Museum, Naha city, Okinawa, Japan. The collaboration program on plasma and nuclear fusion started from 2001 under the auspices of Japanese Society of Promotion of Science (JSPS) and Chinese Academy of Sciences (CAS). This year is the last year of the CUP. This seminar was organized in the framework of the CUP. In the seminar, 29 oral talks were presented, having 14 Chinese and 30 Japanese participants. These presentations covered key topics related to the collaboration categories: (1) improvement of core plasma properties, (2) basic research on fusion reactor technologies, and (3) theory and numerical simulation. This seminar aims at summarizing the results obtained through the collaborations for 10 years, and discussing future prospects of China-Japan collaboration in plasma and nuclear fusion research areas. (author)

  7. Development of Pipeline Database and CAD Model for Selection of Core Security Zone in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Choi, Seong Soo; Kwon, Tae Gyun; Baek, Hun Hyun; Kwon, Min Jin

    2008-07-01

    The objective of the project is to develop the pipeline database which can be used for selection of core security zones considering safety significance of pipes and to develop CAD model for 3-dimensional visualization of core security zones, for the purpose of minimizing damage and loss, enforcing security and protection on important facilities, and improving plant design preparing against emergency situations such as physical terrors in nuclear power plants. In this study, the pipeline database is developed for selection of core security zones considering safety significance of safety class 1 and 2 pipes. The database includes the information on 'pipe-room information-surrogate component' mapping, initiating events which may occur and accident mitigation functions which may be damaged by the pipe failure, and the drawing information related to 2,270 pipe segments of 30 systems. For the 3-dimensional visualization of core security zones, the CAD models on the containment building and the auxiliary building are developed using 3-D MAX tool and the demo program which can visualize the direct-X model converted from the 3-D MAX model is also developed. In addition to this, the coordinate information of all the buildings and their rooms is generated using AUTO CAD tool in order to be used as an input for 3-dimensional browsing of the VIP program

  8. An option for the Brazilian nuclear project: necessity of fast breeder reactors and core design for an experimental fast reactor

    International Nuclear Information System (INIS)

    Ishiguro, Y.

    1983-01-01

    In order to assure the continued utilization of fission energy, development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed as the best type for future Brazilian nuclear systems. The inherent safety characteristics are superior to current FBRs and an efficient utilization of the abundant thorium is possible. A first step and a basic tool for the development of FBR technologies is the construction and operation of an experimental fast reactor (EFR). A series of core designs for a 90 MW EFR is studied and possible options and the magnitudes of principal parameters are identified. Flexible modifications of the core and sufficiently high fast fluxes for fuel and materials irradiations appear possible. (Author) [pt

  9. TMI-2 [Three Mile Island Nuclear Power Station] fuel canister and core sample handling equipment used in INEL hot cells

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Shurtliff, W.T.; Lynch, R.J.; Croft, K.M.; Whitmill, L.J.; Allen, S.M.

    1987-01-01

    This paper describes the specialized remote handling equipment developed and used at the Idaho National Engineering Laboratory (INEL) to handle samples obtained from the core of the damaged Unit 2 reactor at Three Mile Island Nuclear Power Station (TM-2). Samples of the core were removed, placed in TMI-2 fuel canisters, and transported to the INEL. Those samples will be examined as part of the analysis of the TMI-2 accident. The equipment described herein was designed for removing sample materials from the fuel canisters, assisting with initial examination, and processing samples in preparation for detailed examinations. The more complex equipment used microprocessor remote controls with electric motor drives providing the required force and motion capabilities. The remaining components were unpowered and manipulator assisted

  10. Effect of buoyancy on fuel containment in an open-cycle gas-core nuclear rocket engine.

    Science.gov (United States)

    Putre, H. A.

    1971-01-01

    Analysis aimed at determining the scaling laws for the buoyancy effect on fuel containment in an open-cycle gas-core nuclear rocket engine, so conducted that experimental conditions can be related to engine conditions. The fuel volume fraction in a short coaxial flow cavity is calculated with a programmed numerical solution of the steady Navier-Stokes equations for isothermal, variable density fluid mixing. A dimensionless parameter B, called the Buoyancy number, was found to correlate the fuel volume fraction for large accelerations and various density ratios. This parameter has the value B = 0 for zero acceleration, and B = 350 for typical engine conditions.

  11. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  12. The domestic development of rhodium self-powered detector used in the core of Qinshan third nuclear power plant

    International Nuclear Information System (INIS)

    Xiong Weihua; Zhang Zhenhua; Yu Yijun; Zhang Yun; Wu Jun; Deng Peng

    2009-01-01

    This article introduced Qinshan third nuclear power plant's Vanadium detector's principle of work, the domestically development's earlier period preparation, the craft processing process, the domestically sample's experiment as well as the sample in core demonstration test. Elaborated process of manufacture's quality control request and the essential craft, and the factory manufacture experiment situation, and to the installation and trial run process, the modification factor and the test result has carried on the introduction and the analysis. (authors)

  13. Piecewise linear approximation: application to control rod step counting in a nuclear reactor core and image contours characterization

    International Nuclear Information System (INIS)

    Kaoutar, M.

    1986-09-01

    After a survey of main algorithms for piecewise linear approximation, a new method is suggested. It consists of two stages: a sequential detection stage and an optimization stage, which derives from general dynamic clustering principle. It is applied to control rod step counting in a nuclear reactor core and images contours characterization. Another version of our method is presented. Its originality cames from the variability of the line segments number during iterations. A comparative study is made by comparing the results of the proposed method with of another methods already existing thereby it attests the efficiency and reliability of our method [fr

  14. MONJU experimental data analysis and its feasibility evaluation to build up the standard data base for large FBR nuclear core design

    International Nuclear Information System (INIS)

    Sugino, K.; Iwai, T.

    2006-01-01

    MONJU experimental data analysis was performed by using the detailed calculation scheme for fast reactor cores developed in Japan. Subsequently, feasibility of the MONJU integral data was evaluated by the cross-section adjustment technique for the use of FBR nuclear core design. It is concluded that the MONJU integral data is quite valuable for building up the standard data base for large FBR nuclear core design. In addition, it is found that the application of the updated data base has a possibility to considerably improve the prediction accuracy of neutronic parameters for MONJU. (authors)

  15. Treatment of core components from nuclear power plants with PWR and BWR reactors - 16043

    International Nuclear Information System (INIS)

    Viermann, Joerg; Friske, Andreas; Radzuweit, Joerg

    2009-01-01

    During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK R cask. The MOSAIK R family of casks is subject of a separate presentation at the ICEM 09 conference. Besides the MOSAIK R cask the treatment system developed by GNS comprised underwater shears to cut the components to size as well as different types of equipment to handle the components, the shears and the MOSAIK R casks in the fuel pool. Over a decade of experience it showed that this system although effective needed improvement for BWR plants where many water channels and control rods had to be replaced after a certain operating time. Because of the large numbers of components the time period needed to cut the components in the pool had a too big influence on other operational work like rearranging of fuel assemblies in the pool. The system was therefore further developed and again a suitable cask was the heart of the solution. GNS developed the type MOSAIK R 80 T, a cask that is capable to ship the unsegmented components with a length of approx. 4.5 m from the Power plants to an external treatment centre. This treatment centre consisting of a hot cell installation with a scrap shear, super-compactor and a heavy

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  17. Strategic plan for the development of core technologies for the Korean advanced nuclear power reactor for export

    International Nuclear Information System (INIS)

    Moon, Joo Hyun; Cho, Young Ho

    2010-01-01

    With the soaring oil price and worsening global warming, nuclear power has attracted considerable attention on a global scale and a new large market of nuclear power plants (NPPs) is expected. The Korean government aims to export up to 10 NPPs by 2012, based on the successful export of 2 NPPs to the UAE in 2009. It is also going to develop a follow-up model of the Advanced Power Reactor (APR) 1400, and join the world's NPP market under the banner of Korea's original reactor type. For this, it promulgated the strategic plan, NuTech 2012, a technology development plan intended for the early acquisition of core technologies for the Korean advanced NPP design and domestic production of the main components in NPP. This paper introduces the strategic plan of NuTech 2012. (orig.)

  18. Preliminary Thermo-hydraulic Core Design Analysis of Korea Advanced Nuclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Lee, Jeong Ik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    Nclear rockets improve the propellant efficiency more than twice compared to CRs and thus significantly reduce the propellant requirement. The superior efficiency of nuclear rockets is due to the combination of the huge energy density and a single low molecular weight propellant utilization. Nuclear Thermal Rockets (NTRs) are particularly suitable for manned missions to Mars because it satisfies a relatively high thrust as well as a high propellant efficiency. NTRs use thermal energy released from a nuclear fission reactor to heat a single low molecular weight propellant, i. e., Hydrogen (H{sub 2}) and then exhausted the extremely heated propellant through a thermodynamic nozzle to produce thrust. A propellant efficiency parameter of rocket engines is specific impulse (I{sub sp}) which represents the ratio of the thrust over the rate of propellant consumption. The difference of I{sub sp} makes over three times propellant savings of NTRs for a manned Mars mission compared to CRs. NTRs can also be configured to operate bimodally by converting the surplus nuclear energy to auxiliary electric power required for the operation of a spacecraft. Moreover, the concept and technology of NTRs are very simple, already proven, and safe. Thus, NTRs can be applied to various space missions such as solar system exploration, International Space Station (ISS) transport support, Near Earth Objects (NEOs) interception, etc. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The ROK has also begun the research for space nuclear systems as a volunteer of the international space race and a major world nuclear energy country. KANUTER is one of the advanced NTR engines currently under development at KAIST. This bimodal engine is operated in two modes of propulsion with 100 MW

  19. 47{sup th} Annual meeting on nuclear technology (AMNT 2016). About cores, coal and cash

    Energy Technology Data Exchange (ETDEWEB)

    Podivinsky, Tomas Jan

    2016-07-15

    Rationality and - especially with regard to reducing emissions - technological neutrality are two commitments for nuclear fission. The Czech Republic, where conditions are not suitable for economical large-scale operation of facilities based on renewables, there is no alternative in environmental or business policy to the reasonable use of nuclear energy. The aim of the updated Czech energy strategy is to increase the proportion of nuclear energy from 35 % to approx. 50 % of power generation and to cover the rest - together with ultra-high efficiency coal fired power plants - with energy from renewable sources and gas fired power plants.

  20. Assessment of mass fraction and melting temperature for the application of limestone concrete and siliceous concrete to nuclear reactor basemat considering molten core-concrete interaction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jae; Kim, Do Gyeum [Korea Institute of Civil Engineering and Building Technology, Goyang (Korea, Republic of); Cho, Jae Leon [Korea Hydro and Nuclear Power Co., Ulsan (Korea, Republic of); Yoon, Eui Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Myung Suk [Korea Hydro and Nuclear Power Co., Central Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

  1. Pressurized water reactor in-core nuclear fuel management by tabu search

    International Nuclear Information System (INIS)

    Hill, Natasha J.; Parks, Geoffrey T.

    2015-01-01

    Highlights: • We develop a tabu search implementation for PWR reload core design. • We conduct computational experiments to find optimal parameter values. • We test the performance of the algorithm on two representative PWR geometries. • We compare this performance with that given by established optimization methods. • Our tabu search implementation outperforms these methods in all cases. - Abstract: Optimization of the arrangement of fuel assemblies and burnable poisons when reloading pressurized water reactors has, in the past, been performed with many different algorithms in an attempt to make reactors more economic and fuel efficient. The use of the tabu search algorithm in tackling reload core design problems is investigated further here after limited, but promising, previous investigations. The performance of the tabu search implementation developed was compared with established genetic algorithm and simulated annealing optimization routines. Tabu search outperformed these existing programs for a number of different objective functions on two different representative core geometries

  2. Corrosion of cermet cores of fuel plates for nuclear research reactor

    International Nuclear Information System (INIS)

    Durazzo, M.; Ramanathan, L.V.

    1984-01-01

    Materials Testing Reactor (MTR) type fuel plates containing U 3 O 8 -Al cores and clad with Al are used in various research reactor. Preliminary investigations, where in the cladding of samples was drilled to simulate conditions of rupture due to pitting attack, revealed that considerable quantities of H 2 was evolved upon exposure of the core to water. The corrosion of cermets cores of different densities was characterized as a function of H 2 evolution that revealed 3 stages. A first stage consisting of an incubation period followed by initiation of H 2 evolution, a second stage with a constant rate of H 2 evolution and a third stage with a low rate of H 2 evolution. All 3 stages were found to vary as a function of cermet density and water temperature. (Author) [pt

  3. Validation of the CORD-2 System for the Nuclear Design Calculations of the NPP Krsko Core

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2016-01-01

    The CORD-2 package intended for core design calculations of PWRs has be recently updated with some improved models. Since the modifications could substantially influence the obtained results, a technical validation process is required. This paper presents comparison of some calculated and measured parameters of the NPP Krsko core needed to qualify the package. Critical boron concentrations at hot full power for selected cycle burnup points and several parameters obtained during the start-up testing at the beginning of each cycle (hot zero power critical concentration, isothermal temperature coefficient and rods worth) for all 27 finished cycles of operation are considered. In addition, assembly-wise power distribution for some selected cycles is checked. Comparison has shown very good agreement of the CORD-2 calculated values with the selected measured parameter of the NPP Krsko core.(author).

  4. Lighting system for the lower core plate of a nuclear reactor

    International Nuclear Information System (INIS)

    Feuillet, P.; Bonin, J.P.

    1986-01-01

    The invention proposes a grazing lighting system for the lower core plate, creating an excellent contrast and offering a good estimation of the relief; it can stay at the same place during the whole or at least the greater part of the core refueling operation. This lighting system is proposed for a reactor of which the lower core plate has fuel assembly centering elements. It has a sealed vessel with a transparent side wall containing several lights independently controlled and each one illuminating a sector of its wall. The vessel has a bottom aimed at resting on the lower plate and provided with centering and holding means acting with several of the said centering means through the plate, and/or apertures for coolant through the plate, and an upper container provided with gripping and handling elements and sealed conduits for electrical cables feeding the lights [fr

  5. Evaluation of nuclear characteristics of minor actinide loaded core. Analyses of BFS-69 and BFS-66-2 critical experiments

    International Nuclear Information System (INIS)

    Hazama, Taira; Sato, Wakaei

    2010-09-01

    Collaboration with Russian Institute of Physics and Power Engineering named 'Investigation of neutronic-physical characteristics and their change when introducing large quantity of neptunium (Np) at different BFS critical assemblies' has been accomplished. This is the second report of the collaboration to describe experimental information and analysis results on BFS-69 and BFS-66-2 critical experiments. In the experiments, various nuclear characteristics were measured in 2 kinds of cores with/without Np loading of about 8 kg. JAEA's standard analysis results were presented with four kinds of nuclear data (JENDL-3.2, JENDL-3.3, JENDL/AC-2008, and ENDF/BVII). Analytical results show: 1) An overestimation trend has been observed in BFS-69 criticality results, especially with JENDL-3.3 and JENDL/AC-2008. The difference from ENDF/B-II having better results mainly lies in the average cosine of the scattering angle around 1 MeV. 2) A small discrepancy exists in BFS-69 Na void reactivity results with the three JENDL nuclear data. The difference from ENDF/B-II mainly lies in scattering cross sections of sodium around 1 MeV and fission cross section of 239 Pu around 1 keV. 3) The analysis results simulate measured Np effects on nuclear characteristics within experimental errors. (author)

  6. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the

  7. 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

  8. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  9. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    International Nuclear Information System (INIS)

    Mesquita, Amir Z.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L.

    2011-01-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  10. Centering and fixing device of a control rod guide flange on a nuclear reactor core plate

    International Nuclear Information System (INIS)

    Chevereau, G.

    1990-01-01

    The device comprises at least two pins diametrally opposite entering into holes in the core plate and self locking pads sliding into bores in the flange. These pads have a command mechanism blocking the lateral displacement of the pads on the plate [fr

  11. observer-based diagnostics and monitoring of vibrations in nuclear reactor core cooling system

    International Nuclear Information System (INIS)

    Siry, S.A K.

    2007-01-01

    analysis and diagnostics of vibration in industrial systems play a significant rule to prevent severe severe damages . drive shaft vibration is a complicated phenomenon composed of two independent forms of vibrations, translational and torsional. translational vibration measurements in case of the reactor core cooling system are introduced. the system under study consists of the three phase induction motor, flywheel, centrifugal pump, and two coupling between motor-flywheel, and flywheel-pump. this system structure is considered to be one where the blades are pegged into the discs fitting into the shafts. a non-linear model to simulate vibration in the reactor core cooling system will be introduced. simulation results of an operating reactor core cooling system using the actual parameters will be presented to validate the accuracy and reliability of the proposed analytical method the accuracy in analyzing the results depends on the system model. the shortcomings of the conventional model will be avoided through the use of that accurate nonlinear model which improve the simulation of the reactor core cooling system

  12. Development of a standard data base for FBR core nuclear design. 9. Analysis of FCA XVII-1 experiments

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Ishikawa, Makoto; Oigawa, Hiroyuki; Iijima, Susumu

    1998-10-01

    Pnc had developed the adjusted nuclear cross-section library in which the results of the Jupiter experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of Fbr cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for FCA XVII-1 assembly. FCA XVII-1 is a representative mock-up of a MOX fuel sodium cooling FBR core. The criticality, reaction rate ratio, sodium void reactivity worth and 238 U Doppler reactivity worth of FCA XVII-1 were analyzed. The results of C/E values calculated by the standard analytical method for JUPITER experiments are similar to those calculated by the method of JAERI, except for the sodium void reactivity. So, further investigation for sodium void reactivity is necessary. Furthermore, sensitivity analysis shows the characteristics of FCA XVII-1 in comparison with ZPPR-9. (author)

  13. Modelling of a rod bundle under viscous and uncompressible flow by porous media. Applied to nuclear reactor core

    International Nuclear Information System (INIS)

    Ricciardi, Guillaume; Collard, Bruno; Bellizzi, Sergio; Cochelin, Bruno

    2007-01-01

    This study is about the safety of nuclear reactor core submitted to seismic loading. In order to reduce the incertitude margin of the present day codes we propose to develop a numerical code including the non linear behavior of the fluid/structure coupling. The challenge of this work is to find out a tractable model taking the structure complexity into account. In this paper we model the nuclear reactor core mechanical behavior including the dynamics of both fuel assemblies of fluid. Each rod bundle is considered as a deformable porous media, so the velocity field of the fluid and the displacement field of the structure are defined in the whole domain space. Fluid part and structure part are in a first time considered separately, and in second time, the two parts are coupled. The motion equations of the structure are obtained by a Lagrangian formulation, and to allow the fluid structure coupling, the motion equations of the fluid are obtained by an Arbitrary Lagrangian Eulerian formulation. The finite elements method is applied to spatially discretize the equations. Simulations have been performed to analyze the influence of the fluid and structure characteristics, phenomena observed by the experience have been reproduced qualitatively. (author)

  14. Using SAFRAN Software to Assess Radiological Hazards from Dismantling of Tammuz-2 Reactor Core at Al-tuwaitha Nuclear Site

    Science.gov (United States)

    Abed Gatea, Mezher; Ahmed, Anwar A.; jundee kadhum, Saad; Ali, Hasan Mohammed; Hussein Muheisn, Abbas

    2018-05-01

    The Safety Assessment Framework (SAFRAN) software has implemented here for radiological safety analysis; to verify that the dose acceptance criteria and safety goals are met with a high degree of confidence for dismantling of Tammuz-2 reactor core at Al-tuwaitha nuclear site. The activities characterizing, dismantling and packaging were practiced to manage the generated radioactive waste. Dose to the worker was considered an endpoint-scenario while dose to the public has neglected due to that Tammuz-2 facility is located in a restricted zone and 30m berm surrounded Al-tuwaitha site. Safety assessment for dismantling worker endpoint-scenario based on maximum external dose at component position level in the reactor pool and internal dose via airborne activity while, for characterizing and packaging worker endpoints scenarios have been done via external dose only because no evidence for airborne radioactivity hazards outside the reactor pool. The in-situ measurements approved that reactor core components are radiologically activated by Co-60 radioisotope. SAFRAN results showed that the maximum received dose for workers are (1.85, 0.64 and 1.3mSv/y) for activities dismantling, characterizing and packaging of reactor core components respectively. Hence, the radiological hazards remain below the low level hazard and within the acceptable annual dose for workers in radiation field

  15. Systems analysis of nuclear solid-core engines for cis-lunar trajectories

    International Nuclear Information System (INIS)

    Ulrich, T.

    1984-01-01

    This report summarizes the result of a comprehensive study about the use of nuclear engines in cis-lunar space. The nuclear space transportation system elements were defined and the restrictions imposed on the nuclear ferries by the chemical Earth-to-LEO transportation system were analyzed. Operating conditions are met best by tungsten-water-moderated reactors due to a high specific impulse and long durability. Specific transportation cost for LEO-to-GEO and LEO-to-lunar orbit flights were calculated for a transportation system life of 50 years. Average transportation cost were estimated to be about 141 $/kg. No difference was made for both routes mentioned above. An additional analysis of smaller and larger flight units showed only small cost reductions by employing larger ferries but a significant cost increase in case smaller flight units would be used. (orig.) [de

  16. The gamma thermometer as a measuring instrument of the nuclear power of a LWR core

    International Nuclear Information System (INIS)

    Hantouche, C.

    1984-07-01

    After a presentation of the gamma thermometer and its environment, this thesis deals with the calibration of a gamma thermometer or determination of the thermal transfer ratio of the signal. Then, the acquisition systems for signals obtained from gamma thermometers installed in PWR nuclear power plants are presented. One deals also with the nuclear transfer function establishment and deconvolution. Then, one deals with the reconstruction of the axial power of an assembly provided with a gamma thermometer. Qualitative and quantitative studies of measurements obtained from this gamma thermometer are finally presented. 64 refs [fr

  17. A new paradigm for core design aimed at the sustainability of nuclear energy: The solution of the extended equilibrium state

    International Nuclear Information System (INIS)

    Artioli, Carlo; Grasso, Giacomo; Petrovich, Carlo

    2010-01-01

    The future expansion of nuclear energy, a technology identified as one of the main candidates for reducing the world's dependence on fossil fuels, requires a thorough analysis of the sustainability of this energy source for long-term supply. Generation-IV nuclear systems could represent a turning point for energy production by minimizing the environmental footprint of the fuel cycle. A new paradigm is thus required for reactor design, focusing, at the core design level, on both the closure of the fuel cycle and the effective utilization of natural resources. Within this framework, the so-called 'adiabatic core' concept represents a particularly interesting solution. It is based on the idea of ensuring by design a condition of equilibrium in the fuel cycle (i.e., an equilibrium 'fuel vector'), foreseeing nuclear power systems able to maintain a constant total amount of both plutonium and minor actinides (TRU), consuming only uranium (either natural or depleted), while discharging to the environment only fission products and reprocessing losses. Under such a hypothesis, all actinides can be continuously recycled in the same system, reducing both the waste volume and its long-term radiotoxicity, as well as utilizing effectively uranium resources. Two mathematical approaches have been devised to find the 'extended' equilibrium solution for the fuel vector. These methods are compared, validated with the codes MCNPX and FISPACT and applied to the European lead-cooled fast reactor ELSY, confirming the potential of this approach (e.g., a reduction by two orders of magnitude of the TRU mass in the final waste in comparison with the fuel cycle of Light Water Reactors operated in a once-through scenario).

  18. Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system

    International Nuclear Information System (INIS)

    Arigane, Kenji

    1987-04-01

    The reduced enrichment program for the JRR-4 has been progressing based on JAERI's RERTR (Reduced Enrichment Research and Test Reactor) program. The SRAC (JAERI Thermal Reactor Standard Code System for Reactor Design and Analysis) is used for the neutronic design of the JRR-4 LEU Core. This report describes the benchmark calculations on the neutronic characteristics of the JRR-4 HEU Core in order to validate the calculation method. The benchmark calculations were performed on the various kind of neutronic characteristics such as excess reactivity, criticality, control rod worth, thermal neutron flux distribution, void coefficient, temperature coefficient, mass coefficient, kinetic parameters and poisoning effect by Xe-135 build up. As the result, it was confirmed that these calculated values are in satisfactory agreement with the measured values. Therefore, the calculational method by the SRAC was validated. (author)

  19. Nuclear structure calculations in $^{20}$Ne with No-Core Configuration-Interaction model

    OpenAIRE

    Konieczka, Maciej; Satuła, Wojciech

    2016-01-01

    Negative parity states in $^{20}$Ne and Gamow-Teller strength distribution for the ground-state beta-decay of $^{20}$Na are calculated for the very first time using recently developed No-Core Configuration-Interaction model. The approach is based on multi-reference density functional theory involving isospin and angular-momentum projections. Advantages and shortcomings of the method are briefly discussed.

  20. Emotional learning based intelligent controller for a PWR nuclear reactor core during load following operation

    International Nuclear Information System (INIS)

    Khorramabadi, Sima Seidi; Boroushaki, Mehrdad; Lucas, Caro

    2008-01-01

    The design and evaluation of a novel approach to reactor core power control based on emotional learning is described. The controller includes a neuro-fuzzy system with power error and its derivative as inputs. A fuzzy critic evaluates the present situation, and provides the emotional signal (stress). The controller modifies its characteristics so that the critic's stress is reduced. Simulation results show that the controller has good convergence and performance robustness characteristics over a wide range of operational parameters

  1. Modernization of the online core monitoring in the nuclear power plant Gundremmingen

    International Nuclear Information System (INIS)

    Holzer, Robert; Grosshans, Ingo

    2008-01-01

    In NPP Gundremmingen unit B and C the online core monitoring system KSIM is used since 1994. IN 2006 the modernization of the system was started with focus on the actualization of the simulator models according to the state of the art and improved user interface. The new system is operating in parallel operation with the old system and will overtake completely by 2008. The contribution covers a description of the system, the hardware and the system soft ware.

  2. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant; Analisis de documentos de los materiales de la envolvente del nucleo del reactor nuclear de la CLV U-1

    Energy Technology Data Exchange (ETDEWEB)

    Zamora R, L.; Medina F, A. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1999-07-01

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  4. Analysis of hypothetical nuclear excursions in the external core retention system

    International Nuclear Information System (INIS)

    Froehlich, R.; Kussmaul, G.; Schmuck, P.

    1976-01-01

    The core catcher system of the SNR 300 is outside the reactor tank. The probability of recriticality phenomena is reduced by its design, but the licensing procedures still call for the analysis of strong recriticality phenomena in the core catcher system outside the reactor tank in order to achieve a better understanding of the possible physical effects and to get to know the safety limits of the system. For their theoretical investigations, the authors used a two-partner model as presented in fig. 1. At the bottom of the core catcher - which consists of depleted UO 2 - there is a fuel cylinder. Another fuel cylinder (with the same axis) is dropped from a height of 250 cm. The two cylindrical masses are immersed in sodium, but a free fall is assumed since the possibility cannot be excluded that the reactor bottom may be empty or only partially filled with sodium. It was found that under these conditions the strongest excursions may be expected in those cases where prompt criticality does not occur until just before the two partners meet. (orig./AK) [de

  5. Nuclear reactor support and seismic restraint with in-vessel core retention cooling features

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, Tyler A.; Edwards, Michael J.

    2018-01-23

    A nuclear reactor including a lateral seismic restraint with a vertically oriented pin attached to the lower vessel head and a mating pin socket attached to the floor. Thermally insulating materials are disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head.

  6. Neutronics and Thermal Hydraulics Analysis of a Conceptual Ultra-High Temperature MHD Cermet Fuel Core for Nuclear Electric Propulsion

    Directory of Open Access Journals (Sweden)

    Jian Song

    2018-04-01

    Full Text Available Nuclear electric propulsion (NEP offers unique advantages for the interplanetary exploration. The extremely high conversion efficiency of magnetohydrodynamics (MHD conversion nuclear reactor makes it a highly potential space power source in the future, especially for NEP systems. Research on ultra-high temperature reactor suitable for MHD power conversion is performed in this paper. Cermet is chosen as the reactor fuel after a detailed comparison with the (U,ZrC graphite-based fuel and mixed carbide fuel. A reactor design is carried out as well as the analysis of the reactor physics and thermal-hydraulics. The specific design involves fuel element, reactor core, and radiation shield. Two coolant channel configurations of fuel elements are considered and both of them can meet the demands. The 91 channel configuration is chosen due to its greater heat transfer performance. Besides, preliminary calculation of nuclear criticality safety during launch crash accident is also presented. The calculation results show that the current design can meet the safety requirements well.

  7. Difference of reactor core nuclear instrument between AP1000 and CPR1000

    International Nuclear Information System (INIS)

    Zhang Shidong; Zhou Can; Deng Tian

    2014-01-01

    As a typical generation Ⅲ reactor technique, the AP1000 applies many advanced design concepts, simplifies the design, reduces equipment quantities, and thus enhances systematic reliability. The comparison of reactor core measurement instrument differences between AP1000 and CPR1000 from several aspects was involved in the paper. Through analysis and comparison of these differences, passive design concepts and characteristics of AP1000 are familiarized, and conveniences for staffs engaged in CPR1000 to learn and grasp AP1000 technique are provided. It is useful in reactor start up, operation and maintenance. (authors)

  8. Inventories of radioactive fission products in the core of thermal nuclear reactor

    International Nuclear Information System (INIS)

    Marinkovic, N.

    1977-01-01

    As a part of the analysis concerning radiological consequences of a major LWR accident, inventories of the most significant radioactive nuclides and stable fission gases in the core of a PWR type reactor have been calculated. Calculations were performed by the DELFIN code using nuclide data and neutron flux data earlier obtained by the METHUSELAH code. Comparison with simplified calculation method show that it is quite rough for certain nuclides but the accuracy may be sufficient for safety analysis purposes recalling the inaccuracies in the later parts of fission product transport process (author)

  9. Application of genetic algorithms to in-core nuclear fuel management optimization

    International Nuclear Information System (INIS)

    Poon, P.W.; Parks, G.T.

    1993-01-01

    The search for an optimal arrangement of fresh and burnt fuel and control material within the core of a PWR represents a formidable optimization problem. The approach of combining the robust optimization capabilities of the Simulated Annealing (SA) algorithm with the computational speed of a Generalized Perturbation Theory (GPT) based evaluation methodology in the code FORMOSA has proved to be very effective. In this paper, we show that the incorporation of another stochastic search technique, a Genetic Algorithm, results in comparable optimization performance on serial computers and offers substantially superior performance on parallel machines. (orig.)

  10. An evaluation of nuclear design characteristics of duplex burnable poison rods for extended cycle core

    International Nuclear Information System (INIS)

    Lee, D. J.; Kim, M. H.; Song, K. W.

    2003-01-01

    Nuclear design characteristics of duplex burnable poison rod were evaluated for three integral type burnable absorbers; Gadolinia, Erbia and IFBA. Inter-comparison was done for both 12 and 24 month cycle for Korean Standard Nuclear Plant. Fuel assemblies with duplex BP was designed to the equivalent assembly with 8 and 16 gadolinia BP 2 . Duplex BP is composed of inner region of natural U-Gd 2 O 3 , and outer shell of, UO 2 -Er2O 3 . In order to evaluate the duplex BP, assemblies with erbia and IFBA were compared with alternative options. A sensitivity studies were performed to the size of region, compositions and location of duplex BPs. It was shown that duplex BP gave favorable k-infinite curve to burnup, but IFBA provided the least residual reactivity penalty as EOC. Erbia was good for more negative MTCs. IFBA and erbia had better neutronic performance than gadolinia od duplex BP in the aspect of pin power peaking

  11. Study and application of ANISN and DOT-II nuclear cores in reactor physics problems

    International Nuclear Information System (INIS)

    Dias, Artur Flavio

    1980-01-01

    To solve time-independent neutrons and/or gamma rays transport problems in nuclear reactors, two codes available at IPEN were studied and applied to solve benchmark problems. The ANISN code solves the one-dimensional Boltzmann transport equation for neutrons or gamma rays, in plane, spherical, or cylindrical geometries. The DOT-II code solves the same equation in two-dimensional space for plane, cylindrical and circular geometries. General anisotropic scattering allowed in both codes. Moreover, pointwise convergence criteria, and alternate step function difference equations are also used in order to remove the oscillating flux distributions, sometimes found in discrete ordinates solutions. Basic theories and numerical techniques used in these codes are studied and summarized. Benchmark problems have been solved using these codes. Comparisons of the results show that both codes can be used with confidence in the analysis of nuclear problems. (author)

  12. The efficiency and fidelity of the in-core nuclear fuel management code FORMOSA-P

    International Nuclear Information System (INIS)

    Kropaczek, D.J.; Turinsky, P.J.

    1994-01-01

    The second-order generalized perturbation theory (GPT), nodal neutronic model utilized within the nuclear fuel management optimization code FORMOSA-P is presented within the context of prediction fidelity and computational efficiency versus forward solution. Key features of thr GPT neutronics model as implemented within the Simulated Annealing optimization adaptive control algorithm are discussed. Supporting results are then presented demonstrating the superior consistency of adaptive control for both global and local optimization searches. (authors). 15 refs., 1 fig., 4 tabs

  13. Nuclear reactor, reactor core thereof, and device for constituting the reactor

    International Nuclear Information System (INIS)

    Takiyama, Masashi.

    1994-01-01

    A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

  14. Blowdown hydraulic influence on core thermal response in LOFT nuclear experiment L2-3

    International Nuclear Information System (INIS)

    Reeder, D.L.

    1979-01-01

    Experimental research into pressurized water reactor (PWR) loss-of-coolant phenomena conducted in the Loss-of-Fluid Test (LOFT) facility has given results indicating that for very large pipe breaks the core thermal response is tightly coupled to the fluid hydraulic phenomena during the blowdown phase of the loss-of-coolant transient. This summary presents and discusses data supporting this conclusion. LOFT Loss-of-Coolant Experiment (LOCE) L2-3 simulated a complete double-ended offset shear break of a primary coolant reactor vessel inlet pipe in a commercial PWR. The LOFT system conditions at experiment initiation were: fuel rod maximum linear heat generation rate (MLHGR) of 39.4 +- 3 kW/m, hot leg temperature of 593 +- 3 K, core ΔT of 32.2 +- 4 K, system pressure of 15.06 +- 0.03 MPa, and flow rate/system volume of 25.6 +- 0.8 kg/m 3 . These conditions are typical of those in commercial PWR systems at normal operating conditions

  15. Development of intelligent code system to support conceptual design of nuclear reactor core

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki; Tsuchihashi, Keichiro

    1997-01-01

    An intelligent reactor design system IRDS has been developed to support conceptual design of new type reactor cores in the fields of neutronics, thermal-hydraulics and fuel behavior. The features of IRDS are summarized as follows: 1) a variety of computer codes to cover various design tasks relevant to 'static' and 'burnup' problems are implemented, 2) all the information necessary to the codes implemented is unified in a data base, 3) several data and knowledge bases are referred to in order to proceed design process efficiently for non-expert users, 4) advanced man-machine interface to communicate with the system through an interactive and graphical user interface is equipped and 5) a function to search automatically a design window, which is defined as a feasible parameter range to satisfy design requirement and criteria is employed to support the optimization or satisfication process. Applicability and productivity of the system are demonstrated by the design study of fuel pin for new type FBR cores. (author)

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  17. Out-of-core nuclear fuel cycle economic optimization for nonequilibrium cycles

    International Nuclear Information System (INIS)

    Comes, S.A.

    1987-01-01

    A methodology and associated computer code was developed to determine near-optimum out-of-core fuel management strategies. The code, named OCEON (Out-of-Core Economic OptimizationN), identified feed-region sizes and enrichments, and partially burned fuel-reload strategies for each cycle of a multi-cycle planning horizon, subject to cycle-energy requirements and constraints on feed enrichments, discharge burnups, and the moderator temperature coefficient. A zero-dimensional reactor physics model, enhanced by a linear reactivity model to provide batch power shares, performs the initial feed enrichment, burnup and constraint evaluations, while a two-dimensional, nodal code is used to refine the calculations for the final solutions. The economic calculations are performed rapidly using an annuity-factor-based model. Use of Monte Carlo integer programming to select the optimum solutions allows for the determination of a family of near-optimum solutions, from which engineering judgment may be used to select an appropriate strategy. Results from various nonequilibrium cycle energy requirement cases typically show a large number of low-cost solutions near the optimum. This confirms that the Monte Carlo integer programming approach of generating a family of solutions will be most useful for selecting optimum strategies when other considerations, such as incore loading pattern concerns, must be addressed

  18. The development of the Nuclear Electric core performance and fault transient analysis code package in support of Sizewell B

    International Nuclear Information System (INIS)

    Hall, P.; Hutt, P.

    1994-01-01

    This paper describes Nuclear Electric's (NE) development of an integrated code package in support of all its reactors including Sizewell B, designed for the provision of fuel management design, core performance studies, operational support and fault transient analysis. The package uses the NE general purpose three-dimensional transient reactor physics code PANTHER with cross-sections derived in the PWR case from the LWRWIMS LWR lattice neutronics code. The package also includes ENIGMA a generic fuel performance code and for PWR application VIPRE-01 a subchannel thermal hydraulics code, RELAP5 the system thermal hydraulics transient code and SCORPIO an on-line surveillance system. The paper describes the capabilities and validation of the elements of this package for PWR, how they are coupled within the package and the way in which they are being applied for Sizewell B to on-line surveillance and fault transient analysis. (Author)

  19. The self-consistent energy system with an enhanced non-proliferated core concept for global nuclear energy utilization

    International Nuclear Information System (INIS)

    Kawashima, Masatoshi; Arie, Kazuo; Araki, Yoshio; Sato, Mitsuyoshi; Mori, Kenji; Nakayama, Yoshiyuki; Nakazono, Ryuichi; Kuroda, Yuji; Ishiguma, Kazuo; Fujii-e, Yoichi

    2008-01-01

    A sustainable nuclear energy system was developed based on the concept of Self-Consistent Nuclear Energy System (SCNES). Our study that trans-uranium (TRU) metallic fuel fast reactor cycle coupled with recycling of five long-lived fission products (LLFP) as well as actinides is the most promising system for the sustainable nuclear utilization. Efficient utilization of uranium-238 through the SCNES concept opens the doors to prolong the lifetime of nuclear energy systems towards several tens of thousand years. Recent evolution of the concept revealed compatibility of fuel sustainability, minor actinide (MA) minimization and non-proliferation aspects for peaceful use of nuclear energy systems through the discussion. As for those TRU compositions stabilized under fast neutron spectra, plutonium isotope fractions are remained in the range of reactor grade classification with high fraction of Pu240 isotope. Recent evolution of the SCNES concept has revealed that TRU recycling can cope with enhancing non-proliferation efforts in peaceful use with the 'no-blanket and multi-zoning core' concept. Therefore, the realization of SCNES is most important. In addition, along the process to the goals, a three-step approach is proposed to solve concurrent problems raised in the LWR systems. We discussed possible roles and contribution to the near future demand along worldwide expansion of LWR capacities by applying the 1st generation SCNES. MA fractions in TRU are more than 10% from LWR discharged fuels and even higher up to 20% in fuels from long interim storages. TRU recycling in the 1st generation SCNES system can reduce the MA fractions down to 4-5% in a few decades. This capability significantly releases 'MA' pressures in down-stream of LWR systems. Current efforts for enhancing capabilities for energy generation by LWR systems are efficient against the global warming crisis. In parallel to those movements, early realization of the SCNES concept can be the most viable decision

  20. Needs and accuracy requirements for fission product nuclear data in the physics design of power reactor cores

    International Nuclear Information System (INIS)

    Rowlands, J.L.

    1978-01-01

    The fission product nuclear data accuracy requirements for fast and thermal reactor core performance predictions were reviewed by Tyror at the Bologna FPND Meeting. The status of the data was assessed at the Meeting and it was concluded that the requirements of thermal reactors were largely met, and the yield data requirements of fast reactors, but not the cross section requirements, were met. However, the World Request List for Nuclear Data (WRENDA) contains a number of requests for fission product capture cross sections in the energy range of interest for thermal reactors. Recent reports indicate that the fast reactor reactivity requirements might have been met by integral measurements made in zero power critical assemblies. However, there are requests for the differential cross sections of the individual isotopes to be determined in addition to the integral data requirements. The fast reactor requirements are reviewed, taking into account some more recent studies of the effects of fission products. The sodium void reactivity effect depends on the fission product cross sections in a different way to the fission product reactivity effect in a normal core. This requirement might call for different types of measurement. There is currently an interest in high burnup fuel cycles and alternative fuel cycles. These might require more accurate fission product data, data for individual isotopes and data for capture products. Recent calculations of the time dependence of fission product reactivity effects show that this is dependent upon the data set used and there are significant uncertainties. Some recent thermal reactor studies on approximations in the treatment of decay chains and the importance of xenon and samarium poisoning are also summarized. (author)

  1. EdF in the core of UK's nuclear industry... before expecting more

    International Nuclear Information System (INIS)

    Moal, C.

    2008-01-01

    With the announcement at the end of September 2008 of EdF's friendly takeover bid on British Energy, the French group confirms its will of dominating the European nuclear industry before going back to the assault of the US market. Together, EdF and British Energy (owner of 8 NPPs (9.5 GW) and 1 coal-fired power plant (2 GW)) will make a turnover of 11.9 billion euro with 19800 employees and 85.6 TWh of production. Short paper. (J.S.)

  2. 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.

  3. Nuclear reactor core support incorporating also a cooling fluid flow system

    International Nuclear Information System (INIS)

    Pennell, W.E.

    1975-01-01

    A description is given of a core bearing plate with several modular intake units having cooling fluid intake openings on their lower extensions, and on their upper ends located above the bearing plate, at least one fuel assembly which is thus in communication with the area under the bearing plate through the modular intake unit. The means for introducing the cooling fluid into the reactor vessel area are located under the bearing plate. The lower ends of the modular intake have ribs arranged essentially on a plane and join together with openings provided between the seals, in such a manner that the ribs form a barrier. The cooling fluid intake openings are located above this barrier, so that the cooling fluid is compelled to cross it before penetrating into the modular intake units [fr

  4. Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems

    International Nuclear Information System (INIS)

    Chow, S.

    1976-01-01

    A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. In the power generation mode, the plasma and propellant flows are shut off, and the driver elements supply thermal power to the power conversion system, which generates electricity for primary electric propulsion purposes

  5. Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

    International Nuclear Information System (INIS)

    Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F.

    2009-01-01

    The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U 235 (typically Pu 242 , Np 237 , U 238 , Th 232 ). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

  6. Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F. [CEA, DEN, Dosimetry Command Control and Instrumentation Laboratory, F-13109 Saint-Paul-lez-Durance (France)

    2009-07-01

    The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U{sup 235} (typically Pu{sup 242}, Np{sup 237}, U{sup 238}, Th{sup 232}). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

  7. The radial distribution of the neutron field in the core of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Binh, Nguyen Duc; Tuan, Nguyen Minh; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Determination of the radial distribution of the thermal neutron field in the core of the Dalat reactor is done by the Cu foil activation method. The measured data are fitted by the least square method to determine several physical parameters of the reactor, as follows: 1. Buckling B{sub r}{sup 2}=(84.6{+-}5.5)10{sup -4}/cm{sup 2}. 2. The effective radius R{sub eff}=(27.6{+-}1.0)cm. 3. The extrapolation distance {lambda}=(8.7{+-}1.0)cm. 4. The unequal coefficient of the effective multiplication K{sub r}=1.77{+-}0.11. (author). 2 refs., 4 figs., 1 tab.

  8. Study on a nuclear spaceship for interplanetary cruise. Core design of a small fast reactor

    International Nuclear Information System (INIS)

    Kitamura, Taku; Yoshida, Yutaka; Honma, Yuji; Narabayashi, Tadashi; Shimazu, Yoichiro; Tsuji, Masashi

    2009-01-01

    In 21st century, the field which needs nuclear power plant systems are not just on the Earth. We considered that the nuclear power is proper for the energy source of the manned spaceship for interplanetary cruise. In this study, we considered the system configuration of the spaceship, the design of power generating system, some navigational plans to reach the Mars. The system configuration of the spaceship studied in our laboratory has one or two Fast Reactor with liquid sodium coolant as main heat source, dozens of Stirling Engines as main power generators and some Plasma Rockets called VASIMR as propulsion system. Because the Fast Reactor need not thick and heavy pressure vessel and the sodium has high performance of heat transfer, they are the best suited to the space nuclear reactor system. In addition, Stirling Engine has high theoretical thermal efficiency and need not water, steam generators, steam condenser and so on. This results in absence of sodium-water reaction and significant weight saving of power generator system. The VASIMR studied at ASPL is an advanced electric propulsion device which is able to convert large amount of electric power into great propulsion force. At reactor designing, we are using the SRAC2006 code developed at JAEA and pursuing the optimal fast reactor design for spaceship. We think that smaller reactor is better. To realize a system which has inherent safety, sodium void reactivity should be negative. We adopted the design of the small fast reactor named 4S (Super Safe, Small and Simple) as a reference design. As a result, we verified that a void reactivity had negative value in some of calculation cases and we realized safe, small and simple space fast reactor. In addition, to piece out power generator system in space, we need to consider if the budget of exhaust heat from radiator panels to space needed at this case is realistic. To obtain the optimal trajectory of rapid Mars transit, we made some analysis calculation codes

  9. The use of reliability analysis techniques applied to nuclear power station emergency core cooling systems

    International Nuclear Information System (INIS)

    Danielsen, A.; Snaith, E.R.

    1975-01-01

    A reliability investigation carried out by the Safety and Reliability Services of the UKAEA, and the SSEB, of the essential system/reactor coolant system for a large nuclear power station is described. In AGR type reactors, after all reactor shutdown conditions, it is necessary to restore forced gas circulation and sufficient boiler feed to maintain the heat removal capacity of the boilers. The coolant requirements are provided by several independent mechanical systems of primary coolant fans, feedwater pumps, and valves integrated with electrical power sources, switchgear, and automatic control equipment. Reliability is treated as one aspect of system performance and quantified in terms of failure to meet a specific objective. Based on the reliability performance of the constituent components the optimum system configuration is determined together with the preferred plant operating procedures and maintenance requirements. (author)

  10. Optimization of reload core design for PWR and application to Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Shen Wei; Zhongsheng Xie; Banghua Yin

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the Linear Programming method using enrichments as control variable. In the second stage the optimum BP allocation is determined by the Flexible Tolerance Method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant(QNP)cycle 2 reloading design

  11. Device for refueling a nuclear reactor having a core comprising a plurality of fuel assemblies

    International Nuclear Information System (INIS)

    Van Santen, A.; Elofsson, K.

    1975-01-01

    A nuclear reactor formed of fuel assemblies each including a plurality of parallel fuel rods arranged in a predetermined fuel rod lattice, which rods are freely extractable and insertable at one end of the fuel assembly, is refueled by extracting from one of the fuel assemblies a number of fuel rods substantially less than the total number of fuel rods and replacing these by inserting new fuel rods into the vacated positions. The removal and return of the rods is produced by a tool having a plurality of gripping members capable of engaging shoulders beneath heads formed on the upper ends of the fuel rods. This may be accomplished by providing a tool having a number of gripping members attached to the tool body corresponding to the lattice positions of the fuel rods to be extracted, having gripping members which can be pushed together to grip beneath shoulders on the upper ends of the fuel rods. (Official Gazette)

  12. Surveillance of a nuclear reactor core by use of a pattern recognition method

    International Nuclear Information System (INIS)

    Invernizzi, Michel.

    1982-07-01

    A pattern recognition system is described for the surveillance of a PWR reactor. This report contains four chapters. The first one succinctly deals with statistical pattern recognition principles. In the second chapter we show how a surveillance problem may be treated by pattern recognition and we present methods for surveillances (detection of abnormalities), controls (kind of running recognition) and diagnotics (kind of abnormality recognition). The third chapter shows a surveillance method of a nuclear plant. The signals used are the neutron noise observations made by the ionization chambers inserted in the reactor. Abnormality is defined in opposition with the training set witch is supposed to be an exhaustive summary of normality. In the fourth chapter we propose a scheme for an adaptative recognition and a method based on classes modelisations by hyper-spheres. This method has been tested on simulated training sets in two-dimensional feature spaces. It gives solutions to problems of non-linear separability [fr

  13. Standard deviation of local tallies in global Monte Carlo calculation of nuclear reactor core

    International Nuclear Information System (INIS)

    Ueki, Taro

    2010-01-01

    Time series methodology has been studied to assess the feasibility of statistical error estimation in the continuous space and energy Monte Carlo calculation of the three-dimensional whole reactor core. The noise propagation was examined and the fluctuation of track length tallies for local fission rate and power has been formally shown to be represented by the autoregressive moving average process of orders p and p-1 [ARMA(p,p-1)], where p is an integer larger than or equal to two. Therefore, ARMA(p,p-1) fitting was applied to the real standard deviation estimation of the power of fuel assemblies at particular heights. Numerical results indicate that straightforward ARMA(3,2) fitting is promising, but a stability issue must be resolved toward the incorporation in the distributed version of production Monte Carlo codes. The same numerical results reveal that the average performance of ARMA(3,2) fitting is equivalent to that of the batch method with a batch size larger than 100 and smaller than 200 cycles for a 1,100 MWe pressurized water reactor. (author)

  14. The loadings and strength of nuclear power plant structures in core damage accidents

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1994-01-01

    The reactor cavity of VVER-91 NPP is a thick-walled, cylindrical reinforced concrete structure. In case of molten core-water reaction during the severe reactor accident the load carrying capacity of the cavity structure is of interest against the short impulse type loading caused by the steam explosion phenomenon. The assumed size of the impulse was 20 kPa-s and the duration was 10 ms. This investigation was divided in several phases. First, the elastic response of the cavity was determined using the ABAQUS code. Next, the static response of the cavity was evaluated using elasto-plastic properties of reinforcement and concrete and also taking into account the cracking of the concrete. This analysis was done with the aid of ABAQUS/STANDARD and ANSYS codes and the obtained results agreed reasonably with each other. In order to obtain a qualitative picture of the behaviour of the structure under the impulse load a simplified single degree of freedorn model was developed. The hoop reinforcement of the cavity was taken as an elasto-plastic spring and the wall concrete acted as a mass. Using this model the suitable amount of hoop reinforcement was determined. In next phase, the dynamic analysis of the structure was attempted using elasto-plastic material properties and concrete cracking. (13 refs., 57 figs.)

  15. Coarse-grained parallel genetic algorithm applied to a nuclear reactor core design optimization problem

    International Nuclear Information System (INIS)

    Pereira, Claudio M.N.A.; Lapa, Celso M.F.

    2003-01-01

    This work extends the research related to generic algorithms (GA) in core design optimization problems, which basic investigations were presented in previous work. Here we explore the use of the Island Genetic Algorithm (IGA), a coarse-grained parallel GA model, comparing its performance to that obtained by the application of a traditional non-parallel GA. The optimization problem consists on adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a 3-enrichment zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. Our IGA implementation runs as a distributed application on a conventional local area network (LAN), avoiding the use of expensive parallel computers or architectures. After exhaustive experiments, taking more than 1500 h in 550 MHz personal computers, we have observed that the IGA provided gains not only in terms of computational time, but also in the optimization outcome. Besides, we have also realized that, for such kind of problem, which fitness evaluation is itself time consuming, the time overhead in the IGA, due to the communication in LANs, is practically imperceptible, leading to the conclusion that the use of expensive parallel computers or architecture can be avoided

  16. Remote viewing system for in core inspection of nuclear reactors (Paper No. 022)

    International Nuclear Information System (INIS)

    Modi, R.K.; Radke, M.G.; Ramaswamy, N.V.; Ramakumar, M.S.; Das, N.C.

    1987-02-01

    A remote viewing optical instrument and illuminator have been designed and fabricated to view and photograph the various components inside the core of FBTR, Kalpakkam. These instruments have resulted in development of indegenious technology and saving of foreign exchange of the order of Rs. 20 lakhs. 'Periscope' consists of several relay optics, a scanning device attached to objective end prism and an eye piece module. Objects located at various positions, can be brought to field of view by a scanning prism. Objects situated at a distance between 1 meter to infinity can be focussed with magnifying power of 2x. 'Projector' is designed to provide illumination level of 200 lux, at the farthest surface which is about 2.8 meters from light source. Provision is made to change the lamp module to obtain more intense light beam. Both instruments are inserted through two identical experimental canals on 3600mm thick small rotating plug of concrete. They are housed in 101.6 mm O, D, and 5000 mm long S.S. tube. The outer surface of the tube is hard chrome plated for better seal and ease of decontamination. During shut down the ambient is filled with argon gas and sodium aerosol at the temperature of about 150 degrees. The instruments have to withstand high level of gamma and other particulate radiations. 'Periscope' is flushed with argon and maintained at 5 psig to avoid any possibility of reactor gases from escaping to atmosphere. 'Projector is continuously flushed with argon gas at 5 psig for cooling the lamp enclosure. The instruments are provided with the biological shield of 700 mm thick lead plugs. (author). 4 figs

  17. Study of the thermo-hydrodynamic phenomena in the nuclear core during reflood phase

    International Nuclear Information System (INIS)

    Murao, Yoshio

    1983-03-01

    This paper describes the development of the core thermo-hydrodynamic model on the reflood phenomena during a loss-of-coolant accident in a light water reactor. This model was developed based on the physical understanding in order to obtain the flexibility of application to safety analysis. For this purpose, the flow pattern was modeled and the fundamental equations were derived. The equations were used to know the suitable variables for assembling the thermo-hydrodynamic model of each flow regime in a reflood analysis code. Then the hydrodynamic models and the heat transfer models of all flow regimes and the quench model were derived. Some of them were newly developed. It was found that water accumulation above the quench front occurred in some cases, however the criteria was not clarified. One-dimensional forced-feed reflood tests were performed and the models were assessed and partly improved by using the data of the tests. The verified models were built in a one-dimensional reflood analysis code and totally assessed with the data of the test mentioned above. Except for the location just below a grid spacer and cases of high flooding rate, the calculational results indicated good comparison with the experimental results when the water accumulation was assumed above the quench front. Additionally the test data from the other test facility were used for the verification of the model. The results also showed good comparison with the experimental results. It was found that better comparisons were obtained when the water accumulation was not assumed above quench front. From these assessment of the model, it was found that the model derived here describes the over-all reflood phenomena, while it has to be partly improved and the water accumulation phenomena should be further investigated. (author)

  18. United States Department of Energy's reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    International Nuclear Information System (INIS)

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage

  19. A feasibility study for the application of enriched gadolinia burnable absorber rods in nuclear core design

    International Nuclear Information System (INIS)

    Lee, Chung Chan; Zee, Sung Quun; Kim, Kang Seog; Song, Jae Seung

    2000-12-01

    An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. It is estimated that the cycle length increases by 2 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 of which the cycle length is assumed 2 years. For the advanced integral reactor SMART in which ultra long cycle length and soluble boron-free operation concept is applied, natural gadolinia burnable absorber rods fail to control the excess reactivity. On the other hand, enriched gadolinia rods are successful in controling the excess reactivity. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions

  20. Comparison of thermohydraulic and nuclear aspects in a standard HEU core and a typical LEU core for the HFR Petten. A case study

    International Nuclear Information System (INIS)

    Pruimboom, H.; Tas, A.

    1985-01-01

    Within the framework of the RERTR program various HEU-LEU core calculations have been performed by ANL in a cooperative effort with ECN and JRC Petten. The main purpose of this work has been to gain competence in analysing HEU-LEU core conversion for high power Materials Testing Reactors and to assist in a possible HEU-LEU conversion of the HFR Petten. For reference purposes the present HFR standard core (HEU) in the 'old' vessel geometry was calculated at first. As a next step the new vessel geometry and the increased fuel weights were taken into account. Subsequently various LEU HFR core options have been analysed. Main parameters in the LEU study were the uranium loading in the meat, the fuel type, the thickness of the meat, the number of fuel plates per element and the type of burnable poison applied. Though the study has not yet been completed, one of its striking preliminary results concerns the increased power peaking in the LEU fuel elements as compared with the HEU situation. A preliminary analysis of the thermal characteristics of a typical LEU core as compared with a standard HEU core has been made and is presented in the paper. A short survey of the various HEU and LEU calculations is given. The thermal safety analysis procedure for the HFR, as based on the flow instability criterion, is clarified. Finally, the thermal comparison HEU versus LEU and the resulting conclusions are presented. (author)

  1. Pygmy and core polarization dipole modes in 206Pb: Connecting nuclear structure to stellar nucleosynthesis

    Science.gov (United States)

    Tonchev, A. P.; Tsoneva, N.; Bhatia, C.; Arnold, C. W.; Goriely, S.; Hammond, S. L.; Kelley, J. H.; Kwan, E.; Lenske, H.; Piekarewicz, J.; Raut, R.; Rusev, G.; Shizuma, T.; Tornow, W.

    2017-10-01

    A high-resolution study of the electromagnetic response of 206Pb below the neutron separation energy is performed using a (γ → ,γ‧) experiment at the HI γ → S facility. Nuclear resonance fluorescence with 100% linearly polarized photon beams is used to measure spins, parities, branching ratios, and decay widths of excited states in 206Pb from 4.9 to 8.1 MeV. The extracted ΣB (E 1) ↑ and ΣB (M 1) ↑ values for the total electric and magnetic dipole strength below the neutron separation energy are 0.9 ± 0.2 e2fm2 and 8.3 ± 2.0 μN2, respectively. These measurements are found to be in very good agreement with the predictions from an energy-density functional (EDF) plus quasiparticle phonon model (QPM). Such a detailed theoretical analysis allows to separate the pygmy dipole resonance from both the tail of the giant dipole resonance and multi-phonon excitations. Combined with earlier photonuclear experiments above the neutron separation energy, one extracts a value for the electric dipole polarizability of 206Pb of αD = 122 ± 10 mb /MeV. When compared to predictions from both the EDF+QPM and accurately calibrated relativistic EDFs, one deduces a range for the neutron-skin thickness of Rskin206 = 0.12- 0.19 fm and a corresponding range for the slope of the symmetry energy of L = 48- 60 MeV. This newly obtained information is also used to estimate the Maxwellian-averaged radiative cross section 205Pb (n , γ)206Pb at 30 keV to be σ = 130 ± 25 mb. The astrophysical impact of this measurement-on both the s-process in stellar nucleosynthesis and on the equation of state of neutron-rich matter-is discussed.

  2. Development of a standard data base for FBR core nuclear design (XIII). Analysis of small sample reactivity experiments at ZPPR-9

    International Nuclear Information System (INIS)

    Sato, Wakaei; Fukushima, Manabu; Ishikawa, Makoto

    2000-09-01

    A comprehensive study to evaluate and accumulate the abundant results of fast reactor physics is now in progress at O-arai Engineering Center to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as future commercial FBRs. The present report summarizes the analytical results of sample reactivity experiments at ZPPR-9 core, which has not been evaluated by the latest analytical method yet. The intention of the work is to extend and further generalize the standard data base for FBR core nuclear design. The analytical results of the sample reactivity experiments (samples: PU-30, U-6, DU-6, SS-1 and B-1) at ZPPR-9 core in JUPITER series, with the latest nuclear data library JENDL-3.2 and the analytical method which was established by the JUPITER analysis, can be concluded as follows: The region-averaged final C/E values generally agreed with unity within 5% differences at the inner core region. However, the C/E values of every sample showed the radial space-dependency increasing from center to core edge, especially the discrepancy of B-1 was the largest by 10%. Next, the influence of the present analytical results for the ZPPR-9 sample reactivity to the cross-section adjustment was evaluated. The reference case was a unified cross-section set ADJ98 based on the recent JUPITER analysis. As a conclusion, the present analytical results have sufficient physical consistency with other JUPITER data, and possess qualification as a part of the standard data base for FBR nuclear design. (author)

  3. Characterization of the neutronic fields obtained by means of neutron traps inside the nuclear reactor core IPEN/MB-01

    International Nuclear Information System (INIS)

    Mura, Luiz Ernesto Credidio

    2011-01-01

    This paper presents the results of the neutron flux values obtained from the deployment of a Flux Trap of neutrons in the reactor core IPEN/MB-01. We analyzed several configurations of Flux Traps deployed in the reactor core IPEN/MB-01 in order to get elected to Flux Trap configuration more efficient. To characterize the neutron spectrum were irradiated in the center of the Flux Trap activation detectors of different materials (Au, Sc, In, Ti, Ni). The respective gamma spectroscopy of these elements after irradiation with and without cadmium cover, provided the experimental values of the nuclear reaction rates (saturation activity) by the target nuclei and their uncertainties used as input to the code SANDBP who calculated the energy spectrum of neutrons in the center of the 'Flux-Trap' in 50 energy groups, using the input spectra calculated at the irradiation position (center of the 'Flux Trap') by codes for Reactor Physics. The results found an increase in the thermal neutron flux in the center of the Flux Trap configuration 203 for the standard configuration (default) of about 350% without having the need to increase the reactor power. We also made comparisons between the spectra obtained by SANDBP deployed, compared to those calculated by MCNP-4C code and XSDRNPM. The spatial characterization of the thermal neutron flux is made with activation foils in the form of an infinitely dilute bulk alloy of 1% Au and 99% Al in some internal points of the configuration 203 (axially to Flux Trap a nd adjacent radial) and the results showed a significant increase in the magnitude of their values when compared to standard rectangular configuration. (author)

  4. TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

    1984-06-01

    The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

  5. The nuclear core of sun: theoretical perspectives and characterization of the scientific performances of the GOLF / SOHO experiment detector

    International Nuclear Information System (INIS)

    Dzitko, H.

    1995-01-01

    The subject of this thesis is the nuclear core of the sun. The first part is theoretical and concerns neutrino flux predictions. A precise description of the solar plasma is necessary to predict boron, beryllium and CNO cycle neutrinos. We treat here the nuclear reaction rates. They are mainly determined by the cross sections and the enhancement factors due to plasma particles, the co-called screening factors. We have discussed the various possible formalisms that could be used in stellar evolution and performed direct calculations of screened cross sections. We concluded that the screening prescriptions which have been used so far in stellar evolution should be replaced by the Mitler formalism. Next, we examine the cross section uncertainties and we show that it is possible to get a better agreement between theory and experiment. Discrepancies between the gallium experiments and the calculations suggest that we should go beyond the classical solar model. This has motivated our study on possible magnetic fields deeply buried in the solar core. We discuss here the influence of a magnetic pressure perturbation on solar evolution. In the experimental part of this work, we deal with the GOLF experiment, one of the three helio-seismological experiments on board the space probe SOHO. The purpose of this instrument is the study of the global oscillation modes in the frequency range 10 -7 to 6 10 -3 Hz with a sensitivity for frequencies higher than 2 10 -4 Hz of about 1 mm/s over 20 days of continuous integration at counting rates of 12 10 6 cs/s. One part of this work was devoted to the precise characterization of the photomultipliers and their associated electronics in order to select them according to their intrinsic performances. This step was followed by long duration tests of three weeks simulating as well as possible the flight conditions. We show that the detection chain effectively meets the stability requirements of around 10 -7 by velocity measurement. We also

  6. GARLIC-B. A digital code for real-time calculation of the transient behaviour of nodal and global core and plant parameters of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Ercan, Y.; Hoeld, A.; Lupas, O.

    1982-04-01

    A program description of the code GARLIC-B is given. The code is based on a nonlinear transient model for BWR nuclear power plants which consist of a 3D-core, a top plenum, steam removal and feed water systems and a downcomer with main coolant recirculation pumps. The core is subdivided into a number of superboxes and flow channels with different coolant mass flow rates. Subcooled boiling within these channels has an important reactivity feed back effect and has to be taken also into account. The code computes the local and global core and plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. (orig./GL) [de

  7. Nuclear spins, magnetic moments and quadrupole moments of Cu isotopes from N = 28 to N = 46: probes for core polarization effects

    CERN Document Server

    Vingerhoets, P; Avgoulea, M; Billowes, J; Bissell, M L; Blaum, K; Brown, B A; Cheal, B; De Rydt, M; Forest, D H; Geppert, Ch; Honma, M; Kowalska, M; Kramer, J; Krieger, A; Mane, E; Neugart, R; Neyens, G; Nortershauser, W; Otsuka, T; Schug, M; Stroke, H H; Tungate, G; Yordanov, D T

    2010-01-01

    Measurements of the ground-state nuclear spins, magnetic and quadrupole moments of the copper isotopes from 61Cu up to 75Cu are reported. The experiments were performed at the ISOLDE facility, using the technique of collinear laser spectroscopy. The trend in the magnetic moments between the N=28 and N=50 shell closures is reasonably reproduced by large-scale shell-model calculations starting from a 56Ni core. The quadrupole moments reveal a strong polarization of the underlying Ni core when the neutron shell is opened, which is however strongly reduced at N=40 due to the parity change between the $pf$ and $g$ orbits. No enhanced core polarization is seen beyond N=40. Deviations between measured and calculated moments are attributed to the softness of the 56Ni core and weakening of the Z=28 and N=28 shell gaps.

  8. Radiation resistant fiber Bragg grating in random air-line fibers for sensing applications in nuclear reactor cores.

    Science.gov (United States)

    Zaghloul, Mohamed A S; Wang, Mohan; Huang, Sheng; Hnatovsky, Cyril; Grobnic, Dan; Mihailov, Stephen; Li, Ming-Jun; Carpenter, David; Hu, Lin-Wen; Daw, Joshua; Laffont, Guillaume; Nehr, Simon; Chen, Kevin P

    2018-04-30

    This paper reports the testing results of radiation resistant fiber Bragg grating (FBG) in random air-line (RAL) fibers in comparison with FBGs in other radiation-hardened fibers. FBGs in RAL fibers were fabricated by 80 fs ultrafast laser pulse using a phase mask approach. The fiber Bragg gratings tests were carried out in the core region of a 6 MW MIT research reactor (MITR) at a steady temperature above 600°C and an average fast neutron (>1 MeV) flux >1.2 × 10 14 n/cm 2 /s. Fifty five-day tests of FBG sensors showed less than 5 dB reduction in FBG peak strength after over 1 × 10 20 n/cm 2 of accumulated fast neutron dose. The radiation-induced compaction of FBG sensors produced less than 5.5 nm FBG wavelength shift toward shorter wavelength. To test temporal responses of FBG sensors, a number of reactor anomaly events were artificially created to abruptly change reactor power, temperature, and neutron flux over short periods of time. The thermal sensitivity and temporal responses of FBGs were determined at different accumulated doses of neutron flux. Results presented in this paper reveal that temperature-stable Type-II FBGs fabricated in radiation-hardened fibers can survive harsh in-pile conditions. Despite large parameter drift induced by strong nuclear radiation, further engineering and innovation on both optical fibers and fiber devices could lead to useful fiber sensors for various in-pile measurements to improve safety and efficiency of existing and next generation nuclear reactors.

  9. A flow test for calibrating 177 core tubes of 1/5-scale reactor flow model for Yonggwang nuclear units 3 and 4

    International Nuclear Information System (INIS)

    Lee, Byung Jin; Jang, Ho Cheol; Cheong, Jong Sik; Kuh, Jung Eui

    1990-01-01

    A flow test was performed to find out the hydraulic characteristics of every one of 177 core tubes, representing a fuel assembly respectively, as a preparatory step of 1/5 scale reactor flow model test for Yonggwang Nuclear Units (hereafter YGN) 3 and 4. The axial hydraulic resistance of the fuel assembly was simulated in the square core tube with six orifice plates positioned along the tube length; core support structure below each fuel assembly was done in the core upstream geometry section of the test loop. For each core tube the pressure differentials across the inlet, exit orifice plate and overall tube length were measured, along with the flow rates and temperatures of the test fluid. The measured pressure drops were converted to pressure loss or flow metering coefficients. The metering coefficient of the inlet orifice plate was sensitive to the configuration and location of the upstream geometry. The hydraulic resistance of the core tubes were reasonably coincided with a target value and consistent. The polynomial curve fits of the calibrated coefficients for the 177 core tubes were obtained with reasonable data scatters

  10. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  11. Transient core characteristics of small molten salt reactor coupling problem between heat transfer/flow and nuclear fission reaction

    International Nuclear Information System (INIS)

    Yamamoto, Takahisa; Mitachi, Koshi

    2004-01-01

    This paper performed the transient core analysis of a small Molten Salt Reactor (MSR). The emphasis is that the numerical model employed in this paper takes into account the interaction among fuel salt flow, nuclear reaction and heat transfer. The model consists of two group diffusion equations for fast and thermal neutron fluexs, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis are that (1) fission reaction (heat generation) rate significantly increases soon after step reactivity insertion, e.g., the peak of fission reaction rate achieves about 2.7 times larger than the rated power 350 MW when the reactivity of 0.15% Δk/k 0 is inserted to the rated state, and (2) the self-control performance of the small MSR effectively works under the step reactivity insertion of 0.56% Δk/k 0 , putting the fission reaction rate back on the rated state. (author)

  12. Interactive Real-time Simulation of a Nuclear Reactor Emergency Core Cooling System on a Desktop Computer

    International Nuclear Information System (INIS)

    Muncharoen, C.; Chanyotha, S.; Bereznai, G.

    1998-01-01

    The simulation of the Emergency Core Cooling System for a 900 MW nuclear power plant has been developed by using object oriented programming language. It is capable of generating code that executes in real-time on a PENTIUM 100 or equivalent personal computer. Graphical user interface ECCS screens have been developed using Lab VIEW to allow interactive control of ECCS. The usual simulator functions, such as freeze, run, iterate, have been provided, and a number of malfunctions may be activated. A large pipe break near the reactor inlet header has been simulated to verify the response of the ECCS model. LOCA detection, ECC initiation, injection and recovery phased are all modeled, and give results consistent with safety analysis data for a 100% break. With stand alone ECCS simulation, the changes of flow and pressure in ECCS can be observed. The operator can study operational procedures and get used to LOCA in case of the LOCA. Practicing with malfunction, the operator will improve problem solving skills and gain a deeper comprehension of ECCS

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  14. 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.

  15. Holddown arrangement for removable cover of a pressurized nuclear-reactor core and method of using same

    International Nuclear Information System (INIS)

    Beine, B.

    1976-01-01

    A pressurized nuclear-reactor core is surrounded by a burst shield having a side wall formed with a plurality of longitudinal passages in which are received longitudinal prestressing elements whose upper ends extend beyond the upper edge of the side wall. The cover is formed with a plurality of holes that register with the passages in the side wall so that the cover can be set over the top of the side wall with the upper ends of the prestressing elements projecting beyond the cover. Each prestressing element is provided at its upper end with an anchor body which can bear in force-transmitting relationship either with the side wall of the burst shield through a sleeve received in the cover and having a lower end standing on the upper edge of the side wall and an upper end adjacent the anchor body, or with the cover by means of a removable nut screwed on to the anchor body and engageable in force-transmitting relationship with the cover. In use the anchor body is dimensioned to pass through the cover on the top of the side wall so that when this cover is to be removed the nut is unscrewed and the anchor body bears through the sleeve on the upper edge of the side wall in order that the side wall not be destressed. 8 claims, 6 drawing figures

  16. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

  17. Mathematical model for the preliminary analysis of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239a

    Energy Technology Data Exchange (ETDEWEB)

    Grey, J.; Chow, S.

    1976-06-30

    The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. Such a concept could be particularly useful for missions which require both relatively high acceleration (e.g., for planetocentric maneuvers) and high performance at low acceleration (e.g., on heliocentric trajectories or for trajectory shaping). The first volume develops the mathematical model of the system.

  18. Application of TEMPPC code to the IEA-R1 nuclear reactor core hydrothermal calculations operating at 2 MW for determining the minimal coolant flow

    International Nuclear Information System (INIS)

    Frajndlich, R.; Sousa, J.A. de.

    1985-01-01

    A thermohydraulic study of the IEA-R1 nuclear reactor core on steady-state operating condition and forced convection, is presented. The objective of this calculation is to obtain the minimal flow rate of coolant necessary at the reactor core, limited by the temperature associated to the beginning of nucleate boiling over the fuel plates at a normal operating power (2MW) for a certain inlet coolant temperature. The coolant system safety level is also calculated in this paper, which is divided in three steps: thermohydraulic calculation, without using the uncertainty factors and, after that, considering these factor by two methods: the statistical and the conventional ones. Whichever the method accepted, the results obtained by the program TEMPPC show a great safety margin with respect to the termohydraulic parameters from the IEA-R1 nuclear reactor. (Author) [pt

  19. 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

  20. Natural analogue of nuclear waste glass in a geologic formation. Study on long-term behavior of volcanic glass shards collected from drill cores

    International Nuclear Information System (INIS)

    Yoshikawa, Hideki; Yui, Mikazu; Futakuchi, Katsuhito; Hiroki, Minenari

    2005-01-01

    Alteration of the volcanic glass in geologic formation was investigated as one of the natural analog for a glass of high-level nuclear waste in geological disposal. We analyzed some volcanic glasses included in the core sample of the bore hole and estimated the history of its burying and observed its alteration using the polarizing microscope. Some information at the piling up temperature and the piling up time was collected. (author)

  1. Physical start up of the Dalat nuclear research reactor with the core configuration having a central neutron trap

    International Nuclear Information System (INIS)

    Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

    1994-01-01

    After the reactor has reached physical criticality with the core configuration exempt from central neutron trap on 1 November 1983, the core configuration with a central neutron trap has been arranged in the reactor and the reactor has reached physical criticality with this core configuration at 17h48 on 18 December 1983. The integral worths of different control rods are determined with accuracy. 2 refs., 24 figs., 18 tabs

  2. Ex-core detector response caused by control rod misalignment observed during operation of the reactor on the nuclear ship Mutsu

    International Nuclear Information System (INIS)

    Itagaki, Masafumi; Miyoshi, Yoshinori; Gakuhari, Kazuhiko; Okada, Noboru; Sakai, Tomohiro

    1993-01-01

    Unexpected deviations of ex-core neutron detector signals were observed during a voyage of the Japanese nuclear ship, Mutsu. From detailed three-dimensional analyses, this phenomenon was determined to be caused by an asymmetrical neutron source distribution in the core due to a small misalignment between the two control rods of a control rod group. A systematic ex-core detector response experiment was performed during the Mutsu's third experimental voyage to gain some understanding of the relationship between the control rod pattern and the detector response characteristics. Results obtained from analyses of the experiment indicate that the Crump-Lee technique, using calculated three-dimensional source distributions for various control rod patterns, provides good agreement between the calculated and measured detector responses. Xenon transient analyses were carried out to generate accurate three-dimensional source distributions for predicting the time-dependent detector response characteristics. Two types of ex-core detector responses are caused by changes in the control rod pattern in the Mutsu reactor: the detector response ratio tends to decrease with the withdrawal of a group of control rods as a pair, and a difference in the positions of the control rods in a group causes signal deviations among the four ex-core detectors. Control rod misalignment does not greatly affect the mean value of the four detector signals, and the deviation can be minimized if the two rods within a group are set at the same elevation at the time of detector calibration

  3. Guides about nuclear energy in South Korea; Reperes sur l'energie nucleaire en Coree du Sud

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-03-01

    This document summarizes the main information on nuclear energy in South Korea: number of reactors in operation, type, date of commissioning, nuclear facilities under construction, nuclear share in power production, companies and organizations (Korea electric power company (KEPCO), Korea atomic energy institute (KAERI), Korea institute of nuclear safety (KINS), Korea nuclear energy foundation (KNEF), Korea hydro and nuclear power (KHNP), nuclear environment technology institute (NETEC), Korea basic science institute (KBSI)), nuclear fuel fabrication, research works on waste disposal, nuclear R and D in fission and fusion, safety of nuclear facilities, strategies under study (1000 MWe Korea standard nuclear power plant (KSNP), 1400 MWe advanced power reactor (APR), small power water cooled reactors (system-integrated modular advanced reactor (SMART) research program), development of fast reactors (Kalimer research program), development of the process of direct use of PWR fuel in Candu (DUPIC), use of reprocessing uranium, transmutation of trans-uranian and wastes (KOMAC program), first dismantling experience (Triga Mark II and III research reactors). (J.S.)

  4. Fluid structure interaction studies on acoustic load response of light water nuclear reactor core internals under blowdown condition

    International Nuclear Information System (INIS)

    Moses Lemuel Raj, G.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

    1998-12-01

    Acoustic load evaluation within two phase medium and the related fluid-structure interaction analysis in case of Loss of Coolant Accidents (LOCA) for light water reactor systems is an important inter-disciplinary area. The present work highlights the development of a three-dimensional finite element code FLUSHEL to analyse LOCA induced depressurization problems for Pressurised Water Reactor (PWR) core barrel and Boiling Water Reactor (BWR) core shroud. With good comparison obtained between prediction made by the present code and the experimental results of HDR-PWR test problem, coupled fluid-structure interaction analysis of core shroud of Tarapur Atomic Power Station (TAPS) is presented for recirculation line break. It is shown that the acoustic load induced stresses in the core shroud are small and downcomer acoustic cavity modes are decoupled with the shell multi-lobe modes. Thus the structural integrity of TAPS core shroud for recirculation line break induced acoustic load is demonstrated. (author)

  5. Development of core design/analysis technology for integral reactor; verification of SMART nuclear design by Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Hong, In Seob; Han, Beom Seok; Jeong, Jong Seong [Seoul National University, Seoul (Korea)

    2002-03-01

    The objective of this project is to verify neutronics characteristics of the SMART core design as to compare computational results of the MCNAP code with those of the MASTER code. To achieve this goal, we will analyze neutronics characteristics of the SMART core using the MCNAP code and compare these results with results of the MASTER code. We improved parallel computing module and developed error analysis module of the MCNAP code. We analyzed mechanism of the error propagation through depletion computation and developed a calculation module for quantifying these errors. We performed depletion analysis for fuel pins and assemblies of the SMART core. We modeled a 3-D structure of the SMART core and considered a variation of material compositions by control rods operation and performed depletion analysis for the SMART core. We computed control-rod worths of assemblies and a reactor core for operation of individual control-rod groups. We computed core reactivity coefficients-MTC, FTC and compared these results with computational results of the MASTER code. To verify error analysis module of the MCNAP code, we analyzed error propagation through depletion of the SMART B-type assembly. 18 refs., 102 figs., 36 tabs. (Author)

  6. Nuclear facilities maintenance in the core of management-advanced trend in IBM Maximo asset management applications

    International Nuclear Information System (INIS)

    Seino, Satoshi; Ujihara, Satoshi; Kikuyama, Kaoru

    2009-01-01

    European and US plant owners have attached importance to plant maintenance, such as prompt grasp of plant states, implementation of maintenance and planning of maintenance programs, as one of asset management. The US advanced trend was introduced in this feature article through the applications of IBM Maximo Asset Management for nuclear facilities maintenance. World trends of nuclear power and related problems, need of nuclear facilities management, key items for introduction of maintenance management systems, required systems for nuclear maintenance management and introduction of functions of the IBM strategic asset management solution-Maximo were described respectively. (T. Tanaka)

  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. Distribution of radionuclides in a marine sediment core off the waterspout of the nuclear power plants in Daya Bay, northeastern South China Sea

    International Nuclear Information System (INIS)

    Zhou, Peng; Li, Dongmei; Li, Haitao; Fang, Hongda; Huang, Chuguang; Zhang, Yusheng; Zhang, Hongbiao; Zhao, Li; Zhou, Junjie; Wang, Hua; Yang, Jie

    2015-01-01

    A sediment core was collected and dated using 210 Pb ex dating method off the waterspout of nuclear power base of Daya Bay, northeastern South China Sea. The γ-emitting radionuclides were analyzed using HPGe γ spectrometry, gross alpha and beta radioactivity as well as other geochemical indicators were deliberated to assess the impact of nuclear power plants (NPP) operation and to study the past environment changes. It suggested that NPP provided no new radioactivity source to sediment based on the low specific activity of 137 Cs. Two broad peaks of TOC, TC and LOI accorded well with the commercial operations of Daya Bay NPP (1994.2 and 1994.5) and LNPP Phase I (2002.5 and 2003.3), implying that the mass input of cooling water from NPP may result into a substantial change in the ecological environment and Daya Bay has been severely impacted by human activities. - Graphical abstract: A sediment core was collected and dated using 210 Pb ex dating method, off the waterspout of nuclear power base of Daya Bay, northeastern South China Sea. The γ-emitting radionuclides analyzed using the HPGe γ spectrometry, gross alpha and beta radioactivity as well as other geochemical indicators were deliberated to study the past environment changes and assess the impact of nuclear power plants (NPP) operation. NPP provided no new radioactivity source to sediment, but the mass input of cooling water from nuclear power plants may result into a substantial change in the ecological environment and Daya Bay has been severely impacted by human activities. - Highlights: • A sediment core collected from Daya Bay, South China Sea was dated using 210 Pb ex method. • The γ-emitting radionuclides, gross α and β, TOC, TIC, TC, LOI were deliberated to assess the impact of nuclear power plants (NPP) operation. • The low activity of 137 Cs in sediment suggested NPP provided no new radioactivity source. • Two peaks of TOC, TC and LOI implied that the mass input of NPP cooling water may

  9. Restraint system for core elements of a reactor core

    International Nuclear Information System (INIS)

    Class, G.

    1975-01-01

    In a nuclear reactor, a core element bundle formed of a plurality of side-by-side arranged core elements is surrounded by restraining elements that exert a radially inwardly directly restraining force generating friction forces between the core elements in a restraining plane that is transverse to the core element axes. The adjoining core elements are in rolling contact with one another in the restraining plane by virtue of rolling-type bearing elements supported in the core elements. (Official Gazette)

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

    International Nuclear Information System (INIS)

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

    1986-11-01

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

  11. Improvement of core monitoring code cecor by the virtual segmentation of the self powered neutron detector loaded at Korean Standard Nuclear Plant

    International Nuclear Information System (INIS)

    Choi, T.; Jung, Y.S.

    2006-01-01

    Full text: Full text: Korean Standard Nuclear Plant uses Self Powered Neutron Detectors (SPNDs) to measure the neutron flux in the reactor core. The SPND's height is 40 cm and is located axially at the five different positions and 45 radial places. The design code simulated a reactor core is calculated by segmentation of the core. The segmentation is called as 'node', of which size is normally 20 cm. The axial height of the detector is larger than that of the node, and the larger detector's height maybe product some error on the axially complex shape. The analysis with the detector's signals showed some errors at the non-cosine axial flux shape. In order to reduce the errors for the shape, we tried to divide the detector by introducing the virtual boundary in the detector. Then, each axially 5 detectors had two virtual segmentations respectively and the detector's signal was divided by the inputs. So the more virtual detector's signals were gotten, the more accurate axial shape was produced. The result with virtual segmentations in a detector gave less deviation than the case without virtual segmentation (the current model). After the middle of cycle at the initial core specially, the axial neutron flux shape is changed to the saddle type one. The current model gave some error in Root Mean Square (RMS) between the measured value and the calculated one. The virtual segmentation model gave the better agreement at that time

  12. In-core program for on line measurements of neutron, photon and nuclear heating parameters inside Jules Horowitz MTR reactor

    International Nuclear Information System (INIS)

    Lyoussi, A.; Reynard-Carette, C.

    2014-01-01

    Accurate on-line measurements of key parameters inside experimental channels of Material Testing Reactor are necessary to dimension the irradiation devices and consequently to conduct smart experiments on fuels and materials under suitable conditions. In particular the quantification of nuclear heating, a relevant parameter to reach adapted thermal conditions, has to be improved. These works focus on an important collaborative program between CEA and Aix-Marseille University called INCORE (Instrumentation for Nuclear radiations and Calorimetry On-line in Reactor) dedicated to the development of a new measurement methodology to quantify both nuclear heating and accurate radiation flux levels (neutrons and photons). The methodology, which is based on experiments carried out under irradiation conditions with a multi-sensor device (ionization chamber, fission chamber, gamma thermometer, calorimeter, SPND, SPGD) as well as works performed out-of nuclear/radiative environment on a reference sensor used to measure nuclear heating (calorimeter), is presented (authors)

  13. Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

    2015-01-01

    As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

  14. Synthesis of the IRSN report on the topic of water way answers to implement in case of accident with core meltdown occurring on operating pressurized water nuclear reactors

    International Nuclear Information System (INIS)

    2009-06-01

    This report briefly discusses the efficiency of technical measures adopted for the implementation of water ways as answers to an accident with core meltdown in operating pressurized water nuclear reactors. While mentioning the importance of the hydro-geological characteristics of the various sites, the IRSN asks EDF to plan and implement means to prevent any rejection through water ways for some of these sites, to investigate the possibility of building a geotechnical enclosure, to define a storing-control-treatment-rejection chain which would guarantee an efficient management of the water to be pumped, to study retention phenomena for strontium and caesium isotopes in sands and gravels

  15. Computer code and users' guide for the preliminary analysis of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239b

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, R.A.; Smith, W.W.

    1976-06-30

    The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. The second volume describes the computer code and users' guide for the preliminary analysis of the system.

  16. Distribution of radionuclides in a marine sediment core off the waterspout of the nuclear power plants in Daya Bay, northeastern South China Sea.

    Science.gov (United States)

    Zhou, Peng; Li, Dongmei; Li, Haitao; Fang, Hongda; Huang, Chuguang; Zhang, Yusheng; Zhang, Hongbiao; Zhao, Li; Zhou, Junjie; Wang, Hua; Yang, Jie

    2015-07-01

    A sediment core was collected and dated using (210)Pbex dating method off the waterspout of nuclear power base of Daya Bay, northeastern South China Sea. The γ-emitting radionuclides were analyzed using HPGe γ spectrometry, gross alpha and beta radioactivity as well as other geochemical indicators were deliberated to assess the impact of nuclear power plants (NPP) operation and to study the past environment changes. It suggested that NPP provided no new radioactivity source to sediment based on the low specific activity of (137)Cs. Two broad peaks of TOC, TC and LOI accorded well with the commercial operations of Daya Bay NPP (1994.2 and 1994.5) and LNPP Phase I (2002.5 and 2003.3), implying that the mass input of cooling water from NPP may result into a substantial change in the ecological environment and Daya Bay has been severely impacted by human activities. Copyright © 2015 Elsevier Ltd. All rights reserved.

  17. 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

  18. Optimization of fuel core loading pattern design in a VVER nuclear power reactors using Particle Swarm Optimization (PSO)

    International Nuclear Information System (INIS)

    Babazadeh, Davood; Boroushaki, Mehrdad; Lucas, Caro

    2009-01-01

    The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor (K eff ) in order to extract the maximum energy, and keeping the local power peaking factor (P q ) lower than a predetermined value to maintain fuel integrity. In this research, a new strategy based on Particle Swarm Optimization (PSO) algorithm has been developed to optimize the fuel core loading pattern in a typical VVER. The PSO algorithm presents a simple social model by inspiration from bird collective behavior in finding food. A modified version of PSO algorithm for discrete variables has been developed and implemented successfully for the multi-objective optimization of fuel loading pattern design with constraints of keeping P q lower than a predetermined value and maximizing K eff . This strategy has been accomplished using WIMSD and CITATION calculation codes. Simulation results show that this algorithm can help in the acquisition of a new pattern without contravention of the constraints.

  19. Nuclear reactor with a fixed system of neutron poison, which can be burnt up, introduced into the reactor core

    International Nuclear Information System (INIS)

    Mueller, E.; Roegler, H.J.; Wickert, M.

    1985-01-01

    The fixed system consists of neutron poison which can be burnt up, in an uneven distribution, and with adjustable absorber rods for output control, which are driven into the reactor core from the side along the fuel elements. There is an excess of neutron poison which can be burnt up, overall, on the side of the reactor core away from the absorber rods. The reactor core is free of neutron poison which can be burnt up on the side where the absorber rods are driven in, so that the ratio of maximum to mean power density with reference to a possible absorber rod positions is less than for homogeneous distribution of the neutron poison which can be burnt up. (orig./HP) [de

  20. Feasibility study of passive gamma spectrometry of molten core material from Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy - low-volatile FP and special nuclear material inventory analysis and fundamental characteristics of gamma-rays from fuel debris

    International Nuclear Information System (INIS)

    Sagara, Hiroshi; Tomikawa, Hirofumi; Watahiki, Masaru; Kuno, Yusuke

    2014-01-01

    The technologies applied to the analysis of the Three Mile Island accident were examined in a feasibility study of gamma spectrometry of molten core material from the Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy. The focus is on low-volatile fission products and heavy metal inventory analysis, and the fundamental characteristics of gamma-rays from fuel debris with respect to passive measurements. The inventory ratios of the low-volatile lanthanides, "1"5"4Eu and "1"4"4Ce, to special nuclear materials were evaluated by the entire core inventories in units 1, 2, and 3 with an estimated uncertainty of 9%-13% at the 1σ level for homogenized molten fuel material. The uncertainty is expected to be larger locally owing to the use of the irradiation cycle averaging approach. The ratios were also evaluated as a function of burnup for specific fuel debris with an estimated uncertainty of 13%-25% at the 1σ level for units 1 and 2, and most of the fuels in unit 3, although the uncertainty regarding the separated mixed oxide fuel in unit 3 would be significantly higher owing to the burnup dependence approach. Source photon spectra were also examined and cooling-time-dependent data sets were prepared. The fundamental characteristics of high-energy gamma-rays from fuel debris were investigated by a bare-sphere model transport calculation. Mass attenuation coefficients of fuel debris were evaluated to be insensitive to its possible composition in a high-energy region. The leakage photon ratio was evaluated using a variety of parameters, and a significant impact was confirmed for a certain size of fuel debris. Its correlation was summarized with respect to the leakage photopeak ratio of source "1"5"4Eu. Finally, a preliminary study using a hypothetical canister model of fuel debris based on the experience at Three Mile Island was presented, and future plans were introduced. (author)

  1. Final Report, Nuclear Energy Research Initiative (NERI) Project: An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model

    International Nuclear Information System (INIS)

    Anistratov, Dmitriy Y.; Adams, Marvin L.; Palmer, Todd S.; Smith, Kord S.; Clarno, Kevin; Hikaru Hiruta; Razvan Nes

    2003-01-01

    OAK (B204) Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model'' The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations

  2. Method and device for catching reactor core melt-down masses in hypothetical accidents of nuclear power plants

    International Nuclear Information System (INIS)

    Morlock, G.; Wiesemes, J.; Bachner, D.

    1977-01-01

    The device is to receive the afterheat of the molten core and in this way to prevent afterflow of coolant and a new criticality. A tank below the reactor pressure vessel, with the proper diameter, contains a store of salt or a salt mixture suitable to receive the afterheat of a core melt-down as heat of fusion or conversion. Above the salt, there is a layer of thermoplastics or of a material forming a hardening foam. Coolant eventually continuing to flow out is separated from the core melt by this barrier layer, and thus the build-up of high steam pressures is prevented. Neutron-absorbing materials, like boron salts mixed to the salts, as well as a subdivision of the salt surface, e.g. by means of canalizing firebricks, prevent the formation of new criticality. Further installations within the tank, like pipings or channels, permit the introduction of water after cooling down of the core or salt melt-down mass and to wash out the brine with all radioactive and other constituents for transport to reprocessing or ultimate storage. (HP) [de

  3. 48{sup th} Annual meeting on nuclear technology (AMNT 2017). Key topic / Enhanced safety and operation excellence. Focus session: Uncertainty analyses in reactor core simulations

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, Winfried [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Forschungszentrum

    2017-12-15

    The supplementation of reactor simulations by uncertainty analyses is becoming increasingly important internationally due to the fact that the reliability of simulation calculations can be significantly increased by the quantification of uncertainties in comparison to the use of so-called conservative methods (BEPU- ''Best-Estimate plus Uncertainties''). While systematic uncertainty analyses for thermo-hydraulic calculations have been performed routinely for a long time, methods for taking into account uncertainties in nuclear data, which are the basis for neutron transport calculations, are under development. The Focus Session Uncertainty Analyses in Reactor Core Simulations was intended to provide an overview of international research and development with respect to supplementing reactor core simulations with uncertainty and sensitivity analyses, in research institutes as well as within the nuclear industry. The presented analyses not only focused on light water reactors, but also on advanced reactor systems. Particular emphasis was put on international benchmarks in the field. The session was chaired by Winfried Zwermann (Gesellschaft fuer Anlagen- und Reaktorsicherheit).

  4. Development of a fiber-coupled laser-induced breakdown spectroscopy instrument for analysis of underwater debris in a nuclear reactor core

    International Nuclear Information System (INIS)

    Saeki, Morihisa; Iwanade, Akio; Ohba, Hironori; Ito, Chikara; Wakaida, Ikuo; Thornton, Blair; Sakka, Tetsuo

    2014-01-01

    To inspect the post-accident nuclear core reactor of the TEPCO Fukushima Daiichi nuclear power plant (F1-NPP), a transportable fiber-coupled laser-induced breakdown spectroscopy (LIBS) instrument has been developed. The developed LIBS instrument was designed to analyze underwater samples in a high-radiation field by single-pulse breakdown with gas flow or double-pulse breakdown. To check the feasibility of the assembled fiber-coupled LIBS instrument for the analysis of debris material (mixture of the fuel core, fuel cladding, construction material and so on) in the F1-NPP, we investigated the influence of the radiation dose on the optical transmittance of the laser delivery fiber, compared data quality among various LIBS techniques for an underwater sample and studied the feasibility of the fiber-coupled LIBS system in an analysis of the underwater sample of the simulated debris in F1-NPP. In a feasible study conducted by using simulated debris, which was a mixture of CeO 2 (surrogate of UO 2 ), ZrO 2 and Fe, we selected atomic lines suitable for the analysis of materials, and prepared calibration curves for the component elements. The feasible study has guaranteed that the developed fiber-coupled LIBS system is applicable for analyzing the debris materials in the F1-NPP. (author)

  5. Repair pathways independent of the Fanconi anemia nuclear core complex play a predominant role in mitigating formaldehyde-induced DNA damage

    International Nuclear Information System (INIS)

    Noda, Taichi; Takahashi, Akihisa; Kondo, Natsuko; Mori, Eiichiro; Okamoto, Noritomo; Nakagawa, Yosuke; Ohnishi, Ken; Zdzienicka, Malgorzata Z.; Thompson, Larry H.; Helleday, Thomas; Asada, Hideo

    2011-01-01

    The role of the Fanconi anemia (FA) repair pathway for DNA damage induced by formaldehyde was examined in the work described here. The following cell types were used: mouse embryonic fibroblast cell lines FANCA -/- , FANCC -/- , FANCA -/- C -/- , FANCD2 -/- and their parental cells, the Chinese hamster cell lines FANCD1 mutant (mt), FANCGmt, their revertant cells, and the corresponding wild-type (wt) cells. Cell survival rates were determined with colony formation assays after formaldehyde treatment. DNA double strand breaks (DSBs) were detected with an immunocytochemical γH2AX-staining assay. Although the sensitivity of FANCA -/- , FANCC -/- and FANCA -/- C -/- cells to formaldehyde was comparable to that of proficient cells, FANCD1mt, FANCGmt and FANCD2 -/- cells were more sensitive to formaldehyde than the corresponding proficient cells. It was found that homologous recombination (HR) repair was induced by formaldehyde. In addition, γH2AX foci in FANCD1mt cells persisted for longer times than in FANCD1wt cells. These findings suggest that formaldehyde-induced DSBs are repaired by HR through the FA repair pathway which is independent of the FA nuclear core complex. -- Research highlights: → We examined to clarify the repair pathways of formaldehyde-induced DNA damage. Formaldehyde induces DNA double strand breaks (DSBs). → DSBs are repaired through the Fanconi anemia (FA) repair pathway. → This pathway is independent of the FA nuclear core complex. → We also found that homologous recombination repair was induced by formaldehyde.

  6. Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

    Energy Technology Data Exchange (ETDEWEB)

    Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2016-05-15

    Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

  7. Development of a standard database for FBR core nuclear design (XI). Analysis of the Experimental Fast Reactor 'JOYO' MK-I start-up test and operation data

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2000-03-01

    As a recent research, Japan Nuclear Cycle Development Institute (JNC) develops a database of integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor 'JOYO' MK-I core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. On the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of 'JOYO' MK-I core in comparison with ZPPR-9 core of JUPITER experiments. (J.P.N.)

  8. Nuclear data sensitivity and uncertainty for the Canadian supercritical water-cooled reactor II: Full core analysis

    International Nuclear Information System (INIS)

    Langton, S.E.; Buijs, A.; Pencer, J.

    2015-01-01

    Highlights: • H-2, Pu-239, and Th-232 make large contributions to SCWR modelling sensitivity. • H-2, Pu-239, and Th-232 make large contributions to SCWR modelling uncertainty. • Isotopes of Zr make large contributions to SCWR modelling uncertainty. - Abstract: Uncertainties in nuclear data are a fundamental source of uncertainty in reactor physics calculations. To determine their contribution to uncertainties in calculated reactor physics parameters, a nuclear data sensitivity and uncertainty study is performed on the Canadian supercritical water reactor (SCWR) concept. The nuclear data uncertainty contributions to the neutron multiplication factor k eff are 6.31 mk for the SCWR at the beginning of cycle (BOC) and 6.99 mk at the end of cycle (EOC). Both of these uncertainties have a statistical uncertainty of 0.02 mk. The nuclear data uncertainty contributions to Coolant Void Reactivity (CVR) are 1.0 mk and 0.9 mk for BOC and EOC, respectively, both with statistical uncertainties of 0.1 mk. The nuclear data uncertainty contributions to other reactivity parameters range from as low as 3% of to as high as ten times the values of the reactivity coefficients. The largest contributors to the uncertainties in the reactor physics parameters are Pu-239, Th-232, H-2, and isotopes of zirconium

  9. PWR in-core nuclear fuel management optimization utilizing nodal (non-linear NEM) generalized perturbation theory

    International Nuclear Information System (INIS)

    Maldonado, G.I.; Turinsky, P.J.; Kropaczek, D.J.

    1993-01-01

    The computational capability of efficiently and accurately evaluate reactor core attributes (i.e., k eff and power distributions as a function of cycle burnup) utilizing a second-order accurate advanced nodal Generalized Perturbation Theory (GPT) model has been developed. The GPT model is derived from the forward non-linear iterative Nodal Expansion Method (NEM) strategy, thereby extending its inherent savings in memory storage and high computational efficiency to also encompass GPT via the preservation of the finite-difference matrix structure. The above development was easily implemented into the existing coarse-mesh finite-difference GPT-based in-core fuel management optimization code FORMOSA-P, thus combining the proven robustness of its adaptive Simulated Annealing (SA) multiple-objective optimization algorithm with a high-fidelity NEM GPT neutronics model to produce a powerful computational tool used to generate families of near-optimum loading patterns for PWRs. (orig.)

  10. Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core

    International Nuclear Information System (INIS)

    Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G.

    2006-01-01

    Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of 62 149 Sm and its dependence on the shift of a resonance position E r (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73≤ΔE r ≤62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant α. We obtain new, more accurate limits of -4x10 -17 ≤α·/α≤3x10 -17 yr -1 . Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress

  11. Physical start up of the Dalat nuclear research reactor with the core configuration exempt from neutron trap

    International Nuclear Information System (INIS)

    Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

    1994-01-01

    The nominal power of the reconstructed Dalat reactor is of 500 KW. After a meticulous preparation the Russian and Vietnamese teams have proceeded to the physical reactor start-up in November 1983 with the core configuration exempt from the neutron trap. The reactor has reached the physical criticality at 19h50 on 1 November 1983. The report delineates different steps of the start-up procedure. 2 refs., 3 figs., 7 tabs

  12. Nuclear Bombs and Coral: Guam Coral Core Reveals Operation-Specific Radiocarbon Signals from the Pacific Proving Grounds

    Science.gov (United States)

    Andrews, A. H.

    2016-12-01

    Radiocarbon (14C) analyses on a coral core extracted from the western Central Pacific (Guam) has revealed a series of early peaks in the marine bomb 14C record. The typical marine bomb 14C signal, one that is phase lagged and attenuated relative to atmospheric bomb 14C, is present in the coral core and is consistent with other North Pacific records. However, 14C levels that are well above what can be explained by air-sea diffusion alone punctuate this pattern. This anomaly has been demonstrated to a limited extent in other coral cores of the Indo-Pacific region, but is unmatched relative to the magnitude and temporal resolution recorded in the Guam coral core. Other records have shown an early Δ14C rise on the order of 40-50‰ above pre-bomb levels, with a subsequent decline before continuing the gradual Δ14C rise that is indicative of air-sea diffusion of 14CO2. The Guam coral Δ14C record provided three strong pulses in 1954-55, 1956-57, and 1958-59 that are superimposed on the pre-bomb to initial Δ14C rise from atmospheric bomb 14C. Each of these peaks can be directly linked to testing of thermonuclear devices in the Pacific Proving Grounds at Eniwetok and Bikini Atoll of the Marshall Islands. The measurable lag in reaching Guam can be tied to ocean surface currents and can be traced to other regional Δ14C records from corals, providing a transport timeline to places as distant as the Indonesian throughflow, Okinawa and Palmyra.

  13. Design study of nuclear power systems for deep space explorers. (2) Electricity supply capabilities of solid cores

    International Nuclear Information System (INIS)

    Yamaji, Akifumi; Takizuka, Takakazu; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime

    2009-01-01

    This study has been carried out in series with the other study, 'Criticality of Low Enriched Uranium Fueled Core' to explore the possibilities of a solid reactor electricity generation system for supplying propulsion power of a deep space explorer. The design ranges of two different systems are determined with respect to the electric power, the radiator mass, and the operating temperatures of the heat-pipes and thermoelectric converters. The two systems are the core surface cooling with heat-pipe system (CSHP), and the core direct cooling with heat-pipe system (CDHP). The evaluated electric powers widely cover the 1 to 100 kW range, which had long been claimed to be the range that lacked the power sources in space. Therefore, the concepts shown by this study may lead to a breakthrough of the human activities in space. The working temperature ranges of the main components, namely the heat-pipes and thermoelectric converters, are wide and covers down to relatively low temperatures. This is desirable from the viewpoints of broadening the choices, reducing the development needs, and improving the reliabilities of the devices. Hence, it is advantageous for an early establishment of the concept. (author)

  14. 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)

  15. The fuzzy clearing approach for a niching genetic algorithm applied to a nuclear reactor core design optimization problem

    International Nuclear Information System (INIS)

    Sacco, Wagner F.; Machado, Marcelo D.; Pereira, Claudio M.N.A.; Schirru, Roberto

    2004-01-01

    This article extends previous efforts on genetic algorithms (GAs) applied to a core design optimization problem. We introduce the application of a new Niching Genetic Algorithm (NGA) to this problem and compare its performance to these previous works. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a three-enrichment zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. After exhaustive experiments we observed that our new niching method performs better than the conventional GA due to a greater exploration of the search space

  16. Cooling system of the core of a nuclear reactor while it is being stopped or normally operating

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1986-01-01

    The present invention proposes a cooling system with intermediate gas flow which ensures the reactor core cooling when the primary pumps are stopped either directly by means of main heat-exchange circuits ensuring normally the reactor operation, or by means of separated loops, these ones being able so to operate in an autonomous way for they produce their own electricity needs and also an excedent which is added to the power plant production. The cooling circuit and the heat exchanger are described in detail [fr

  17. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    International Nuclear Information System (INIS)

    Walton, J.T.

    1992-11-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code

  18. Operating system design of parallel computer for on-line management of nuclear pressurised water reactor cores

    International Nuclear Information System (INIS)

    Gougam, F.

    1991-04-01

    This study is part of the PHAETON project which aims at increasing the knowledge of safety parameters of PWR core and reducing operating margins during the reactor cycle. The on-line system associates a simulator process to compute the three dimensional flux distribution and an acquisition process of reactor core parameters from the central instrumentation. The 3D flux calculation is the most time consuming. So, for cost and safety reasons, the PHAETON project proposes an approach which is to parallelize the 3D diffusion calculation and to use a computer based on parallel processor architecture. This paper presents the design of the operating system on which the application is executed. The routine interface proposed, includes the main operations necessary for programming a real time and parallel application. The primitives include: task management, data transfer, synchronisation by event signalling and by using the rendez-vous mechanisms. The primitives which are proposed use standard softwares like real-time kernel and UNIX operating system [fr

  19. Experimental Study of Hydraulic Control Rod Drive Mechanism for Passive IN-core Cooling System of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    CAREM 25 (27 MWe safety systems using hydraulic control rod drives (CRD) studied critical issues that were rod drops with interrupted flow [3]. Hydraulic control rod drive suggested fast shutdown condition using a large gap between piston and cylinder in order to fast drop of neutron absorbing rods. A Passive IN-core Cooling system (PINCs) was suggested for safety enhancement of pressurized water reactors (PWR), small modular reactor (SMR), sodium fast reactor (SFR) in UNIST. PINCs consist of hydraulic control rod drive mechanism (Hydraulic CRDM) and hybrid control rod assembly with heat pipe combined with control rod. The schematic diagram of the hydraulic CRDM for PINCs is shown in Fig. 1. The experimental results show the steady state and transient behavior of the upper cylinder at a low pressure and low temperature. The influence of the working fluid temperature and cylinder mass are investigated. Finally, the heat removal between evaporator section and condenser section is compared with or without the hybrid control rod. Heat removal test of the hybrid heat pipe with hydraulic CRDM system showed the heat transfer coefficient of the bundle hybrid control rod and its effect on evaporator pool. The preliminary test both hydraulic CRDM and heat removal system was conducted, which showed the possibility of the in-core hydraulic drive system for application of PINCs.

  20. Repair pathways independent of the Fanconi anemia nuclear core complex play a predominant role in mitigating formaldehyde-induced DNA damage

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Taichi [Department of Biology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Department of Dermatology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Takahashi, Akihisa [Department of Biology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Kondo, Natsuko [Particle Radiation Oncology Research Center, Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Mori, Eiichiro [Department of Biology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Okamoto, Noritomo [Department of Otorhinolaryngology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Nakagawa, Yosuke [Department of Oral and Maxillofacial Surgery, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Ohnishi, Ken [Department of Biology, Ibaraki Prefectual University of Health Sciences, 4669-2 Ami, Ami-mati, Inasiki-gun, Ibaraki 300-0394 (Japan); Zdzienicka, Malgorzata Z. [Department of Molecular Cell Genetics, Collegium Medicum in Bydgoszcz, Nicolaus-Copernicus-University in Torun, ul. Sklodowskiej-Curie 9, 85-094 Bydgoszcz (Poland); Thompson, Larry H. [Biosciences and Biotechnology Division, L452, Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA 94551-0808 (United States); Helleday, Thomas [Gray Institute for Radiation Oncology and Biology, University of Oxford, Old Road Campus Research Building, Off Roosevelt Drive, Oxford, OX3 7DQ (United Kingdom); Department of Genetics, Microbiology and Toxicology Stockholm University, SE-106 91 Stockholm (Sweden); Asada, Hideo [Department of Dermatology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); and others

    2011-01-07

    The role of the Fanconi anemia (FA) repair pathway for DNA damage induced by formaldehyde was examined in the work described here. The following cell types were used: mouse embryonic fibroblast cell lines FANCA{sup -/-}, FANCC{sup -/-}, FANCA{sup -/-}C{sup -/-}, FANCD2{sup -/-} and their parental cells, the Chinese hamster cell lines FANCD1 mutant (mt), FANCGmt, their revertant cells, and the corresponding wild-type (wt) cells. Cell survival rates were determined with colony formation assays after formaldehyde treatment. DNA double strand breaks (DSBs) were detected with an immunocytochemical {gamma}H2AX-staining assay. Although the sensitivity of FANCA{sup -/-}, FANCC{sup -/-} and FANCA{sup -/-}C{sup -/-} cells to formaldehyde was comparable to that of proficient cells, FANCD1mt, FANCGmt and FANCD2{sup -/-} cells were more sensitive to formaldehyde than the corresponding proficient cells. It was found that homologous recombination (HR) repair was induced by formaldehyde. In addition, {gamma}H2AX foci in FANCD1mt cells persisted for longer times than in FANCD1wt cells. These findings suggest that formaldehyde-induced DSBs are repaired by HR through the FA repair pathway which is independent of the FA nuclear core complex. -- Research highlights: {yields} We examined to clarify the repair pathways of formaldehyde-induced DNA damage. Formaldehyde induces DNA double strand breaks (DSBs). {yields} DSBs are repaired through the Fanconi anemia (FA) repair pathway. {yields} This pathway is independent of the FA nuclear core complex. {yields} We also found that homologous recombination repair was induced by formaldehyde.

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

    International Nuclear Information System (INIS)

    Kim, Hyun Koon

    1992-02-01

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

  2. A porous medium approach for the fluid structure interaction modelling of a water pressurized nuclear reactor core fuel assemblies: simulation and experimentation

    International Nuclear Information System (INIS)

    Ricciardi, G.

    2008-10-01

    The designing of a pressurized water reactor core subjected to seismic loading, is a major concern of the nuclear industry. We propose, in this PhD report, to establish the global behaviour equations of the core, in term of a porous medium. Local equations of fluid and structure are space averaged on a control volume, thus we define an equivalent fluid and an equivalent structure, of which unknowns are defined on the whole space. The non-linear fuel assemblies behaviour is modelled by a visco-elastic constitutive law. The fluid-structure coupling is accounted for by a body force, the expression of that force is based on empirical formula of fluid forces acting on a tube subject to an axial flow. The resulting equations are solved using a finite element method. A validation of the model, on three experimental device, is proposed. The first one presents two fuel assemblies subjected to axial flow. One of the two fuel assemblies is deviated from its position of equilibrium and released, while the other is at rest. The second one presents a six assemblies row, immersed in water, placed on a shaking table that can simulate seismic loading. Finally, the last one presents nine fuel assemblies network, arranged in a three by three, subject to an axial flow. The displacement of the central fuel assembly is imposed. The simulations are in agreement with the experiments, the model reproduces the influence of the flow of fluid on the dynamics and coupling of the fuel assemblies. (author)

  3. A preliminary feasibility study of passive in-core thermionic reactors for highly compact space nuclear power systems

    International Nuclear Information System (INIS)

    Parlos, A.G.; Khan, E.U.; Frymire, R.; Negron, S.; Thomas, J.K.; Peddicord, K.L.

    1991-01-01

    Results of a preliminary feasibility study on a new concept for a highly compact space reactor power systems are presented. Notwithstanding the preliminary nature of the present study, the results which include a new space reactor configuration and its associated technologies indicate promising avenues for the devleopment of highly compact space reactors. The calculations reported in this study include a neutronic design trade-off study using a two-dimensioinal neutron transport model, as well as a simplified one-dimensional thermal analysis of the reactor core. In arriving at the most desirable configuration, various options have been considered and analyzed, and their advantages/disadvantages have been compared. However, because of space limitation, only the most favorable reactor configuration is presented in this summary

  4. Contingency strategy for insufficient full core off load capability in spent fuel pool for Chinshan nuclear power station

    International Nuclear Information System (INIS)

    Huang, Pinghue

    2012-01-01

    The spent fuel pool (SFP) at Taiwan Power Company's (TUC's) Chinshan plant lost the full core off load (FCO) capability in 2010, even with the second SFP repacking project to expand the capacity as reported in 12PBNC. The TEPC had originally planned to move some spent fuel assemblies from SFP to dry storage facility, however, the dry storage project had seriously fell behind. Thus, it is required to address insufficient FCO capability, and the following contingency measures have been employed: The first step was to explore whether there was a specific regulatory requirement for FCO capability, and none were identified. Also, the industrial experiences were explored. The refueling strategy is changed from FCO to in-core shuffling. A feasibility evaluation performed indicates the Technical Specifications require: alternate method of decay heat removal, and verification of shutdown margin for each in vessel fuel movement. Specific methods have been successfully established. A safety evaluation for operation without FCO capability was performed, and no safety concerns were identified. The risk for operation without FCO capability was assessed. The previous operational experiences were identified. Moreover, such works are not expected in subsequent cycles. The new fuel vault is used to store new fuel assemblies. The criticality analysis has been performed and some new approaches are proposed to enhance the storage flexibility as reported in 17PBNC. An inter-unit transfer cask has been designed to transfer spent fuel from the SFP of one unit to the other. The FCO capability can be effectively extended for three more years with this consideration. The TPC discussed the contingency strategy with the ROCAEC in May 2006, and the ROCAEC's concurrence was attained. With the proposed strategy, Chinshan units have been operating smoothly

  5. Contingency strategy for insufficient full core off load capability in spent fuel pool for Chinshan nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Pinghue [Taiwan Power Company, Taipei (China)

    2012-03-15

    The spent fuel pool (SFP) at Taiwan Power Company's (TUC's) Chinshan plant lost the full core off load (FCO) capability in 2010, even with the second SFP repacking project to expand the capacity as reported in 12PBNC. The TEPC had originally planned to move some spent fuel assemblies from SFP to dry storage facility, however, the dry storage project had seriously fell behind. Thus, it is required to address insufficient FCO capability, and the following contingency measures have been employed: The first step was to explore whether there was a specific regulatory requirement for FCO capability, and none were identified. Also, the industrial experiences were explored. The refueling strategy is changed from FCO to in-core shuffling. A feasibility evaluation performed indicates the Technical Specifications require: alternate method of decay heat removal, and verification of shutdown margin for each in vessel fuel movement. Specific methods have been successfully established. A safety evaluation for operation without FCO capability was performed, and no safety concerns were identified. The risk for operation without FCO capability was assessed. The previous operational experiences were identified. Moreover, such works are not expected in subsequent cycles. The new fuel vault is used to store new fuel assemblies. The criticality analysis has been performed and some new approaches are proposed to enhance the storage flexibility as reported in 17PBNC. An inter-unit transfer cask has been designed to transfer spent fuel from the SFP of one unit to the other. The FCO capability can be effectively extended for three more years with this consideration. The TPC discussed the contingency strategy with the ROCAEC in May 2006, and the ROCAEC's concurrence was attained. With the proposed strategy, Chinshan units have been operating smoothly.

  6. Applying Sediment Cores and Nuclear Techniques for Pollution Assessment in the Bonny/new Calabar River Estuary, Niger Delta, Nigeria

    International Nuclear Information System (INIS)

    Omokheyeke, O.; Sikoki, F.; Abdelmourhit, L.

    2015-01-01

    Sediment cores were collected from the Bonny Estuary in August 2011 to archive pollution records over the last 80 years. Three sampling locations were selected and used for assessing the extent of pollution in the estuary. Alpha and gamma spectrometer in conjunction with the constant rate of supply (CRS) model were deployed to establish the age of the sediment and the data were validated using 137 Cs profiles and events that occurred within the region. The results of the activities of naturally occurring radionuclides ( 226 Ra, 228 Ra and 40 K) ranged between 15 ± 2 and 34 ± 3 Bq kg -1 ; 32 ± 5 and 48 ± 6 Bq kg -1 ; 264 ± 29 and 462 ± 36 Bq kg -1 , respectively. The profiles of stations 1 and 2 exhibited a relatively constant activity but that of station 3 showed significant decrease with depth which was appropriate for radiometric dating. CRS model displayed an increase in the bulk sediment accumulation over the past 80 years, with special reference to 1938, 1955, 1973 1997 and 2004. The cesium-137 profile was in agreement with the CRS established ages. The sedimentation rates obtained ranged from 0.019 to 0.034 gcm -2 y -1 . The data further revealed that minor and major environmental perturbations occurred in the early 1970s and late 1990s respectively. Core profiles of total hydrocarbon content ranged from 0.001 to 130.80 ppm and total phosphorus 0.01 to 0.499 mg/g. The observed profiles had peaks corresponding chronologically to severe environmental perturbations resulting from massive oil spills and substantial delivery of phosphorus from agricultural runoffs between early 1970s and early 2000s. It is therefore surmised that since the advent of industrial activities the estuary has been subject to perturbations and contamination from human activities which has resulted in adverse fluctuations in environment conditions. (author)

  7. Toward full MOX core design

    International Nuclear Information System (INIS)

    Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

    1999-01-01

    This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

  8. Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

    Directory of Open Access Journals (Sweden)

    Sidi Ali Kamel

    2012-01-01

    Full Text Available The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate’s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

  9. Final Report for DOE Grant DE-FG02-00ER41149 ''Nuclear Physics of Core-Collapse Supernovae''

    International Nuclear Information System (INIS)

    Yong-Zhong Qian

    2004-01-01

    During the funding period from August 15, 2000 to August 14, 2004, the main foci of my research have been implications of abundances in metal-poor stars for nucleosynthetic yields of supernovae and chemical evolution of the universe, effects of neutrino oscillations and neutrino-nucleus interactions on r-process nucleosynthesis, physical conditions in neutrino-driven winds from proto-neutron stars, neutrino driven mechanism of supernova explosion, supernova neutrino signals in terrestrial detectors, and constraints on variations of fundamental couplings and astrophysical conditions from properties of nuclear reactions. Personnel (three graduate students and a postdoctoral research associate) involved in my research are listed in section 2. Completed research projects are discussed in section 3. Publications during the funding period are listed in section 4 and oral presentations in section 5. Remarks about the budget are given in section 6

  10. ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

    Science.gov (United States)

    Walton, J. T.

    1994-01-01

    ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

  11. Optimisation of the core shroud bypass flow in the nuclear power plant Unterweser. Part 2: hardware implementation

    International Nuclear Information System (INIS)

    Ploeger, M.; Weissmueller, W.; Strobel, R.

    2001-01-01

    As result of the thermohydraulic analysis a modification of the upper core shroud former plate was identified. The goal was to machine 36 holes with a diameter of 56 mm into the plate. Among different methods of machining the EDM technique was selected. EDM means ''electric discharge machining''. This method contained some advantages. First, most of the equipment was available and the erodiation tools had to be optimised only. Secondly, EDM does not produce any metal chips which could probably reach the primary circuit. Tests were made using two types of erodiation electrodes. One type like a tube machined a circular groove into the 30 mm thick plate leaving two ligaments of solid material at the bottom of the groove. These were cut out in a second step. The EDM tool for cutting out is shown. The structure of the groove making tool looks quite similar. It differs in the kind of electrode only. Handling the EDM tools in the test facility is ALSO shown. (author)

  12. 77 FR 30435 - In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2012-05-23

    ... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [Docket No. PRM-50-105; NRC-2012-0056] In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission... of operating licenses for nuclear power plants (``NPP'') to operate NPPs with in-core thermocouples...

  13. Pressurized Hybrid Heat Pipe for Passive IN-Core Cooling System (PINCs) in Advanced Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Mo; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2016-05-15

    The representative operating limit of the thermosyphon heat pipe is flooding limit that arises from the countercurrent flow of vapor and liquid. The effect of difference between wetted perimeter and heated perimeter on the flooding limit of the thermosyphons has not been studied; despite the effect of cross-sectional area of the vapor path on the heat transfer characteristics of thermosyphons have been studied. Additionally, the hybrid heat pipe must operate at the high temperature and high pressure environment because it will be inserted to the active core to remove the decay heat. However, the previously studied heat pipes operated below the atmospheric pressure. Therefore, the effect of the unique geometry for hybrid heat pipe and operating pressure on the heat transfer characteristics including the flooding limit of hybrid heat pipe was experimentally measured. Hybrid heat pipe as a new conceptual decay heat removal device was proposed. For the development of hybrid heat pipe operating at high temperature and high pressure conditions, the pressurized hybrid heat pipe was prepared and the thermal performances including operation limits of hybrid heat pipe were experimentally measured. Followings were obtained: (1) As operating pressure of the heat pipe increases, the evaporation heat transfer coefficient increases due to heat transfer with convective pool boiling mode. (2) Non-condensable gas charged in the test section for the pressurization lowered the condensation heat transfer by impeding the vapor flow to the condenser. (3) The deviations between experimentally measured flooding limits for hybrid heat pipes and the values from correlation for annular thermosyphon were observed.

  14. DEPTH-CHARGE static and time-dependent perturbation/sensitivity system for nuclear reactor core analysis. Revision I

    International Nuclear Information System (INIS)

    White, J.R.

    1985-04-01

    This report provides the background theory, user input, and sample problems required for the efficient application of the DEPTH-CHARGE system - a code black for both static and time-dependent perturbation theory and data sensitivity analyses. The DEPTH-CHARGE system is of modular construction and has been implemented within the VENTURE-BURNER computational system at Oak Ridge National Laboratory. The DEPTH module (coupled with VENTURE) solves for the three adjoint functions of Depletion Perturbation Theory and calculates the desired time-dependent derivatives of the response with respect to the nuclide concentrations and nuclear data utilized in the reference model. The CHARGE code is a collection of utility routines for general data manipulation and input preparation and considerably extends the usefulness of the system through the automatic generation of adjoint sources, estimated perturbed responses, and relative data sensitivity coefficients. Combined, the DEPTH-CHARGE system provides, for the first time, a complete generalized first-order perturbation/sensitivity theory capability for both static and time-dependent analyses of realistic multidimensional reactor models. This current documentation incorporates minor revisions to the original DEPTH-CHARGE documentation (ORNL/CSD-78) to reflect some new capabilities within the individual codes

  15. Proteomics Core

    Data.gov (United States)

    Federal Laboratory Consortium — Proteomics Core is the central resource for mass spectrometry based proteomics within the NHLBI. The Core staff help collaborators design proteomics experiments in a...

  16. Nuclear

    International Nuclear Information System (INIS)

    2014-01-01

    This document proposes a presentation and discussion of the main notions, issues, principles, or characteristics related to nuclear energy: radioactivity (presence in the environment, explanation, measurement, periods and activities, low doses, applications), fuel cycle (front end, mining and ore concentration, refining and conversion, fuel fabrication, in the reactor, back end with reprocessing and recycling, transport), the future of the thorium-based fuel cycle (motivations, benefits and drawbacks), nuclear reactors (principles of fission reactors, reactor types, PWR reactors, BWR, heavy-water reactor, high temperature reactor of HTR, future reactors), nuclear wastes (classification, packaging and storage, legal aspects, vitrification, choice of a deep storage option, quantities and costs, foreign practices), radioactive releases of nuclear installations (main released radio-elements, radioactive releases by nuclear reactors and by La Hague plant, gaseous and liquid effluents, impact of releases, regulation), the OSPAR Convention, management and safety of nuclear activities (from control to quality insurance, to quality management and to sustainable development), national safety bodies (mission, means, organisation and activities of ASN, IRSN, HCTISN), international bodies, nuclear and medicine (applications of radioactivity, medical imagery, radiotherapy, doses in nuclear medicine, implementation, the accident in Epinal), nuclear and R and D (past R and D programmes and expenses, main actors in France and present funding, main R and D axis, international cooperation)

  17. The Next Generation Fornax Survey (NGFS). IV. Mass and Age Bimodality of Nuclear Clusters in the Fornax Core Region

    Science.gov (United States)

    Ordenes-Briceño, Yasna; Puzia, Thomas H.; Eigenthaler, Paul; Taylor, Matthew A.; Muñoz, Roberto P.; Zhang, Hongxin; Alamo-Martínez, Karla; Ribbeck, Karen X.; Grebel, Eva K.; Ángel, Simón; Côté, Patrick; Ferrarese, Laura; Hilker, Michael; Lançon, Ariane; Mieske, Steffen; Miller, Bryan W.; Rong, Yu; Sánchez-Janssen, Ruben

    2018-06-01

    We present the analysis of 61 nucleated dwarf galaxies in the central regions (≲R vir/4) of the Fornax galaxy cluster. The galaxies and their nuclei are studied as part of the Next Generation Fornax Survey using optical imaging obtained with the Dark Energy Camera mounted at Blanco/Cerro Tololo Inter-American Observatory and near-infrared data obtained with VIRCam at VISTA/ESO. We decompose the nucleated dwarfs in nucleus and spheroid, after subtracting the surface brightness profile of the spheroid component and studying the nucleus using point source photometry. In general, nuclei are consistent with colors of confirmed metal-poor globular clusters, but with significantly smaller dispersion than other confirmed compact stellar systems in Fornax. We find a bimodal nucleus mass distribution with peaks located at {log}({{ \\mathcal M }}* /{M}ȯ )≃ 5.4 and ∼6.3. These two nucleus subpopulations have different stellar population properties: the more massive nuclei are older than ∼2 Gyr and have metal-poor stellar populations (Z ≤ 0.02 Z ⊙), while the less massive nuclei are younger than ∼2 Gyr with metallicities in the range 0.02 < Z/Z ⊙ ≤ 1. We find that the nucleus mass ({{ \\mathcal M }}nuc}) versus galaxy mass ({{ \\mathcal M }}gal}) relation becomes shallower for less massive galaxies starting around 108 M ⊙, and the mass ratio {η }n={{ \\mathcal M }}nuc}/{{ \\mathcal M }}gal} shows a clear anticorrelation with {{ \\mathcal M }}gal} for the lowest masses, reaching 10%. We test current theoretical models of nuclear cluster formation and find that they cannot fully reproduce the observed trends. A likely mixture of in situ star formation and star cluster mergers seems to be acting during nucleus growth over cosmic time.

  18. Investigation into fuel pin reshuffling options in PWR in-core fuel management for enhancement of efficient use of nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Daing, Aung Tharn, E-mail: atdaing@khu.ac.kr; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

    2014-07-01

    Highlights: • This paper discusses an alternative option, fuel pin reshuffling for maximization of cycle energy production. • The prediction results of isotopic compositions of each burnt pin are verified. • The operating performance is analyzed at equilibrium core with fuel pin reshuffling. • The possibility of reuse of spent fuel pins for reduction of fresh fuel assemblies is investigated. - Abstract: An alternative way to enhance efficient use of nuclear fuel is investigated through fuel pin reshuffling options within PWR fuel assembly (FA). In modeling FA with reshuffled pins, as prerequisite, the single pin calculation method is proposed to estimate the isotopic compositions of each pin of burnt FA in the core-wide environment. Subsequently, such estimation has been verified by comparing with the neutronic performance of the reference design. Two scenarios are concerned, i.e., first scenario was targeted on the improvement of the uniform flux spatial distribution and on the enhancement of neutron economy by simply reshuffling the existing fuel pins in once-burnt fuel assemblies, and second one was focused on reduction of fresh fuel loading and discharged fuel assemblies with more economic incentives by reusing some available spent fuel pins still carrying enough reactivity that are mechanically sound ascertained. In scenario-1, the operating time was merely somewhat increased for few minutes when treating eight FAs by keeping enough safety margins. The scenario-2 was proved to reduce four fresh FAs loading without largely losing any targeted parameters from the safety aspect despite loss of 14 effective full power days for operation at reference plant full rated power.

  19. Intelligent uranium fission converter for neutron production on the periphery of the nuclear reactor core (MARIA reactor in Swierk - Poland)

    Energy Technology Data Exchange (ETDEWEB)

    Gryzinski, M.A.; Wielgosz, M. [National Centre for Nuclear Research, Andrzeja Soltana 7, 05-400 Otwock-Swierk (Poland)

    2015-07-01

    The multipurpose, high flux research reactor MARIA in Otwock - Swierk is an open-pool type, water and beryllium moderated and graphite reflected. There are two not occupied experimental H1 and H2 horizontal channels with complex of empty rooms beside them. Making use of these two channels is not in conflict with other research or commercial employing channels. They can work simultaneously, moreover commercial channels covers the cost of reactor working. Such conditions give beneficial possibility of creating epithermal neutron stand for researches in various field at the horizontal channel H2 of MARIA reactor (co-organization of research at H1 channel is additionally planned). At the front of experimental channels the neutron flux is strongly thermalized - neutrons with energies above 0.625 eV constitute only ∼2% of the total flux. This thermalized neutron flux will be used to achieve high flux of epithermal neutrons at the level of 2x10{sup 9} n cm{sup -2}s{sup -1} by uranium neutron converter (fast neutron production - conversion of reactor core thermal neutrons to fast neutrons - and then filtering, moderating and finally cutting of unwanted gamma radiation). The intelligent converter will be placed in the reactor pool, near the front of the H2 channel. It will replace one graphite block at the periphery of MARIA graphite reflector. The converter will consist of 20 fuel elements - low enriched uranium plates. A fuel plate will be a part which will measure 110 mm wide by 380 mm long and will consist of a thin layer of uranium sealed between two aluminium plates. These plates, once assembled, form the fuel element used in converter. The plates will be positioned vertically. There are several important requirements which should be taken into account at the converter design stage: -maximum efficiency of the converter for neutrons conversion, -cooling of the converter need to be integrated with the cooling circuit of the reactor pool and if needed equipped with

  20. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    International Nuclear Information System (INIS)

    Basol, Mit; Kielb, John F.; MuHooly, John F.; Smit, Kobus

    2007-01-01

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

  1. Nuclear export of human hepatitis B virus core protein and pregenomic RNA depends on the cellular NXF1-p15 machinery.

    Science.gov (United States)

    Yang, Ching-Chun; Huang, Er-Yi; Li, Hung-Cheng; Su, Pei-Yi; Shih, Chiaho

    2014-01-01

    Hepatitis B virus (HBV) core protein (HBc) can shuttle between nucleus and cytoplasm. Cytoplasm-predominant HBc is clinically associated with severe liver inflammation. Previously, we found that HBc arginine-rich domain (ARD) can associate with a host factor NXF1 (TAP) by coimmunoprecipitation. It is well known that NXF1-p15 heterodimer can serve as a major export receptor of nuclear mRNA as a ribonucleoprotein complex (RNP). In the NXF1-p15 pathway, TREX (transcription/export) complex plays an important role in coupling nuclear pre-mRNA processing with mRNA export in mammalian cells. Here, we tested the hypothesis whether HBc and HBV specific RNA can be exported via the TREX and NXF1-p15 mediated pathway. We demonstrated here that HBc can physically and specifically associate with TREX components, and the NXF1-p15 export receptor by coimmunoprecipitation. Accumulation of HBc protein in the nucleus can be induced by the interference with TREX and NXF1-p15 mediated RNA export machinery. HBV transcripts encodes a non-spliced 3.5 kb pregenomic RNA (pgRNA) which can serve as a template for reverse transcription. Cytoplasmic HBV pgRNA appeared to be reduced by siRNA treatment specific for the NXF1-p15 complex by quantitative RT-qPCR and Northern blot analyses. This result suggests that the pgRNA was also exported via the NXF1-p15 machinery. We entertain the hypothesis that HBc protein can be exported as an RNP cargo via the mRNA export pathway by hijacking the TREX and NXF1-p15 complex. In our current and previous studies, HBc is not required for pgRNA accumulation in the cytoplasm. Furthermore, HBc ARD can mediate nuclear export of a chimeric protein containing HBc ARD in a pgRNA-independent manner. Taken together, it suggests that while both pgRNA and HBc protein exports are dependent on NXF1-p15, they are using the same export machinery in a manner independent of each other.

  2. Thermal-hydraulic study of fixed bed nuclear reactor (FBNR), in FCC, BCC and pseudo-random configurations of the core through CFD method

    International Nuclear Information System (INIS)

    Luna, M.; Chavez, I.; Cajas, D.; Santos, R.

    2015-01-01

    The study of thermal-hydraulic performance of a fixed bed nuclear reactor (FBNR) core and the effect of the porosity was studied by the CFD method with 'SolidWorks' software. The representative sections of three different packed beds arrangements were analyzed: face-centered cubic (FCC), body-centered cubic (BCC), and a pseudo-random, with values of porosity of 0.28, 0.33 and 0.53 respectively. The minimum coolant flow required to avoid the phase change for each one of the configurations was determined. The results show that the heat transfer rate increases when the porosity value decreases, and consequently the minimum coolant flow in each configuration. The results of minimum coolant flow were: 728.51 kg/s for the FCC structure, 372.72 kg/s for the BCC, and 304.96 kg/s for the pseudo-random. Meanwhile, the heat transfer coefficients in each packed bed were 6480 W/m 2 *K, 3718 W/m 2 *K and 3042 W/m 2 *K respectively. Finally the pressure drop was calculated, and the results were 0.588 MPa for FCC configuration, 0.033 MPa for BCC and 0.017 MPa for the pseudo-random one. This means that with a higher porosity, the fluid can circulate easier because there are fewer obstacles to cross, so there are fewer energy losses. (authors)

  3. Equipment for the conditioning of core components in the fuel element storage pool with particular respect to the design required by the conditions for nuclear facilities in operation and the surveillance in accordance with atomic rules and regulations

    International Nuclear Information System (INIS)

    Dumpe, J.; Schwiertz, V.; Geiser, C.; Prucker, E.

    2001-01-01

    In nuclear power plants worn out and activated parts from the reactor core (core components) which are placed into the fuel element storage pool arise on a regular basis during the technical maintenance and the review. The disposal of these core components due to radiation protection aspects is only feasible within the fuel element storage pool during the operation of the NPP using techniques of the under water conditioning. Therefore, special GNS equipment is used for the conditioning, using under water conditioning equipment, such as UWS, BZ, and ZVA, a number of lifting and auxiliary equipment for mounting and dismantling purposes and the handling of the core components and the waste casks within the fuel element storage pool. These components must meet particular safety requirements with regard to their integrity and reliability. They are designed according to the requirements on nuclear components (KTA). The manipulating equipment must be partly redundant and the protection goals for nuclear accidents must be met. The Bavarian Ministry for Development and Environment tasked the TUeV Sueddeutschland with the surveillance and control. The conditioning equipment of GNS is therefore designed in co-ordination with the examiner of the Governmental Regulating Agency, in particular respect to all safety aspects. Furthermore the examiners perform reviews of the construction and the documentation during the design and construction phase. (orig.)

  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. Steam generation: fossil-fired systems: utility boilers; industrial boilers; boiler auxillaries; nuclear systems: boiling water; pressurized water; in-core fuel management; steam-cycle systems: condensate/feedwater; circulating water; water treatment

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    A survey of development in steam generation is presented. First, fossil-fired systems are described. Progress in the design of utility and industrial boilers as well as in boiler auxiliaries is traced. Improvements in coal pulverizers, burners that cut pollution and improve efficiency, fans, air heaters and economisers are noted. Nuclear systems are then described, including the BWR and PWR reactors, in-core fuel management techniques are described. Finally, steam-cycle systems for fossil-fired and nuclear power plants are reviewed. Condensate/feedwater systems, circulating water systems, cooling towers, and water treatment systems are discussed

  6. 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)

  7. 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)

  8. Heterogeneous gas core reactor

    International Nuclear Information System (INIS)

    Diaz, N.J.; Dugan, E.T.

    1983-01-01

    A heterogeneous gas core nuclear reactor is disclosed comprising a core barrel provided interiorly with an array of moderator-containing tubes and being otherwise filled with a fissile and/or fertile gaseous fuel medium. The fuel medium may be flowed through the chamber and through an external circuit in which heat is extracted. The moderator may be a fluid which is flowed through the tubes and through an external circuit in which heat is extracted. The moderator may be a solid which may be cooled by a fluid flowing within the tubes and through an external heat extraction circuit. The core barrel is surrounded by moderator/coolant material. Fissionable blanket material may be disposed inwardly or outwardly of the core barrel

  9. IMPROVEMENT OF STRATEGIC MANIPULATED FEDERAL PROPERTY THE EXAMPLE NON-CORE ASSETS OF JSC «CENTER OF NUCLEAR INDUSTRY NONCORE ASSETS» STATE CORPORATION «ROSATOM»

    OpenAIRE

    Ilya I. Rodin

    2015-01-01

    The article describes the main measures to improve the management of assets, federally-owned or private of public corporations - an inventory of the property, the recognition of non-core assets, the organization of decision-making systems, the sale of non-core assets at market value. The article provides the rationale for the creation within the large state-owned corporations specialized management companies responsible for the restructuring of non-core assets and improve management of the pr...

  10. Nuclear

    International Nuclear Information System (INIS)

    Anon.

    2000-01-01

    The first text deals with a new circular concerning the collect of the medicine radioactive wastes, containing radium. This campaign wants to incite people to let go their radioactive wastes (needles, tubes) in order to suppress any danger. The second text presents a decree of the 31 december 1999, relative to the limitations of noise and external risks resulting from the nuclear facilities exploitation: noise, atmospheric pollution, water pollution, wastes management and fire prevention. (A.L.B.)

  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. 78 FR 56174 - In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2013-09-12

    ... 52 [Docket No. PRM-50-105; NRC-2012-0056] In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission. ACTION: Petition for rulemaking; denial...-core thermocouples at different elevations and radial positions throughout the reactor core to enable...

  13. LMFBR core design analysis

    International Nuclear Information System (INIS)

    Cho, M.; Yang, J.C.; Yoh, K.C.; Suk, S.D.; Soh, D.S.; Kim, Y.M.

    1980-01-01

    The design parameters of a commercial-scale fast breeder reactor which is currently under construction by regeneration of these data is preliminary analyzed. The analysis of nuclear and thermal characteristics as well as safety features of this reactor is emphasized. And the evaluation of the initial core mentioned in the system description is carried out in the areas of its kinetics and control system, and, at the same time, the flow distribution of sodium and temperature distribution of the initial FBR core system are calculated. (KAERI INIS Section)

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  15. Transformer core

    NARCIS (Netherlands)

    Mehendale, A.; Hagedoorn, Wouter; Lötters, Joost Conrad

    2008-01-01

    A transformer core includes a stack of a plurality of planar core plates of a magnetically permeable material, which plates each consist of a first and a second sub-part that together enclose at least one opening. The sub-parts can be fitted together via contact faces that are located on either side

  16. Transformer core

    NARCIS (Netherlands)

    Mehendale, A.; Hagedoorn, Wouter; Lötters, Joost Conrad

    2010-01-01

    A transformer core includes a stack of a plurality of planar core plates of a magnetically permeable material, which plates each consist of a first and a second sub-part that together enclose at least one opening. The sub-parts can be fitted together via contact faces that are located on either side

  17. Oklo natural reactor. Study of uranium and rare earths migration on a core drilled through a reaction zone. Application to determination of the date of the nuclear reaction by measurement of fission products

    International Nuclear Information System (INIS)

    Ruffenach, J.C.

    1977-01-01

    Isotopic and chemical analysis of uranium and five rare earths: neodymium, samarium, europium, gadolinium and dysprosium were effected on fourteen samples taken in the same core drilled through a reaction zone of the Oklo uranium deposit. This study points out the general stability of uranium and fission rare earths; spatial distributions of these elements are quite analogous. Migrations have affected about 5% only of fission neodymium in the core of the reaction zone; corresponding values for samarium and gadolinium are slightly higher. These migration phenomena have carried rare earths to no more than 80 cm out of the core. By study of the europium it is shown that nuclear reactions have stayed in the ground since the time of reactions. On the other hand it is shown by analysis of the dysprosium that rare earths have not undergone an important movement. This study allow also the datation of nuclear reactions from the measurement of the quantity of fission neodymium produced. A value of 1.98x10 9 years is obtained slightly higher than the value obtained by geochronology [fr

  18. IMPROVEMENT OF STRATEGIC MANIPULATED FEDERAL PROPERTY THE EXAMPLE NON-CORE ASSETS OF JSC «CENTER OF NUCLEAR INDUSTRY NONCORE ASSETS» STATE CORPORATION «ROSATOM»

    Directory of Open Access Journals (Sweden)

    Ilya I. Rodin

    2015-01-01

    Full Text Available The article describes the main measures to improve the management of assets, federally-owned or private of public corporations - an inventory of the property, the recognition of non-core assets, the organization of decision-making systems, the sale of non-core assets at market value. The article provides the rationale for the creation within the large state-owned corporations specialized management companies responsible for the restructuring of non-core assets and improve management of the property. Also calculated the cost-effectiveness of the proposed measures on the example of the State Atomic Energy Corporation «Rosatom».

  19. Wire core reactor for NTP

    International Nuclear Information System (INIS)

    Harty, R.B.

    1991-01-01

    The development of the wire core system for Nuclear Thermal Propulsion (NTP) that took place from 1963 to 1965 is discussed. A wire core consists of a fuel wire with spacer wires. It's an annular flow core having a central control rod. There are actually four of these, with beryllium solid reflectors on both ends and all the way around. Much of the information on the concept is given in viewgraph form. Viewgraphs are presented on design details of the wire core, the engine design, engine weight vs. thrust, a technique used to fabricate the wire fuel element, and axial temperature distribution

  20. Reload core safety verification

    International Nuclear Information System (INIS)

    Svetlik, M.; Minarcin, M.

    2003-01-01

    This paper presents a brief look at the process of reload core safety evaluation and verification in Slovak Republic. It gives an overview of experimental verification of selected nuclear parameters in the course of physics testing during reactor start-up. The comparison of IAEA recommendations and testing procedures at Slovak and European nuclear power plants of similar design is included. An introduction of two level criteria for evaluation of tests represents an effort to formulate the relation between safety evaluation and measured values (Authors)

  1. Core lifter

    Energy Technology Data Exchange (ETDEWEB)

    Pavlov, N G; Edel' man, Ya A

    1981-02-15

    A core lifter is suggested which contains a housing, core-clamping elements installed in the housing depressions in the form of semirings with projections on the outer surface restricting the rotation of the semirings in the housing depressions. In order to improve the strength and reliability of the core lifter, the semirings have a variable transverse section formed from the outside by the surface of the rotation body of the inner arc of the semiring aroung the rotation axis and from the inner a cylindrical surface which is concentric to the outer arc of the semiring. The core-clamping elements made in this manner have the possibility of freely rotating in the housing depressions under their own weight and from contact with the core sample. These semirings do not have weakened sections, have sufficient strength, are inserted into the limited ring section of the housing of the core lifter without reduction in its through opening and this improve the reliability of the core lifter in operation.

  2. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  3. Heterogeneous gas core reactor

    International Nuclear Information System (INIS)

    Han, K.I.

    1977-01-01

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

  4. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  5. Investigation of EAS cores

    Directory of Open Access Journals (Sweden)

    Shaulov S.B.

    2017-01-01

    Full Text Available The development of nuclear-electromagnetic cascade models in air in the late forties have shown informational content of the study of cores of extensive air showers (EAS. These investigations were the main goal in different experiments which were carried out over many years by a variety of methods. Outcomes of such investigations obtained in the HADRON experiment using an X-ray emulsion chamber (XREC as a core detector are considered. The Ne spectrum of EAS associated with γ-ray families, spectra of γ-rays (hadrons in EAS cores and the Ne dependence of the muon number, ⟨Nμ⟩, in EAS with γ-ray families are obtained for the first time at energies of 1015–1017 eV with this method. A number of new effects were observed, namely, an abnormal scaling violation in hadron spectra which are fundamentally different from model predictions, an excess of muon number in EAS associated with γ-ray families, and the penetrating component in EAS cores. It is supposed that the abnormal behavior of γ-ray spectra and Ne dependence of the muon number are explained by the emergence of a penetrating component in the 1st PCR spectrum ‘knee’ range. Nuclear and astrophysical explanations of the origin of the penetrating component are discussed. The necessity of considering the contribution of a single close cosmic-ray source to explain the PCR spectrum in the knee range is noted.

  6. Ice Cores

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Records of past temperature, precipitation, atmospheric trace gases, and other aspects of climate and environment derived from ice cores drilled on glaciers and ice...

  7. Nuclear astrophysics

    International Nuclear Information System (INIS)

    Haxton, W.C.

    1992-01-01

    The problem of core-collapse supernovae is used to illustrate the many connections between nuclear astrophysics and the problems nuclear physicists study in terrestrial laboratories. Efforts to better understand the collapse and mantle ejection are also motivated by a variety of interdisciplinary issues in nuclear, particle, and astrophysics, including galactic chemical evolution, neutrino masses and mixing, and stellar cooling by the emission of new particles. The current status of theory and observations is summarized

  8. AGR core safety assessment methodologies

    International Nuclear Information System (INIS)

    McLachlan, N.; Reed, J.; Metcalfe, M.P.

    1996-01-01

    To demonstrate the safety of its gas-cooled graphite-moderated AGR reactors, nuclear safety assessments of the cores are based upon a methodology which demonstrates no component failures, geometrical stability of the structure and material properties bounded by a database. All AGRs continue to meet these three criteria. However, predictions of future core behaviour indicate that the safety case methodology will eventually need to be modified to deal with new phenomena. A new approach to the safety assessment of the cores is currently under development, which can take account of these factors while at the same time providing the same level of protection for the cores. This approach will be based on the functionality of the core: unhindered movement of control rods, continued adequate cooling of the fuel and the core, continued ability to charge and discharge fuel. (author). 5 figs

  9. Feasibility study of the design of homogeneously mixed thorium-uranium oxide and all-uranium fueled reactor cores for civil nuclear marine propulsion - 15082

    International Nuclear Information System (INIS)

    Alam, S.B.; Lindley, B.A.; Parks, G.T.

    2015-01-01

    In this reactor physics study, we attempt to design a civil marine reactor core that can operate over a 10 effective-full-power-years life at 333 MWth using ThUO 2 and all-UO 2 fuel. We use WIMS to develop subassembly designs and PANTHER to examine whole-core arrangements, optimizing: subassembly and core geometry; fuel enrichment; burnable and moveable poison design; and whole-core loading patterns. We compare designs with a 14% fissile loading for ThUO 2 and all-UO 2 fuel in 13*13 assemblies with ZrB 2 integral fuel burnable absorber pins for reactivity control. Taking advantage of self-shielding effects, the ThUO 2 option shows greater promise in the final burnable poison design while maintaining low, stable reactivity with minimal burnup penalty. For the final poisoning design with ZrB 2 , ThUO 2 contributes 2.5% more initial reactivity suppression, although the all-UO 2 design exhibits lower reactivity swing. All the candidate materials show greater rod worth for the ThUO 2 design. For both fuels, B 4 C has the highest reactivity worth, providing 10% higher control rod worth for ThUO 2 fuel than all-UO 2 . Finally, optimized assemblies were loaded into a 3D reactor model in PANTHER. The PANTHER results show that after 10 years, the core is on the border of criticality, confirming the fissile loading is well-designed. (authors)

  10. Core cooling systems

    International Nuclear Information System (INIS)

    Hoeppner, G.

    1980-01-01

    The reactor cooling system transports the heat liberated in the reactor core to the component - heat exchanger, steam generator or turbine - where the energy is removed. This basic task can be performed with a variety of coolants circulating in appropriately designed cooling systems. The choice of any one system is governed by principles of economics and natural policies, the design is determined by the laws of nuclear physics, thermal-hydraulics and by the requirement of reliability and public safety. PWR- and BWR- reactors today generate the bulk of nuclear energy. Their primary cooling systems are discussed under the following aspects: 1. General design, nuclear physics constraints, energy transfer, hydraulics, thermodynamics. 2. Design and performance under conditions of steady state and mild transients; control systems. 3. Design and performance under conditions of severe transients and loss of coolant accidents; safety systems. (orig./RW)

  11. A study to promote the establishment of R and D collaboration center for nuclear fundamental core technology development between Korea and China

    International Nuclear Information System (INIS)

    Kim, Hark Rho; Chang, J. H.; Song, C. H.; Choi, B. H.; Kim, Y. J.; Choi, C. O.; Lee, C. H.; Lee, Y. J.; Chun, S. Y.; Kim, H. J.

    2005-05-01

    This study constructed the infra for Joint R and D Collaboration Center through activating the working group activities for PRC-ROK Joint Committee on Nuclear Energy Group of Scientific Research and Development

  12. Distribution of equilibrium burnup for an homogeneous core with fuel elements of slightly enriched uranium (0.85% U-235) at Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Sidelnik, J.I.; Perez, R.A.; Salom, G.F.

    1987-01-01

    At Atucha I, the present fuel management with natural uranium comprises three burnup areas and one irradiation path, sometimes performing four steps in the reactor core, according to the requirements. The discharge burnup is 6.0 Mw d/kg U for a waste reactivity of 6.5 m k and a heavy water purity of 99.75%. This is a preliminary study to obtain the distribution of equilibrium burnup of an homogeneous core with slightly enriched uranium (0.85% by weight U-235), using the time-averaged method implemented in the code PUMA and a representative model of one third of core and fixed rod position. It was found a strategy of three areas and two paths that agrees with the present limits of channel power and specific power in fuel rod. The discharge burnup obtained is 11.6 Mw d/kg U. This strategy is calculated with the same method and a full core representation model is used to verify the obtained results. (Author)

  13. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  14. Understanding Core-Collapse Supernovae

    Science.gov (United States)

    Hix, W. R.; Lentz, E. J.; Baird, M.; Messer, O. E. B.; Mezzacappa, A.; Lee, C.-T.; Bruenn, S. W.; Blondin, J. M.; Marronetti, P.

    2010-03-01

    Our understanding of core-collapse supernovae continues to improve as better microphysics is included in increasingly realistic neutrino-radiationhydrodynamic simulations. Recent multi-dimensional models with spectral neutrino transport, which slowly develop successful explosions for a range of progenitors between 12 and 25 solar mass, have motivated changes in our understanding of the neutrino reheating mechanism. In a similar fashion, improvements in nuclear physics, most notably explorations of weak interactions on nuclei and the nuclear equation of state, continue to refine our understanding of how supernovae explode. Recent progresses on both the macroscopic and microscopic effects that affect core-collapse supernovae are discussed.

  15. 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.)

  16. Detection of local sodium boiling in the nuclear boiling generator in KNK II and in the cores of SNR 300 and SNR 2

    International Nuclear Information System (INIS)

    Erhardt, J.; Hoppe, P.

    1977-03-01

    As a basis of a global detection system, the detection of local boiling in sodium cooled reactors via surveillance of the neutron flux background noise is of special importance. With the help of parameter studies it is investigated in the present report, which parts of the core of SNR 300 and SNR 2 could be monitored with such a detection system. As a comparison the detection sensibility of the planned boiling generator in KNK II is determined

  17. SCORPIO - VVER core surveillance system

    International Nuclear Information System (INIS)

    Zalesky, K.; Svarny, J.; Novak, L.; Rosol, J.; Horanes, A.

    1997-01-01

    The Halden Project has developed the core surveillance system SCORPIO which has two parallel modes of operation: the Core Follow Mode and the Predictive Mode. The main motivation behind the development of SCORPIO is to make a practical tool for reactor operators which can increase the quality and quantity of information presented on core status and dynamic behavior. This can first of all improve plant safety as undesired core conditions are detected and prevented. Secondly, more flexible and efficient plant operation is made possible. So far the system has only been implemented on western PWRs but the basic concept is applicable to a wide range of reactor including WWERs. The main differences between WWERs and typical western PWRs with respect to core surveillance requirements are outlined. The development of a WWER version of SCORPIO was initiated in cooperation with the Nuclear Research Institute at Rez and industry partners in the Czech Republic. The first system will be installed at the Dukovany NPP. (author)

  18. Reactor core structure

    International Nuclear Information System (INIS)

    Higashinakagawa, Emiko; Sato, Kanemitsu.

    1992-01-01

    Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)

  19. 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.)

  20. Emergency core cooling system

    International Nuclear Information System (INIS)

    Kato, Ken.

    1989-01-01

    In PWR type reactors, a cooling water spray portion of emergency core cooling pipelines incorporated into pipelines on high temperature side is protruded to the inside of an upper plenum. Upon rupture of primary pipelines, pressure in a pressure vessel is abruptly reduced to generate a great amount of steams in the reactor core, which are discharged at a high flow rate into the primary pipelines on high temperature side. However, since the inside of the upper plenum has a larger area and the steam flow is slow, as compared with that of the pipelines on the high temperature side, ECCS water can surely be supplied into the reactor core to promote the re-flooding of the reactor core and effectively cool the reactor. Since the nuclear reactor can effectively be cooled to enable the promotion of pressure reduction and effective supply of coolants during the period of pressure reduction upon LOCA, the capacity of the pressure accumulation vessel can be decreased. Further, the re-flooding time for the reactor is shortened to provide an effect contributing to the improvement of the safety and the reduction of the cost. (N.H.)

  1. Plasma core reactor applications

    International Nuclear Information System (INIS)

    Latham, T.S.; Rodgers, R.J.

    1976-01-01

    Analytical and experimental investigations are being conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride (UF 6 ) fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Power, in the form of thermal radiation emitted from the high-temperature nuclear fuel, is transmitted through fused-silica transparent walls to working fluids which flow in axial channels embedded in segments of the cavity walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration; each cavity is approximately 1 m in diameter by 4.35 m in length. Axial working fluid channels are located along a fraction of each cavity peripheral wall

  2. Comparison of Transport Calculation Between 2D/1D synthesis and RAPTOR-M3G at Core Barrel of Korea Standard Nuclear Plant(KSNP), OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Maeng, Young Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou [Korea Reactor Integrity Surveillance Technology, Daejeon (Korea, Republic of); Yoo, Chun Sung [Korea Atomic Energy Research Institutes, Daejeon (Korea, Republic of)

    2013-10-15

    The DORT code for 2D/1D synthesis has been actively applied to calculate the fast neutron (E>1.0MeV) fluence exposure of RPV. RAPTOR-M3G code is also applied for the comparison of 2D/1D synthesis, and it was found that 2D/1D synthesis method generally provided more conservative results than RAPTOR-M3G at both RPV and surveillance capsule locations. As a result, definitely RAPTOR-M3G for 3D calculation must apply for accurate evaluation of the integrity and ageing of RPV and internal structures. Therefore, the purpose of this paper is to compare the differences in terms of geometric aspect of KSNP model between 2D/1D synthesis and RAPTOR-M3G at core barrel area. 2D/1D synthesis method shows still higher results at the shortest distance of bypass water region. The reason is that 2D/1D synthesis method has excessive conservatism because of having just one model of R-θ and R-Z separately. Angles (5, 25, 45, 65 and 90 degrees) that RAPTOR-M3G results are higher than 2D/1D synthesis results seem to have almost regular interval. The reason can be that neutron flux to reach to barrel is affected by the nearest core definitely and all of near core areas including bypass water. RAPTOR-M3G performing 3D calculation can be applied to various reactor structures, because the code can simulate the model realistically and reasonably in geometric view points. Understanding the phenomenon that 45 degree shows downward peak, in spite of baffle corner location, remains.

  3. Comparison of Transport Calculation Between 2D/1D synthesis and RAPTOR-M3G at Core Barrel of Korea Standard Nuclear Plant(KSNP), OPR-1000

    International Nuclear Information System (INIS)

    Maeng, Young Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Chun Sung

    2013-01-01

    The DORT code for 2D/1D synthesis has been actively applied to calculate the fast neutron (E>1.0MeV) fluence exposure of RPV. RAPTOR-M3G code is also applied for the comparison of 2D/1D synthesis, and it was found that 2D/1D synthesis method generally provided more conservative results than RAPTOR-M3G at both RPV and surveillance capsule locations. As a result, definitely RAPTOR-M3G for 3D calculation must apply for accurate evaluation of the integrity and ageing of RPV and internal structures. Therefore, the purpose of this paper is to compare the differences in terms of geometric aspect of KSNP model between 2D/1D synthesis and RAPTOR-M3G at core barrel area. 2D/1D synthesis method shows still higher results at the shortest distance of bypass water region. The reason is that 2D/1D synthesis method has excessive conservatism because of having just one model of R-θ and R-Z separately. Angles (5, 25, 45, 65 and 90 degrees) that RAPTOR-M3G results are higher than 2D/1D synthesis results seem to have almost regular interval. The reason can be that neutron flux to reach to barrel is affected by the nearest core definitely and all of near core areas including bypass water. RAPTOR-M3G performing 3D calculation can be applied to various reactor structures, because the code can simulate the model realistically and reasonably in geometric view points. Understanding the phenomenon that 45 degree shows downward peak, in spite of baffle corner location, remains

  4. 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

  5. United States Nuclear Regulatory Commission research program on core debris/concrete interactions and ex-vessel fission-product release

    International Nuclear Information System (INIS)

    Burson, S.B.

    1987-01-01

    The study of core debris/concrete interaction phenomena has been a significant element of the NRC's Severe Accident Research Program for a number of years. The CORCON and VANESA codes used to predict the consequences of high-temperature debris attack on concrete and fission-product aerosol release are state-of-the-art computational tools. The major thrust of current NRC sponsored research focuses on the refinement, verification, and validation of these codes. An overview of the analytical and experimental aspects of the NRC research program is presented

  6. Top-Level Software for VVER-1000 In-core Monitoring System under Implementation of Expanded Nuclear Fuel Diversification Program in Ukraine

    International Nuclear Information System (INIS)

    Khalimonchuk, V.A.

    2015-01-01

    The paper considers the possibility and expediency of developing mathematical software for VVER-1000 ICMS in Ukraine. This mathematical software is among the most important conditions for implementation of the expanded nuclear fuel diversification program. The top-level software is to be developed based on SSTC own studies in the development of codes for power distribution recovery, which were successfully used previously for RBMK-1000 safety analysis

  7. LWR-core behaviour project

    International Nuclear Information System (INIS)

    Paratte, J.M.

    1982-07-01

    The LWR-Core behaviour project concerns the mathematical simulation of a light water reactor in normal operation (emergency situations excluded). Computational tools are assembled, i.e. programs and libraries of data. These computational tools can likewise be used in nuclear power applications, industry and control applications. The project is divided into three parts: the development and application of calculation methods for quantisation determination of LWR physics; investigation of the behaviour of nuclear fuels under radiation with special attention to higher burnup; simulation of the operating transients of nuclear power stations. (A.N.K.)

  8. Studies on WWER core diagnostics

    International Nuclear Information System (INIS)

    Lunin, G.L.; Mitin, V.I.; Bulavin, V.V.

    1987-01-01

    The reliability and safety of nuclear power plants have decisive meaning under the situation that nuclear power generation steadily increases, and among various measures aiming at ensuring the reliability and safety in the operation of nuclear power plants, the countermeasures for protecting reactor core, main process equipment and high pressure circuits from damage have the important role, and the monitoring of condition and the organization of forecast, which are carried out continuously or periodically during the operation of nuclear power stations using the diagnostic expert system specially developed for the purpose, are included in them. Such monitoring enables the early detection of mechanical damage, increase of vibration, defects caused during operation and so on in reactor cores and primary and secondary circuits, and the continuous watching of defect developments. Also boiling in a core is detected, the place of abnormality occurrence is identified, and the intensity and characteristics of boiling are determined, thus the occurrence of dangerous condition is prevented. The developments of an in-core monitoring system and noise diagnostic systems are reported. (Kako, I.)

  9. Collapsing stellar cores and supernovae

    Energy Technology Data Exchange (ETDEWEB)

    Epstein, R J [Nordisk Inst. for Teoretisk Atomfysik, Copenhagen (Denmark); Noorgaard, H [Nordisk Inst. for Teoretisk Atomfysik, Copenhagen (Denmark); Chicago Univ., IL (USA). Enrico Fermi Inst.); Bond, J R [Niels Bohr Institutet, Copenhagen (Denmark); California Inst. of Tech., Pasadena (USA). W.K. Kellogg Radiation Lab.)

    1979-05-01

    The evolution of a stellar core is studied during its final quasi-hydrostatic contraction. The core structure and the (poorly known) properties of neutron rich matter are parametrized to include most plausible cases. It is found that the density-temperature trajectory of the material in the central part of the core (the core-center) is insensitive to nearly all reasonable parameter variations. The central density at the onset of the dynamic phase of the collapse (when the core-center begins to fall away from the rest of the star) and the fraction of the emitted neutrinos which are trapped in the collapsing core-center depend quite sensitively on the properties of neutron rich matter. We estimate that the amount of energy Ecm which is imparted to the core-mantle by the neutrinos which escape from the imploded core-center can span a large range of values. For plausible choices of nuclear and model parameters Ecm can be large enough to yield a supernova event.

  10. Core BPEL

    DEFF Research Database (Denmark)

    Hallwyl, Tim; Højsgaard, Espen

    The Web Services Business Process Execution Language (WS-BPEL) is a language for expressing business process behaviour based on web services. The language is intentionally not minimal but provides a rich set of constructs, allows omission of constructs by relying on defaults, and supports language......, does not allow omissions, and does not contain ignorable elements. We do so by identifying syntactic sugar, including default values, and ignorable elements in WS-BPEL. The analysis results in a translation from the full language to the core subset. Thus, we reduce the effort needed for working...

  11. Contribution to the study of the evolution of nuclear fuel composition in PWR type reactors. Reactor cores in three and four regions of enrichment

    International Nuclear Information System (INIS)

    Teixeira, M.C.C.

    1976-01-01

    The main methods for calculations of fuel depletion are studied and some approaches to do it are mentioned; the LEOPARD Code is described and full details are given for each subroutine, flow charts are included; the method given by the code for calculation of fuel depletion is described; some imperfections from the IPR's version are listed, and corrected, for instance: the method for burn-up calculation of heavy isotopes; the results of calculations for a reference reactor based on data of the Preliminary Safety Analysis Report (PSAR) for Angra I Nuclear Plant are presented and discussed. (author) [pt

  12. Analysis of core damage frequency: Nuclear power plant Dukovany, VVER/440 V-213 Unit 1, internal events. Volume 1: Main report

    International Nuclear Information System (INIS)

    Pugila, W.J.

    1994-01-01

    This report presents the final results from the Level 1 probabilistic safety assessment (PSA) for the Dukovany VVER/440 V-213 nuclear power plant, Unit 1. Section 1.1 describes the objectives of this study. Section 1.2 discusses the approach that was used for completing the Dukovany PSA. Section 1.3 summarizes the results of the PSA. Section 1.4 provides a comparison of the results of the Dukovany PSA with the results of other PSAs for different types of reactors worldwide. Section 1.5 summarizes the conclusions of the Dukovany PSA

  13. Validation of a new version of software for monitoring the core of nuclear power plant of Laguna Verde Unit 2, at the end of Cycle 10

    International Nuclear Information System (INIS)

    Hernandez, G.; Calleros, G.; Mata, F.

    2009-10-01

    This work shows the differences observed in thermal limits established in the technical specifications of operation, among the new software, installed at the end of Cycle 10 of Unit 2 of nuclear power plant of Laguna Verde, and the old software that was installed from the beginning of the cycle. The methodology allowed to validate the new software during the coast down stage, before finishing the cycle, for what could be used as tool during the shutdown of Unit 2 at the end of Cycle 10. (Author)

  14. True 'core' splitting. Trade unions and citizens' groups in conflict over nuclear power. Die eigentliche Kernspaltung. Gewerkschaften und Buergerinitiativen im Streit um die Atomkraft

    Energy Technology Data Exchange (ETDEWEB)

    Hallerbach, J [ed.

    1978-01-01

    Massive unemployment and the fear of a barred future put at present the unions and civil initiative to the apparent alternatives; securing work places or securing life and future. How the 'atomic fight' is fought and its result can have considerable consequences for our society. This volume presents a dialogue: Firstly the situation and environment must be understood giving rise to the controversial arguments. Reports, analyses and interviews are presented on this as basic structure for the future discussion. The quality and direction of the technical progress are dealt with in the core of the discussion. Is atomic technology acceptable. Who should decide and whom does it serve. What is progress going to look like anyway.

  15. Fostering of ability to solve problems toward consensus-making. From teaching practice on the use of nuclear power as a core of energy issues

    International Nuclear Information System (INIS)

    Harada, Tadanori

    2005-01-01

    In Hiroshima, it is practicing the peace education which aimed to bring up the citizen who practices world peace. In this research, in the nuclear power generation, at the teaching materials, it did the curriculum development to bring up the problem-solving ability to have paid to the consensus building. After practicing a class for the ninth grade life, it got to actually feel ''the problem solving depend on our future''. It understood the following point from this practice. (1) It thinks that the student wants to know the truth. (2) In to devise a way of guiding a teacher, the student becomes able to develop independent learning. (3) If there is not a mistake in the way of taking a problem, it is possible to do a student and a discussion even if it is the problem which touches a sense of values. (4) The understanding of a student is promoted when learning the difference of the mechanism of the atomic bomb and the nuclear power generation. (author)

  16. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

  17. Emergency core cooling system

    International Nuclear Information System (INIS)

    Ando, Masaki.

    1987-01-01

    Purpose: To actuate an automatic pressure down system (ADS) and a low pressure emergency core cooling system (ECCS) upon water level reduction of a nuclear reactor other than loss of coolant accidents (LOCA). Constitution: ADS in a BWR type reactor is disposed for reducing the pressure in a reactor container thereby enabling coolant injection from a low pressure ECCS upon LOCA. That is, ADS has been actuated by AND signal for a reactor water level low signal and a dry well pressure high signal. In the present invention, ADS can be actuated further also by AND signal of the reactor water level low signal, the high pressure ECCS and not-operation signal of reactor isolation cooling system. In such an emergency core cooling system thus constituted, ADS operates in the same manner as usual upon LOCA and, further, ADS is operated also upon loss of feedwater accident in the reactor pressure vessel in the case where there is a necessity for actuating the low pressure ECCS, although other high pressure ECCS and reactor isolation cooling system are not operated. Accordingly, it is possible to improve the reliability upon reactor core accident and mitigate the operator burden. (Horiuchi, T.)

  18. A Calculation of Nuclear Heating by Activation Product of Structure Materials for the In-core Irradiation Hole in the HANARO

    International Nuclear Information System (INIS)

    Noh, Tae Yang; Park, B. G.; Kim, M. S.

    2016-01-01

    Only delayed gamma heating is considered in this paper. Contribution of the delayed gamma heating is expected to be negligible for the reactor power. For the neutron irradiation, however, the contribution of delayed gamma heating is not negligible issue, and it should be evaluated for safety analysis. Additionally, in the case of temperature-sensitive irradiation targets, the delayed gamma heating should be evaluated precisely. For the HANARO, the delayed gamma heating has been evaluated by modifying the library data of the calculation code or by assuming the heating to be conservative value based on prompt gamma heating. For the method of modifying the library data, however, it should be able to estimate isotopes which contribute to heat generation exactly. And furthermore, it should be concerned to determine modified emission yield of gamma-rays depending on the half-life. For the method of assuming conservative value, it is hard to determine whether the assumed heating value is enough conservative or not. In this study, a methodology for evaluation of nuclear heating by structure materials irradiated for a long time is established with the ORIGEN and MCNP codes. And this method is applied to determine the nuclear heating of the RI capsule in the IR2 irradiation hole in the HANARO. In this paper, the methodology for evaluation of heat generation by irradiated structure materials was established by using the ORIGEN and MCNP codes. From this result, the contribution by farther structures was expected to be negligible. Meanwhile, heat generation by delayed gamma-ray was calculated less than 0.03% of heat generation by prompt radiations. The result of this study indicates that there are some remaining issues for the real situation of the neutron irradiation at HANARO.

  19. Simulation an Accelerator driven Subcritical Reactor core with thorium fuel

    International Nuclear Information System (INIS)

    Shirmohammadi, L.; Pazirandeh, A.

    2011-01-01

    The main purpose of this work is simulation An Accelerator driven Subcritical core with Thorium as a new generation nuclear fuel. In this design core , A subcritical core coupled to an accelerator with proton beam (E p =1 GeV) is simulated by MCNPX code .Although the main purpose of ADS systems are transmutation and use MA (Minor Actinides) as a nuclear fuel but another use of these systems are use thorium fuel. This simulated core has two fuel assembly type : (Th-U) and (U-Pu) . Consequence , Neutronic parameters related to ADS core are calculated. It has shown that Thorium fuel is use able in this core and less nuclear waste ,Although Iran has not Thorium reserves but study on Thorium fuel cycle can open a new horizontal in use nuclear energy as a clean energy and without nuclear waste

  20. A Bayesian reliability study on motorized valves for the emergency core cooling, heat transport isolation and shutdown cooling systems at Gentilly-2 Nuclear Generating Station

    International Nuclear Information System (INIS)

    Smith, J.E.; Rennick, D.F.; Nainer, A.

    1996-01-01

    The objective of this is to examine operational data on 32 motorized valves in the emergency core cooling, shutdown cooling and heat transport isolation systems and determine if the evidence would support a reduction in testing frequency of these valves. The methodology used is to examine the data which has accumulated on motorized valve failures since Gentilly-2 first entered service, compare these data with similar data from other sources, and determine whether the evidence indicate that demand-based, wear out type failure mechanisms play a significant role in the recorded failures. The statistical data are then updated, using a Bayesian updating procedure, to obtain revised time based failure rates and demand based probabilities of failure on demand for the motorized valves. The revised failure rates and probabilities are then applied to the fault tree models for the systems of interest to determine what effects there would be, with the current test intervals and with extended test intervals, on the probability of failure of the systems. (author)