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Sample records for annular core pulse reactor

  1. Facility modernization Annular Core Research Reactor

    International Nuclear Information System (INIS)

    The Annular Core Research Reactor (ACRR) has undergone numerous modifications since its conception in response to program needs. The original reactor fuel, which was special U-ZrH TRIGA fuel designed primarily for pulsing, has been replaced with a higher pulsing capacity BeO fuel. Other advanced operating modes which use this increased capability, in addition to the pulse and steady state, have been incorporated to tailor power histories and fluences to the experiments. Various experimental facilities have been developed that range from a radiography facility to a 50 cm diameter External Fuel Ring Cavity (FREC) using 180 of the original ZrH fuel elements. Currently a digital reactor console is being produced with GA, which will give enhanced monitoring capabilities of the reactor parameters while leaving the safety-related shutdown functions with analog technology. (author)

  2. Characterization of Novel Calorimeters in the Annular Core Research Reactor

    Science.gov (United States)

    Hehr, Brian D.; Parma, Edward J.; Peters, Curtis D.; Naranjo, Gerald E.; Luker, S. Michael

    2016-02-01

    A series of pulsed irradiation experiments have been performed in the central cavity of Sandia National Laboratories' Annular Core Research Reactor (ACRR) to characterize the responses of a set of elemental calorimeter materials including Si, Zr, Sn, Ta, W, and Bi. Of particular interest was the perturbing effect of the calorimeter itself on the ambient radiation field - a potential concern in dosimetry applications. By placing the calorimeter package into a neutron-thermalizing lead/polyethylene (LP) bucket and irradiating both with and without a cadmium wrapper, it was demonstrated that prompt capture gammas generated inside the calorimeters can be a significant contributor to the measured dose in the active disc region. An MCNP model of the experimental setup was shown to replicate measured dose responses to within 10%. The internal (n,γ) contribution was found to constitute as much as 50% of the response inside the LP bucket and up to 20% inside the nominal (unmodified) cavity environment, with Ta and W exhibiting the largest enhancement due to their sizable (n,γ) cross sections. Capture reactions in non-disc components of the calorimeter were estimated to be responsible for up to a few percent of the measured response. This work was supported by the United States Department of Energy under Contract DE-AC04-94AL85000. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy.

  3. MCNP/MCNPX model of the annular core research reactor.

    Energy Technology Data Exchange (ETDEWEB)

    DePriest, Kendall Russell; Cooper, Philip J.; Parma, Edward J., Jr. (.,; .)

    2006-10-01

    Many experimenters at the Annular Core Research Reactor (ACRR) have a need to predict the neutron/gamma environment prior to testing. In some cases, the neutron/gamma environment is needed to understand the test results after the completion of an experiment. In an effort to satisfy the needs of experimenters, a model of the ACRR was developed for use with the Monte Carlo N-Particle transport codes MCNP [Br03] and MCNPX [Wa02]. The model contains adjustable safety, transient, and control rods, several of the available spectrum-modifying cavity inserts, and placeholders for experiment packages. The ACRR model was constructed such that experiment package models can be easily placed in the reactor after being developed as stand-alone units. An addition to the 'standard' model allows the FREC-II cavity to be included in the calculations. This report presents the MCNP/MCNPX model of the ACRR. Comparisons are made between the model and the reactor for various configurations. Reactivity worth curves for the various reactor configurations are presented. Examples of reactivity worth calculations for a few experiment packages are presented along with the measured reactivity worth from the reactor test of the experiment packages. Finally, calculated neutron/gamma spectra are presented.

  4. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

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

    International Nuclear Information System (INIS)

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

  6. Study of startup conditions of a pulsed annular reactor; Estudo das reacoes de partida de um reator anelar pulsado

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Mario Augusto Bezerra da

    2003-10-15

    A new concept of reactor, which combines features of pulsed and stationary reactors, was proposed so as to produce intense neutronic fluxes. Such a reactor, known as VICHFPR (Very Intense Continuous High Flux Pulsed Reactor), consists of a subcritical core with an annular geometry and pulsed by a rotating reflector which acts as a reactivity modulator as it produces a short pulse (approximately equal to 1 ms) of high intensity, guiding the region near the pulser to super-prompt critical state. This dissertation intends to analyze the startup conditions of a Pulsed Annular Reactor. The evolution of the neutron pulse intensity is analyzed when the reactivity modulator is brought upwards according to a helicoidal path from its initial position (far away from the core), when the multiplication factor has a subcritical value, up to the final position (near the core), in which a super-prompt critical state is reached. Part of the analysis is based on the variation of neutron reflection, which is a uniform function of the exit and reflection angles between the core and the modulator. It must be emphasized that this work is an approximation of the real situation. As the initial and final reactor parameters are known, a programming code in Fortran is worked out to provide the multiplication factor and the flux intensity evolution. According to the results obtained with this code, the conditions under which the modulator must be lifted up during the startup are established. Basically, these conditions are related to the analysis of the rising and the rotation velocities, the reflector saving and the initial distance between the reactor and the modulator. The Pulsed Annular Reactor startup was divided into three stages. Because of its negative reactivity in the first two stages, the neutron multiplication is not large, while the last one, having a positive reactivity, shows an intense multiplication as is usually expected when handling pulsed systems. This last stage is quite

  7. Annular core for modular high temperature gas-cooled reactor (MHTGR)

    International Nuclear Information System (INIS)

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40 % greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93 m tall. Fuel elements contain TRISO-coated microspheres of 19.8 % enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above. (author)

  8. Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

    OpenAIRE

    Kaiser Krista; Chantel Nowlen K.; Russell DePriest K.

    2016-01-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were char...

  9. Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source; Resolucao analitica da equacao de transporte de neutrons em um reator anelar com fonte pulsada rotativa

    Energy Technology Data Exchange (ETDEWEB)

    Teixeira, Paulo Cleber Mendonca

    2002-12-01

    In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) {delta} (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) {delta} (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)

  10. Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

    Science.gov (United States)

    Kaiser, Krista; Chantel Nowlen, K.; DePriest, K. Russell

    2016-02-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  11. Characterization of the Annular Core Research Reactor (ACRR Neutron Radiography System Imaging Plane

    Directory of Open Access Journals (Sweden)

    Kaiser Krista

    2016-01-01

    Full Text Available The Annular Core Research Reactor (ACRR at Sandia National Laboratories (SNL is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1 available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  12. Conceptual design study of Pebble Bed Type High Temperature Gas-cooled Reactor with annular core structure

    International Nuclear Information System (INIS)

    This report presents the Conceptual Design Study of Pebble Bed Type High Temperature Gas-cooled Reactor with Annular Core Structure. From this study, it is made clear that the thermal power of the Pebble Bed Type Reactor can be increased to 500MW through introducing the annular core structure without losing the inherent safe characteristics (in the coolant depressurization accident, the fuel temperature does not exceed the temperature where the fuel defect begins.) This thermal power is two times higher than the inherent safe Pebble Bed Type High temperature Gas-cooled Reactor (MHTGR) designed in West Germany. From this result, it is foreseen that the ratio of the plant cost to the reactor power is reduced and the economy of the plant operation is improved. The reactor performances e.g. fuel burnup and fuel temperature are maintained in same level of the MHTGR. (author)

  13. ACRR [Annular Core Research Reactor] fission product release tests: ST-1 and ST-2

    International Nuclear Information System (INIS)

    Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACER) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model. 8 refs., 8 figs., 2 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    PARMA JR.,EDWARD J.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

  16. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor

    Science.gov (United States)

    Parma, Edward J.; Naranjo, Gerald E.; Lippert, Lance L.; Vehar, David W.

    2016-02-01

    Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR) is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity "bucket" environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters. Work supported by the United States Department of Energy at Sandia National Laboratories. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned

  17. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Parma Edward J.

    2016-01-01

    Full Text Available Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity “bucket” environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters.

  18. Lessons Learned from Sandia National Laboratories' Operational Readiness Review of the Annular Core Research Reactor (ACRR)

    International Nuclear Information System (INIS)

    The Sandia ACRR (a Hazard Category 2 Nuclear Reactor Facility) was defueled in June 1997 to modify the reactor core and control system to produce medical radioisotopes for the Department of Energy (DOE) Isotope Production Program. The DOE determined that an Operational Readiness Review (ORR) was required to confirm readiness to begin operations within the revised safety basis. This paper addresses the ORR Process, lessons learned from the Sandia and DOE ORRS of the ACRR, and the use of the ORR to confirm authorization basis implementation

  19. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

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

    International Nuclear Information System (INIS)

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

  1. Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2010-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The Japanese government approved construction of the HTTR in the 1989 fiscal year budget; construction began at the Oarai Research and Development Center in March 1991 and was completed May 1996. Fuel loading began July 1, 1998, from the core periphery. The first criticality was attained with an annular core on November 10, 1998 at 14:18, followed by a series of start-up core physics tests until a fully-loaded core was developed on December 16, 1998. Criticality tests were carried out into January 1999. The first full power operation with an average core outlet temperature of 850ºC was completed on December 7, 2001, and operational licensing of the HTTR was approved on March 6, 2002. The HTTR attained high temperature operation at 950 ºC in April 19, 2004. After a series of safety demonstration tests, it will be used as the heat source in a hydrogen production system by 2015. Hot zero-power critical, rise-to-power, irradiation, and safety demonstration testing , have also been performed with the HTTR, representing additional means for computational validation efforts. Power tests were performed in steps from 0 to 30 MW, with various tests performed at each step to confirm

  2. Effect of moderator density distribution of annular flow on fuel assembly neutronic characteristics in boiling water reactor cores

    International Nuclear Information System (INIS)

    The effect of the moderator density distribution of annular flow on the fuel assembly neutronic characteristics in a boiling water nuclear reactor was investigated using the SRAC95 code system. For the investigation, a model of annular flow for fuel assembly calculation was utilized. The results of the assembly calculation with the model (Method 1) and those of the fuel assembly calculation with the uniform void fraction distribution (Method 2) were compared. It was found that Method 2 underestimates the infinite multiplication factor in the fuel assembly including the gadolinia rod (type 1 assembly). This phenomenon is explained by the fact that the capture rate in the thermal energy region in gadolinia fuel is estimated to be smaller when the liquid film of annular flow at the fuel rod surface is considered. A burnup calculation was performed under the condition of a void fraction of 65% and a volumetric fraction of the liquid film in liquid phase of 1. It is found that Method 2 underestimates the infinite multiplication factor in comparison to Method 1 in the early stage of burnup, and that Method 2 becomes to overestimate the factor after a certain degree of burnup. This is because Method 2 overestimates the depletion rate of the gadolinia. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2009-11-01

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

  5. Characteristics of Electromagnetic Pulse Coupling into Annular Apertures

    Directory of Open Access Journals (Sweden)

    Yan-Peng Sun

    2013-11-01

    Full Text Available Electromagnetic pulse (EMP coupling into the annular apertures can disturb or damage much electronic equipment. To enhance electronic system’s  capability of anti-electromagnetic interference, the finite difference time domain method (FDTD was employed to study the characteristics of electromagnetic pulse coupling into the cavity enclosures with annular apertures. The coupling characteristics of annular apertures with different shapes (rectangle, square and circle were discussed. It shows that, in the case of the same aperture area, the coupling energy of electromagnetic pulse into the circular annular aperture is smaller than that into the rectangular and the square ones. To the rectangular annular aperture, while the polarization direction of the incident electromagnetic pulse is perpendicular to the long side of the rectangular annular aperture, the coupling energy is larger when the aspect ratio of the rectangular annular aperture is larger. The coupling effect of incident pulse with short pulse width is obviously better than the one with longer pulse width. The resonance phenomenon of the coupled waveform occurs in the cavity.

  6. Limited Diffraction Maps for Pulsed Wave Annular Arrays

    DEFF Research Database (Denmark)

    Fox, Paul D.

    2002-01-01

    A procedure is provided for decomposing the linear field of flat pulsed wave annular arrays into an equivalent set of known limited diffraction Bessel beams. Each Bessel beam propagates with known characteristics, enabling good insight into the propagation of annular fields to be obtained...

  7. An Evaluation of the Annular Fuel and Bottle-Shaped Fuel Concepts for Sodium Fast Reactors

    OpenAIRE

    Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

    2010-01-01

    Two innovative fuel concepts, the internally and externally cooled annular fuel and the bottle-shaped fuel, were investigated with the goal of increasing the power density and reduce the pressure drop in the sodium-cooled fast reactor, respectively. The concepts were explored for both high- and low-conversion core configurations, and metal and oxide fuels. The annular fuel concept is best suited for low-conversion metal-fuelled cores, where it can enable a power uprate of ~20%; the magnitude ...

  8. Core-annular flow through a horizontal pipe: Hydrodynamic counterbalancing of buoyancy force on core

    NARCIS (Netherlands)

    Ooms, G.; Vuik, C.; Poesio, P.

    2007-01-01

    A theoretical investigation has been made of core-annular flow: the flow of a high-viscosity liquid core surrounded by a low-viscosity liquid annular layer through a horizontal pipe. Special attention is paid to the question of how the buoyancy force on the core, caused by a density difference betwe

  9. Developments in fabrication of annular MOX fuel pellet for Indian fast reactor

    International Nuclear Information System (INIS)

    Mechanical rotary presses along with adoption of core rod feature were inducted for fabrication of intricate annular Mixed Oxide (MOX) pellets for Prototype Fast Breeder Reactor (PFBR). In the existing tooling, bottom plungers contain core rod whereas top plungers contain a central hole for the entry of core rod during compaction. Frequent manual clean up of top plungers after few operations were required due to settling of powder in the annular hole of top plungers during compaction. Delay in cleaning can also result in breakage of tooling apart from increase in the dose to extremities of personnel. New design of tooling has been introduced to clean up the top plungers online during the operation of rotary press. It leads to increase in the productivity, reduces the spillage of valuable nuclear material and also reduces man-rem to operators significantly. The present paper describes the modification in tooling design and compaction sequence established for online cleaning of top plungers. (author)

  10. Pulsed fusion reactors

    International Nuclear Information System (INIS)

    This summer school specialized in examining specific fusion center systems. Papers on scientific feasibility are first presented: confinement of high-beta plasma, liners, plasma focus, compression and heating and the use of high power electron beams for thermonuclear reactors. As for technological feasibility, lectures were on the theta-pinch toroidal reactors, toroidal diffuse pinch, electrical engineering problems in pulsed magnetically confined reactors, neutral gas layer for heat removal, the conceptual design of a series of laser fusion power plants with ''Saturn'', implosion experiments and the problem of the targets, the high brightness lasers for plasma generation, and topping and bottoming cycles. Some problems common to pulsed reactors were examined: energy storage and transfer, thermomechanical and erosion effects in the first wall and blanket, the problems of tritium production, radiation damage and neutron activation in blankets, and the magnetic and inertial confinement

  11. Fast reactor core concepts to improve transmutation efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Fujimura, Koji; Kawashima, Katsuyuki [Hitachi Research Laboratory, Hitachi, Ltd., 7-1-1, Omika-cho, Hitachi-shi, Ibaraki, 319-1221 Japan (Japan); Itooka, Satoshi [Hitachi-GE Nuclear Energy, Ltd., 3-1-1, Saiwai-cho, Hitachi-shi, Ibaraki, 317-0073 Japan (Japan)

    2015-12-31

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

  12. KNK II third core: design report for the annular fuel elements on the central position to accommodate material test inserts NZ 402 and NZ 403

    International Nuclear Information System (INIS)

    Since August 1984 irradiation experiments with temperature controlled pressure tube probes are being performed in the central position of KNK II. This is part of a long-term experimental program for the development of irradiation resistant reactor materials, which shall also be continued in the third core. The necessary irradiation channel is provided by a special annular fuel element. The present report describes the annular fuel elements for the third core. Aspects of the subassembly design are considered on the basis of the annular element design for the second core and the standard elements of the third core. Two annular elements NZ 402 and NZ 403 (as reserve) are available. It is demonstrated that the expected loadings will allow an unperturbed operation of the annular elements on the central position of the third core

  13. Power pulse tests on CANDU type fuel elements in TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensor for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (author)

  14. Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (orig.)

  15. Biofilm Community Dynamics in Bench-Scale Annular Reactors Simulating Arrestment of Chloraminated Drinking Water Nitrification

    Science.gov (United States)

    Annular reactors (ARs) were used to study biofilm community succession and provide an ecological insight during nitrification arrestment through simultaneously increasing monochloramine (NH2Cl) and chlorine to nitrogen mass ratios, resulting in four operational periods (I to IV)....

  16. A novel reactor concept for boron neutron capture therapy: annular low-low power reactor (ALLPR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, B.; Levine, S.H. [Department of Nuclear Engineering, Pennsylvania State University, University Park, PA 16802 (United States)

    1998-07-01

    Boron Neutron Capture Therapy (BNC), originally proposed in 50's, has been getting renewed attention over the last {approx}10 years. This is in particular due to its potential for treating deep-seated brain tumors by employing epithermal neutron beams. Large (several MW) research reactors are currently used to obtain epithermal beams for BNCT, but because of cost and licensing issues it is not likely that such high-power reactors can be placed in regular medical centers. This paper describes a novel reactor concept for BNCT devised to overcome this obstacle. The design objective was to produce a beam of epithermal neutrons of sufficient intensity for BNCT at <50 kW using low enriched uranium. It is achieved by the annular reactor design, which is called Annular Low-Low Power Reactor (ALLPR). Preliminary studies using Monte Carlo simulations are summarized in this paper. The ALLPR should be relatively economical to build, and safe and easy to operate. This novel concept may increase the viability of using BNCT in medical centers worldwide. (author)

  17. Safe operation of a TRIGA reactor in the situation of LEU-HEU core conversion

    International Nuclear Information System (INIS)

    Romanian TRIGA reactor was commissioned in 1980. The location of the research institute is Pitesti, 100 Km west of Bucharest. In fact there are two independent cores sharing the same pool. There are a 14 MW Steady State Reactor (SSR), high flux, and materials testing reactor and an Annular Core Pulsing Reactor (ACPR). The SSR reactor is a forced convection reactor cooled via a primary circuit with 4 pumps and 3 heat exchangers. The ACPR is natural convection cooled by the pool water. Modifications performed concerning core configuration resulted in the following. Removal the central pin from the bundle leads to slightly temperature increase of approximately 1% for the corner and edge pins, for the same pin power density. Also, the temperature slightly decreases for the 4 pins adjacent to the water hole. This is caused by the coolant flow redistribution. But, according to preliminary neutronic computations, PPF-s are decreasing, the edge and corner temperatures changes are no more detectable. DNB are decreasing, leading to a safer operation. Fuel management of TRIGA steady state core allows to obtain the requested fluxes for experimental reasons in the safer operation conditions. We can firmly state that the present operation of the reactor and the HEU-LEU core conversion fully respect the provisions of the National Regulatory Body and the IAEA. On the other side, we have to mention the common fact that research reactors cannot sustain themselves in the financial domain. The lack of sufficient financial support leads to shortage of the maintenance programs and to reduce of activities and personnel member; this is a real danger in maintaining the actual standards of nuclear safety. During this transition period, the Romanian TRIGA reactor is used much its capability in the frame of international cooperation this facility can ensure support for various research programmes in the fields of interest

  18. Design and characterization of the annular cathode high current pulsed electron beam source for circular components

    Science.gov (United States)

    Jiang, Wei; Wang, Langping; Wang, Xiaofeng

    2016-08-01

    In order to irradiate circular components with high current pulsed electron beam (HCPEB), an annular cathode based on carbon fiber bunches was designed and fabricated. Using an acceleration voltage of 25 kV, the maximum pulsed irradiation current and energy of this annular cathode can reach 7.9 kA and 300 J, respectively. The irradiation current density distribution of the annular cathode HCPEB source measured along the circumferential direction shows that the annular cathode has good emission uniformity. In addition, four 9310 steel substrates fixed uniformly along the circumferential direction of a metal ring substrate were irradiated by this annular cathode HCPEB source. The surface and cross-section morphologies of the irradiated samples were characterized by scanning electron microscopy (SEM). SEM images of the surface reveal that crater and surface undulation have been formed, which hints that the irradiation energy of the HCPEB process is large enough for surface modification of 9310 steel. Meanwhile, SEM cross-section images exhibit that remelted layers with a thickness of about 5.4 μm have been obtained in all samples, which proves that a good practical irradiation uniformity can be achieved by this annular cathode HCPEB source.

  19. On the levitation force in horizontal core-annular flow with a large viscosity ratio and small density ratio

    NARCIS (Netherlands)

    Ooms, G.; Pourquie, M.J.B.M.; Beerens, J.C.

    2013-01-01

    A numerical study has been made of horizontal core-annular flow: the flow of a high-viscosity liquid core surrounded by a low-viscosity liquid annular layer through a horizontal pipe. Special attention is paid to the question how the buoyancy force on the core, caused by a density difference between

  20. A complete fuel development facility utilizing a dual core TRIGA reactor system

    International Nuclear Information System (INIS)

    A TRIGA Dual Core Reactor System has been chosen by the Romanian Government as the heart of a new fuel development facility which will be operated by the Romanian Institute for Nuclear Technologies. The Facility, which will be operational in 1976, is an integral part of the Romanian National Program for Power Reactor Development, with particular emphasis being placed on fuel development. The unique combination of a new 14 MW steady state TRIGA reactor, and the well-proven TRIGA Annular Core Pulsing Reactor (ACPR) in one below-ground reactor pool resulted in a substantial construction cost savings and gives the facility remarkable experimental flexibility. The inherent safety of the TRIGA fuel elements in both reactor cores means that a secondary containment building is not necessary, resulting in further construction cost savings. The 14 MW steady state reactor gives acceptably high neutron fluxes for long- term testing of various prototype fuel-cladding-coolant combinations; and the TRIGA ACPR high pulse capability allows transient testing of fuel specimens, which is so important for accurate prediction of the performance of power reactor fuel elements under postulated failure conditions. The 14 MW steady state reactor has one large and three small in-core irradiation loop positions, two large irradiation loop positions adjacent to the core face, and twenty small holes in the beryllium reflector for small capsule irradiation. The power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position approaching 3.0 x 1014 n/cm2-sec. The ACPR has one large dry central experimental cavity which can be loaded at pool level through a shielded offset loading tube; a small diameter in-core flux trap; and an in-core pneumatically-operated capsule irradiation position. A peak pulse of 15,000 MW will yield a peak fast neutron flux in the central experimental cavity of about 1.5 x 1017 n/cm2-sec. The pulse width at half maximum during a

  1. High quantum efficiency annular backside silicon photodiodes for reflectance pulse oximetry in wearable wireless body sensors

    DEFF Research Database (Denmark)

    Duun, Sune Bro; Haahr, Rasmus Grønbek; Hansen, Ole;

    2010-01-01

    The development of annular photodiodes for use in a reflectance pulse oximetry sensor is presented. Wearable and wireless body sensor systems for long-term monitoring require sensors that minimize power consumption. We have fabricated large area 2D ring-shaped silicon photodiodes optimized...

  2. Development of space reactor core heat pipes

    International Nuclear Information System (INIS)

    The Space Power Advance Reactor (SPAR) core heat pupes are being developed to transport 15 kW of power at 1400 K. A straight, 2-m-long, 15.9-mm-diam heat pipe was fabricated of low-carbon arc-cast molybdenum and filled with sodium as the working fluid. This nonconcentric, annular, screen-tube-wick pipe was tested successfully at 16.1 kW at 1310 K, at which point a boiling limit was encountered. Follow-on work has produced an as yet untested heat pipe which has its wick centered in the evaporator by spacer wires to alleviate the boiling limit problem. A dual artery wick heat pipe is being fabricated to further improve on the boiling limit and increase redundancy. Because the heat pipe must bend around the radiation shield of the SPAR reactor, a series of bending experiments was performed. Promising results were achieved by filling the pipe completely with sodium and bending at 3650 K

  3. An analytical solution of the transport theory in an annular geometry with a rotating source; Uma solucao analitica da teoria de transporte com fonte rotativa em uma geometria anelar

    Energy Technology Data Exchange (ETDEWEB)

    Teixeira, Paulo Cleber Mendonca; Narain, Rajendra [Pernambuco Univ., Recife, PE (Brazil). Dept. de Energia Nuclear]. E-mail: clebermt@yahoo.com.br

    2002-07-01

    This paper presents an analytical solution for transport equation in a ring reactor with a rotating neutron source of the type S(x){delta}(x-Vt). It is an extension of the previous study of Williams carried out with source of the type S(x){delta}(t). Rotating neutron source is produced in a new concept of pulsed annular reactor for the production of high flux. The solution is obtained by opening of the annular geometry and applying transport theory in one-group, one-dimension, using applied mathematics techniques like Laplace Transforms and Complex Variables. A general solution for flux presented for the rotating source injected in the reactor. Condition for the existence of harmonics were specified depending upon the perimeter of the annular core. The solution is studied to look for flux instability of the harmonics in annular reactor. It is observed that no instability is possible the new reactor concept.(author)

  4. Annular shape silver lined proportional counter for on-line pulsed neutron yield measurement

    Energy Technology Data Exchange (ETDEWEB)

    Dighe, P.M., E-mail: pmdighe@barc.gov.in; Das, D.

    2015-04-01

    An annular shape silver lined proportional counter is developed to measure pulsed neutron radiation. The detector has 314 mm overall length and 235 mm overall diameter. The central cavity of 150 mm diameter and 200 mm length is used for placing the neutron source. Because of annular shape the detector covers >3π solid angle of the source. The detector has all welded construction. The detector is developed in two halves for easy mounting and demounting. Each half is an independent detector. Both the halves together give single neutron pulse calibration constant of 4.5×10{sup 4} neutrons/shot count. The detector operates in proportional mode which gives enhanced working conditions in terms of dead time and operating range compared to Geiger Muller based neutron detectors.

  5. Annular shape silver lined proportional counter for on-line pulsed neutron yield measurement

    Science.gov (United States)

    Dighe, P. M.; Das, D.

    2015-04-01

    An annular shape silver lined proportional counter is developed to measure pulsed neutron radiation. The detector has 314 mm overall length and 235 mm overall diameter. The central cavity of 150 mm diameter and 200 mm length is used for placing the neutron source. Because of annular shape the detector covers >3π solid angle of the source. The detector has all welded construction. The detector is developed in two halves for easy mounting and demounting. Each half is an independent detector. Both the halves together give single neutron pulse calibration constant of 4.5×104 neutrons/shot count. The detector operates in proportional mode which gives enhanced working conditions in terms of dead time and operating range compared to Geiger Muller based neutron detectors.

  6. RETRAC, Reactor Core Accident Simulation

    International Nuclear Information System (INIS)

    1 - Description of program or function: The RETRAC code uses a set of coupled neutron point-kinetics equations and thermal-hydraulic conservation laws to simulate nuclear reactor core behaviour under transient or accident conditions. The reactor core is represented by single equivalent unit cells composed of three regions: fuel, clad, and moderator (coolant). 2 - Method of solution: At each time step, core thermal power is calculated by solving a set of six delayed neutron group kinetics equations with adjusted reactivity feedbacks. The numerical resolution is performed by using the Runge-Kutta-Gill method. The externally inserted reactivity is specified in the input data file, whereas Doppler, fuel, clad, and water temperature reactivity feedbacks are calculated by the code itself. Core cooling is treated as a homogeneous one-dimensional fluid flow through a representative unit cell composed of three successive regions: fuel, clad, and coolant. Several flow regime models are considered for both single- and two-phase states of the coolant. The conservation laws are solved by the method of characteristics coupled with an implicit finite difference scheme to ensure stability and convergence of the numerical algorithm. Validation tests of the RETRAC code were performed by using the International Atomic Energy Agency 10-MW benchmark cores, for protected transients. Further assessment studies are in progress using experimental data. 3 - Restrictions on the complexity of the problem: The RETRAC code uses steady-state thermal-hydraulic correlations. Their use is not always justified, but it seems to be quite useful in quasi-steady cases such as as loss-of-flow transients

  7. Cold neutron tomography of annular coolant flow in a double subchannel model of a boiling water reactor

    International Nuclear Information System (INIS)

    Dryout of the liquid coolant film on fuel pins at the top of boiling water reactor (BWR) cores constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is a limiting factor in the thermal power, and therefore the economy, of BWRs. Ongoing research on multiphase annular flow, specifically the liquid film thickness, is fundamental not only to nuclear reactor safety and operation but also to that of evaporators, condensers, and pipelines in a general industrial context. We have performed cold neutron tomography of adiabatic air water annular flow in a scaled up model of the subchannel geometry found in BWR fuel assemblies today. All imaging has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institut in Switzerland. Neutron tomography is shown to excel in investigating the interactions of air water two phase flows with spacer vanes of different geometry. The high resolution, high contrast measurements provide spatial distributions of the coolant on top of the surfaces of the spacer, including the vanes, and in the subchannel downstream of the spacers.

  8. Cold neutron tomography of annular coolant flow in a double subchannel model of a boiling water reactor

    Science.gov (United States)

    Kickhofel, J. L.; Zboray, R.; Damsohn, M.; Kaestner, A.; Lehmann, E. H.; Prasser, H.-M.

    2011-09-01

    Dryout of the liquid coolant film on fuel pins at the top of boiling water reactor (BWR) cores constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is a limiting factor in the thermal power, and therefore the economy, of BWRs. Ongoing research on multiphase annular flow, specifically the liquid film thickness, is fundamental not only to nuclear reactor safety and operation but also to that of evaporators, condensers, and pipelines in a general industrial context. We have performed cold neutron tomography of adiabatic air water annular flow in a scaled up model of the subchannel geometry found in BWR fuel assemblies today. All imaging has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institut in Switzerland. Neutron tomography is shown to excel in investigating the interactions of air water two phase flows with spacer vanes of different geometry. The high resolution, high contrast measurements provide spatial distributions of the coolant on top of the surfaces of the spacer, including the vanes, and in the subchannel downstream of the spacers.

  9. Advanced core monitoring technology for WWER reactors

    International Nuclear Information System (INIS)

    The Westinghouse BEACON online monitoring system has been developed to provide continuous core monitoring and operational support for pressurized water reactor using movable detectors (fission chamber) and core thermocouples. The basic BEACON core monitoring methodology is described. Traditional WWER reactors use rhodium fixed in-core detectors as the means to provide detailed core power distribution for surveillance purposes. An adapted version of the BEACON advanced core monitoring and support system is described which seems to be, due to the different demand/response requirements, the optimal solution (for routine surveillance and anomaly detection) for WWER reactors with existing fixed in-core detectors. (Z.S.) 4 refs

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

    Energy Technology Data Exchange (ETDEWEB)

    Slutz, S.A.; Gauntt, R.O.; Harms, G.A.; Latham, T.; Roman, W.; Rodgers, R.J. (Sandia National Lab, Albuquerque, NM (United States))

    1994-05-01

    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.

  11. Fuel and core testing plan for a target fueled isotope production reactor

    International Nuclear Information System (INIS)

    pins can be removed after the experiment and using Sandia's metrology lab, relative power profiles (radially and axially) can be determined. In addition to validating neutronic analyses, confirming heat transfer properties of the target/fuel pins and core will be conducted. Fuel/target pin power limits can be verified with out-of-pile (electrical heating) thermal-hydraulic experiments. This will yield data on the heat flux across the Zircaloy clad and establish safety margin and operating limits. Using Sandia's Annular Core Research Reactor (ACRR) a 4 MW TRIGA type research reactor, target/fuel pins can be driven to desired fission power levels for long durations. Post experiment inspection of the pins can be conducted in the Auxiliary Hot Cell Facility to observe changes in the mechanical properties of the LEU matrix and burn-up effects. Transient tests can also be conducted at the ACRR to observe target/fuel pin performance during accident conditions. Target/fuel pins will be placed in double experiment containment and driven by pulsing the ACRR until target/fuel failure is observed. This will allow for extrapolation of analytical work to confirm safety margins.

  12. Experimental study on large diameter drilling in hard rock annular coring

    Institute of Scientific and Technical Information of China (English)

    Yinzhu WU; Guochun YANG; Wenchen WANG

    2008-01-01

    Based on analyzing method of large diameter hard rock drilling at home and abroad, the authors proposed a set of drilling of large diameter hard rock annular coring in low energy consumption, low cost and high efficiency. The prototype of drilling tools was designed and was made. The experimental result of the prototype indicates that this plan and technology are feasible and reach the anticipated object of design. A set of drilling tools has been offered for the constructs of large diameter hard rock coring.

  13. Shut-down margin study for the next generation VVER-1000 reactor including 13 x 13 hexagonal annular assemblies

    International Nuclear Information System (INIS)

    Highlights: → Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 x 13 assemblies are calculated. → The MCNP-5 code is run for many cases with different core burn up at various core temperatures. → There is a substantial drop in SDM in the case of annular fuel for the same power level. → SDM for our proposed VVER-1000 annular pins is calculated for specific average fuel burn up values at the BOC, MOC, and EOC. - Abstract: Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 x 13 assemblies are calculated as the main aim of the present research. We have applied the MCNP-5 code for many cases with different values of core burn up at various core temperatures, and therefore their corresponding coolant densities and boric acid concentrations. There is a substantial drop in SDM in the case of annular fuel for the same power level. Specifically, SDM for our proposed VVER-1000 annular pins is calculated when the average fuel burn up values at the BOC, MOC, and EOC are 0.531, 11.5, and 43 MW-days/kg-U, respectively.

  14. From reactors to long pulse sources

    International Nuclear Information System (INIS)

    We will show, that by using an adapted instrumentation concept, the performance of a continuous source can be emulated by one switch on in long pulses for only about 10% of the total time. This 10 fold gain in neutron economy opens up the way for building reactor like sources with an order of magnitude higher flux than the present technological limits. Linac accelerator driven spallation lends itself favorably for the realization of this kind of long pulse sources, which will be complementary to short pulse spallation sources, the same way continuous reactor sources are

  15. Liner of a pulsed thermonuclear reactor

    International Nuclear Information System (INIS)

    Different flowsheets of thermonuclear reactors being developed now are presented: quasi-stationary reactors, pulsed systems, hybrid reactors, thermonuclear reactors with collapsing liners. Methods of linear acceleration and plasma confinement, effect of linear viscosity and compressibility on linear system efficiency as well as methods of plasma shape formation in liner systems are considered. The problem of liquid metal liners application, i.e. the formation of centrifugal and jet liner as well as compression dynamics of cenrtifugal liner is studied separately. The following different flowsheets of the conversion of thermal power from thermonuclear reactors into electric power are compared: 1) thermonuclear power plant with MHD generator; 2) electric power plant with ''continuous'' and ''periodic'' coolant supply; 3) electric power plant based on tokamak reactor; 4) two-circuit conversion flowsheet with a steam turbine and MHD generator. The conclusion is made that at the present time. The development of quasi-stationary and pulsed thermonuclear systems is carried out intensively, the emphasis being placed on tokamak type quasi-stationary reactors. As for pulsed systems, a certain preference is given at present to ignition systems which according to estimates have definite prospects, but liner systems are also developed

  16. Modeling of thermal hydraulics behaviour in reactor core of reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Reactor TRIGA PUSPATI (RTP) in Malaysian Nuclear Agency (Nuclear Malaysia) is the one and only research reactor in Malaysia and had been used exclusively for research and development (R and D), training for reactor operators and education purposes. The RTP is a 1 MWt pool type reactor with natural convection cooling system and pulsing capability up to 1200 MWt. It went critical on 28 June 1982 and the core configuration has been changed twelve times to date. The core is a mixed type using 20% enriched U-ZrH fuel element containing 8.5, 12 and 20wt% uranium. This paper will discuss the modeling of thermal-hydraulics behaviour in reactor core of RTP using computer code namely PARET. The results of the calculation that were carried out at RTP are modelled and temperature profiles of the thermal hydraulics data at different locations and power levels are developed. s a comparison to the thermal hydraulics calculation using PARET, an experiment were carried out at several different locations and power levels in the reactor core for temperature profile in the core to compare the result obtained from PARET. Finally, an overall analysis of the result of PARET calculation and experimental measurement were exhibited in this paper. (author)

  17. Intrinsically secure fast reactors with dense cores

    Energy Technology Data Exchange (ETDEWEB)

    Slessarev, Igor [29, Res. Tivoli, Allee des Peupliers, 13090 Aix-en-Provence (France)], E-mail: igor.slessarev@free.fr

    2007-11-15

    Secure safety, resistance to weapons material proliferation and problems of long-lived wastes remain the most important 'painful points' of nuclear power. Many innovative reactor concepts have been developed aimed at a radical enhancement of safety. The promising potential of innovative nuclear reactors allows for shifting accents in current reactor safety 'strategy' to reveal this worth. Such strategy is elaborated focusing on the priority for intrinsically secure safety features as well as on sure protection being provided by the first barrier of defence. Concerning the potential of fast reactors (i.e. sodium cooled, lead-cooled, etc.), there are no doubts that they are able to possess many favourable intrinsically secure safety features and to lay the proper foundation for a new reactor generation. However, some of their neutronic characteristics have to be radically improved. Among intrinsically secure safety properties, the following core parameters are significantly important: reactivity margin values, reactivity feed-back and coolant void effects. Ways of designing intrinsically secure safety features in fast reactors (titled hereafter as Intrinsically Secure Fast Reactors - ISFR) can be found in the frame of current reactor technologies by radical enhancement of core neutron economy and by optimization of core compositions. Simultaneously, respecting resistance to proliferation, by using non-enriched fuel feed as well as a core breeding gain close to zero, are considered as the important features (long-lived waste problems will be considered in a separate paper). This implies using the following reactor design options as well as closed fuel cycles with natural U as the reactor feed: {center_dot}Ultra-plate 'dense cores' of the ordinary (monolithic) type with negative total coolant void effects. {center_dot}Modular type cores. Multiple dense modules can be embedded in the common reflector for achieving the desired NPP total

  18. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  19. Validation of reactor core protection system

    International Nuclear Information System (INIS)

    Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

  20. Method of evaluating the reactor core performance

    International Nuclear Information System (INIS)

    Purpose: To enable exact evaluation for the core performance in a short period. Constitution: A reactor core is equally divided into 2, 4 or 8 sections considering the structure of the symmetricalness and calculation for the evaluation the core performance is carried out to at least one region of the divided core. However, the reactor core can not be said to be completely symmetrical and there is a difference more or less, because if identical type fuels are loaded the way of burning is different depending on the positions, thereby causing difference in the total heat calorie generated. Accordingly, the performance evaluation is conducted for the entire core at a predetermined time interval, the compensation value for each of the fuels is calculated based on the result of the calculation for the entire core and the corresponding result of the calculation in each of the divided cores and the compensated values are added to the calculation result for the divided cores to compensate the calculated evaluation value. This enables to shorten the calculation time and improve the calculation accuracy. (Yoshino, Y.)

  1. Cold-neutron tomography of annular flow and functional spacer performance in a model of a boiling water reactor fuel rod bundle

    International Nuclear Information System (INIS)

    Highlights: → Annular flows w/wo functional spacers are investigated by cold neutron imaging. → Liquid film thickness distribution on fuel pins and on spacer vanes is measured. → The influence of the spacers on the liquid film distributions has been quantified. → The cross-sectional averaged liquid hold-up significantly affected by the spacers. → The sapers affect the fraction of the entrained liquid hold up in the gas core. - Abstract: Dryout of the coolant liquid film at the upper part of the fuel pins of a boiling water reactor (BWR) core constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is both a safety concern and a limiting factor in the thermal power and thus for the economy of BWRs. We have investigated adiabatic, air-water annular flows in a scaled-up model of two neighboring subchannels as found in BWR fuel assemblies using cold-neutron tomography. The imaging of the double suchannel has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institute, Switzerland. Cold-neutron tomography is shown here to be an excellent tool for investigating air-water annular flows and the influence of functional spacers of different geometries on such flows. The high-resolution, high-contrast measurements provide the spatial distributions of the coolant liquid film thickness on the fuel pin surfaces as well as on the surfaces of the spacer vanes. The axial variations of the cross-section averaged liquid hold-up and its fraction in the gas core shows the effect of the spacers on the redistribution of the two phases.

  2. LOCA power pulse analysis for CANDU-6 CANFLEX-RU core

    International Nuclear Information System (INIS)

    The power pulses following a large LOCA are analyzed for CANDU-6 reactor core fuelled with CANFLEX-RU fuel. The coupled simulations for reactor physics and channel thermal-hydraulic phenomena are done using RFSP and CATHENA codes. The 55% pump suction, 35% reactor inlet header and 100% reactor outlet header breaks are selected. The highest power pulse is predicted for 100% reactor outlet header break and it is higher than that for the standard 37-element natural fuel. However, the summation of initial stored energy and transient pulse energy of hottest pin has the minimum 17% margin to the fuel break up. Therefore, it is expected that there is no fuel breakup during the LOCA for CANFLEX-RU core

  3. 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

  4. Mechanical core coupling and reactors stability

    International Nuclear Information System (INIS)

    Structural parts of nuclear reactors are complex mechanical systems, able to vibrate with a set of proper frequencies when suitably excited. Cyclical variations in the strain state of the materials, including density perturbations, are produced. This periodic changes may affect reactor reactivity. But a variation in reactivity affects reactor thermal power, thus modifying the temperature field of the abovementiones materials. If the variation in temperature fields is fast enough, thermal-mechanical coupling may produce fast variations in strain states, and this, at its turn, modifies the reactivity, and so on. This coupling between mechanical vibrations of the structure and the materials of the core, with power oscillations of the reactor, not only may not be excluded a priori, but it seems that it has been present in some stage of the incidents or accidents that happened during the development of nuclear reactor technology. The purpose of the present communication is: (a) To review and generalize some mathematical models that were proposed in order to describe thermal-mechanical coupling in nuclear reactors. (b) To discuss some conditions in which significant instabilities could arise, including large amplitude power oscillations coupled with mechanical vibrations whose amplitudes are too small to be excluded by conventional criteria of mechanical design. Enough Certain aspects of thr physical safety of nuclear power reactors, that are objected by people that opposes to the renaissance of nucleoelectric generation, are discussed in the framework of the mathematical model proposed in this paper

  5. Gas-core reactor power transient analysis.

    Science.gov (United States)

    Kascak, A. F.

    1972-01-01

    The nuclear fuel in the gas-core reactor concept is a ball of uranium plasma radiating thermal photons. The photons are met by an inflowing hydrogen stream, which is seeded with submicron size, depleted uranium particles. A 'wall-burnout' condition exists if the thermal photons can reach the cavity liner because of insufficient absorption by the hydrogen. An analysis was conducted in order to determine the time for which the maximum steady state reactor power could be exceeded without damage to the cavity liner due to burnout. Wall-burnout time as a function of the power increase above the initial steady state condition is shown in a graph.

  6. Fluidized-bed atomic layer deposition reactor for the synthesis of core-shell nanoparticles

    International Nuclear Information System (INIS)

    The design of a fluidized bed atomic layer deposition (ALD) reactor is described in detail. The reactor consists of three parts that have all been placed in one protective cabinet: precursor dosing, reactor, and residual gas treatment section. In the precursor dosing section, the chemicals needed for the ALD reaction are injected into the carrier gas using different methods for different precursors. The reactor section is designed in such a way that a homogeneous fluidized bed can be obtained with a constant, actively controlled, reactor pressure. Furthermore, no filters are required inside the reactor chamber, minimizing the risk of pressure increase due to fouling. The residual gas treatment section consists of a decomposition furnace to remove residual precursor and a particle filter and is installed to protect the pump. In order to demonstrate the performance of the reactor, SiO2 particles have been coated with TiO2 using tetrakis-dimethylamino titanium (TDMAT) and H2O as precursors. Experiments with varying pulse times show that saturated growth can be obtained with TDMAT pulse times larger than 600 s. Analysis of the powder with High-Angle Annular Dark-Field Scanning Transmission Electron Microscopy (HAADF-STEM) and energy dispersive X-ray spectroscopy confirmed that after 50 cycles, all SiO2 particles were coated with a 1.6 nm homogenous shell of TiO2

  7. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  8. In-core instrument for nuclear reactor

    International Nuclear Information System (INIS)

    This invention concerns, in particular, an improvement for in-core equipments in a nuclear reactor having sliding members. Deposition layers of particles of metal carbides and metal nitrides are formed at the sliding surface of members in the in-core eqiupments. The matrix materials constituting the members are melted under irradiation of laser beams to form a welded layer integrated with the deposition layer. In this way, since the thickness of the welded layer is remarkably thin as compared with of the substrate material, when the irradiation of the laser beams is interrupted, corrosion resistance in water at high temperature can be improved remarkably since the melted portion is quenched and no chromium carbide is deposited at the crystal boudary. Accordingly, due to excellent corrosion resistance and abrasion resistance of the welded layer relative to the in-core equipments in the reactor having sliding surfaces, sliding incapability does not occur between each of the members under crevice conditions. Accordingly, no withdrawal incapability for equipments, for example, neutron monitors should occur upon periodical inspection. (I.S.)

  9. Sodium fast reactor evaluation: Core materials

    Science.gov (United States)

    Cheon, Jin Sik; Lee, Chan Bock; Lee, Byoung Oon; Raison, J. P.; Mizuno, T.; Delage, F.; Carmack, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.

  10. Core surveillance of boiling-water reactors

    International Nuclear Information System (INIS)

    Methods suitable for a calculational procedure which determines the three-dimensional power distribution in boilingwater reactors on the basis of in-core detector readings are described. A two- dimensional equation based on diffusion theory is set up, and a method for incorporating detector readings in the solution of this equation is presented. A three-dimensional calculational method based on nodal theory is developed. Calculations are carried out using this method, and the results are compared with a three-dimensional nodal theory calculation . Finally, parameters affecting the detector readings are examined. (author)

  11. Homopolar Gun for Pulsed Spheromak Fusion Reactors II

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2004-06-14

    A homopolar gun is discussed that could produce the high currents required for pulsed spheromak fusion reactors even with unit current amplification and open field lines during injection, possible because close coupling between the gun and flux conserver reduces gun losses to acceptable levels. Example parameters are given for a gun compatible with low cost pulsed reactors and for experiments to develop the concept.

  12. Neutronics analysis of the initial core of the TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Highlights: → The TRIGA Mark II Vienna is modeled employing MCNP5. → The model is confirmed through three different experiments. → Initial critical, reactivity distribution and flux mapping experiment. - Abstract: The Atominstitute (ATI) of Vienna University of Technology (VUT) operates a TRIGA Mark II research reactor since March 1962. Its initial criticality was achieved on 7th March 1962 when 57th Fuel Element (FE) was loaded to the core. This paper describes the development of the MCNP model of the TRIGA reactor and its validation through three different experiments i.e. initial criticality, reactivity distribution and a thermal flux mapping experiment in the reactor core. All these experiments were performed on the initial core configuration. The MCNP model includes all necessary core components i.e. FE, Graphite Element GE, neutron Source Element (SE), Central IRradiation channel (CIR) etc. Outside the core, this model simulates the annular grooved graphite reflector, the thermal and thermalizing column, four beam tubes and the reactor water tank up to 100 cm in radial and +60 and -60 cm in axial direction. Each grid position at its exact location is modeled. This model employs the ENDF/B-VI data library except for the Sm-isotopes which are taken from JEFF 3.1 because ENDF/B-VI lacks samarium (Sm) cross sections. For the first experiment, the model predicts an effective multiplication factor (κeff) of 1.00183 with an estimated standard deviation 0.00031 which is very close to the experimental value 1.00114. The second experiment measures the reactivity values of four FE and one GE. In comparison to the MCNP results, the percent difference ranges from 4 to 22. The third experiment verifies the model at a local level with the radial and axial thermal flux density distribution in the core. Though the trends are similar, the MCNP model overestimates the radial thermal flux density in the core and underestimates these results at the core periphery.

  13. The effects of annular flow on dynamics of AP1000 reactor coolant pump rotor

    International Nuclear Information System (INIS)

    The feature of AP1000 RCP rotor system is that the whole rotor system is immersed in the annular flow. The rotor in annular flow induces fluctuating fluid forces, thereby causes vibration and noise, even rotor instability. The effects of annular flow on AP1000 RCP rotor system are different from that in bearings and seals and should be considered in a new approach. Based on the turbulent bulk flow theory and perturbation analysis, the rotor-flow coupled linear dynamic model is developed to predict the dynamics of AP1000 RCP immersed rotor. During the analysis, the rotor eccentricity, stator and rotor wall friction effects are emphasized. The analytic results show the rotor eccentricity induces divergence instability and significant decrease of instability speed for system with moderate or large eccentricity; however, stator and rotor wall friction effects distinctly suppress divergence instability and increase instability speed for system with small or moderate eccentricity. Finally, we can have the conclusion that the flow-structure interaction induced by annular flow has great effects on the dynamics of AP1000 RCP immersed rotor, which should be considered in rotor dynamic analysis and design of AP1000 RCP. The method and results in the paper have theoretical significance and practical importance. (author)

  14. Applications of plasma core reactors to terrestrial energy systems

    Science.gov (United States)

    Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

  15. Performance of commercial off-the-shelf microelectromechanical systems sensors in a pulsed reactor environment

    Energy Technology Data Exchange (ETDEWEB)

    Hobert, Keith Edwin [Los Alamos National Laboratory; Heger, Arlen S [Los Alamos National Laboratory; Mccready, Steven S [Los Alamos National Laboratory

    2010-07-15

    Prompted by the unexpected failure of piezoresistive sensors in both an elevated gamma-ray environment and reactor core pulse tests, we initiated radiation testing of several MEMS piezoresistive accelerometers and pressure transducers to ascertain their radiation hardness. Some commercial off-the-shelf sensors are found to be viable options for use in a high-energy pulsed reactor, but others suffer severe degradation and even catastrophic failure. Although researchers are promoting the use of MEMS devices in radiation-harsh environment, we nevertheless find assurance testing necessary.

  16. Laser anemometer measurements in an annular cascade of core turbine vanes and comparison with theory

    Science.gov (United States)

    Goldman, L. J.; Seashultz, R. G.

    1982-01-01

    Laser measurements were made in an annular cascade of stator vanes operating at an exit critical velocity ratio of 0.78. Velocity and flow angles in the blade to blade plane were obtained at every 10 percent of axial chord within the passage and at 1/2 axial chord downstream of the vanes for radial positions near the hub, mean and tip. Results are presented in both plot and tabulated form and are compared with calculations from an inviscid, quasi three dimensional computer program. The experimental measurements generally agreed well with these theoretical calculations, an indication of the usefulness of this analytic approach.

  17. Development of dual cooled annular fuel and its possibility to enhance both economy and safety of light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yanghyun; Kim, Keonsik; Park, Jeongyong; Yang, Yongsik; Kim, Hyungkyu; In, Wangkee; Song, Kunwoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-03-15

    Over the past few decades, extensive studies have been performed to improve the reliability and safety of light water reactor (LWR) fuel. In recent years, power updating of about 10% is being achieved by modifying safety analysis methodology and subsequent increase in safety margin. But departure from nucleate boiling (DNB) and loss of coolant accident (LOCA) are still two of the most important limiting factors which would restrict power updating more than 10%. Duel cooled annular fuel, cooled in both internal and external cooling channel, has advantages of considerably lower heat flux and lower fuel temperature than conventional solid fuel. While lower heat flus gives higher DNB margin for the same power retie, lower temperature reduces the stored energy of fuel. However, there are many technical issues that should be addressed before any new type of fuel can be considered for application to LWR. This paper describes the key technologies that Korea Atomic Energy Research Institute (KAERI) has developed for dual cooled annular fuel and discusses the feasibility of its application to LWR.

  18. Development of dual cooled annular fuel and its possibility to enhance both economy and safety of light water reactor

    International Nuclear Information System (INIS)

    Over the past few decades, extensive studies have been performed to improve the reliability and safety of light water reactor (LWR) fuel. In recent years, power updating of about 10% is being achieved by modifying safety analysis methodology and subsequent increase in safety margin. But departure from nucleate boiling (DNB) and loss of coolant accident (LOCA) are still two of the most important limiting factors which would restrict power updating more than 10%. Duel cooled annular fuel, cooled in both internal and external cooling channel, has advantages of considerably lower heat flux and lower fuel temperature than conventional solid fuel. While lower heat flus gives higher DNB margin for the same power retie, lower temperature reduces the stored energy of fuel. However, there are many technical issues that should be addressed before any new type of fuel can be considered for application to LWR. This paper describes the key technologies that Korea Atomic Energy Research Institute (KAERI) has developed for dual cooled annular fuel and discusses the feasibility of its application to LWR

  19. High-voltage air-core pulse transformers

    Energy Technology Data Exchange (ETDEWEB)

    Rohwein, G. J.

    1981-01-01

    General types of air core pulse transformers designed for high voltage pulse generation and energy transfer applications are discussed with special emphasis on pulse charging systems which operate up to the multi-megavolt range. The design, operation, dielectric materials, and performance are described. It is concluded that high voltage air core pulse transformers are best suited to applications outside the normal ranges of conventional magnetic core transformers. In general these include charge transfer at high power levels and fast pulse generation with comparatively low energy. When properly designed and constructed, they are capable of delivering high energy transfer efficiency and have demonstrated superior high voltage endurance. The principal disadvantage of high voltage air core transformers is that they are not generally available from commercial sources. Consequently, the potential user must become thoroughly familiar with all aspects of design, fabrication and system application before he can produce a high performance transformer system. (LCL)

  20. On the oxidation of uraninite from natural reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Cui, D.; Eriksen, T.; Eklund, U.B.

    1999-07-01

    Natural nuclear reactors provide unique evidence in helping to understand the processes that might occur over long timescales in radioactive waste disposal sites. In the presented work, the extent and kinetics of oxidation of core material from the Oklo-Bangombe natural reactors are investigated. The X-ray powder diffraction analysis shows that the uraninites core samples from the Bangombe Reactor and Oklo Reactor 2, and Oklo Reactor 13 have the same unit-cell parameters as synthetic UO{sub 2.25}. A significant amount of fourmarierite, Pb(UO{sub 2}){sub 4}O{sub 3}(OH){sub 4}.4H{sub 2}O, was identified in the core samples from two shallow reactors Bangombe and Oklo 2, but not in the deeper reactor Oklo 13. The results of U(IV)/U(IV) measurements indicate that the extent of oxidative weathering of shallow reactors (Bangombe and Oklo 2) is greater than for the deeper reactor Oklo 13. Evaporable organic compounds found in the uraninite inclusion containing bitumen at the edge of Okelobondo Reactor (400 C) and in the black shale immediately above the Bangombe Reactor (260 C) may work as a reducing buffer or/and a hydrophobic water shield to depress the oxidative dissolution of the uraninite cores.

  1. Stability of core-annular flow of power-law fluids in the presence of interfacial surfactant

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    The shear-thinning influence on the core-annular flow stability of two immiscible power-law fluids is considered by making a linear stability analysis.The flow is driven by an axial pressure gradient in a straight pipe with the interface between the two fluids occupied by an insoluble surfactant.Given the basic flow for this core-annular arrangement,the analytical solution is obtained with respect to the power-law fluid model.The linearized equations for the evolution of infinitesimal disturbances are derived and the stability problem is formulated as a generalized matrix eigenvalue problem,which is solved by using the software package Matlab based on the QZ algorithm.The shear-thinning property is found to have marked influence on the power-law fluid core-annular flow stability,which is reflected in various aspects.First,the capillary instability is magnified by the shear-thinning property,which may lead to an essential difference between power-law and Newtonian fluid flows.Especially when the interface is close to the pipe wall,the power-law fluid flow may be unstable while the Newtonian fluid flow is stable.Second,under disturbances to the interface a velocity discontinuity at the interface appears which is destabilizing to the flow.The magnitude of this velocity discontinuity is affected by the power-law index and the flow stability is influenced correspondingly.Besides,the shear-thinning property may induce new stability modes which do not appear in the Newtonian fluid flow.The flow stability shows much dependence on the interface location,the role of which was neglected in most previous studies.The shear-thinning fluid flow is more unstable to long wave disturbances when the interface is close to the pipe wall,while the Newtonian fluid flow is more unstable when the interface is close to the pipe centerline.But this trend is changed by the addition of interfacial surfactant,for which the power-law fluid flow is more stable no matter where the interface is

  2. Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

    Energy Technology Data Exchange (ETDEWEB)

    Philbin, J.S.; Nelson, W.E.; Rosenstroch, B.

    1978-10-01

    The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO/sub 2/ fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject.

  3. Hanging core support system for a nuclear reactor. [LMFBR

    Science.gov (United States)

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-04-26

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

  4. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  5. State space modeling of reactor core in a pressurized water reactor

    Science.gov (United States)

    Ashaari, A.; Ahmad, T.; Shamsuddin, Mustaffa; M, Wan Munirah W.; Abdullah, M. Adib

    2014-07-01

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  6. 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)

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

    International Nuclear Information System (INIS)

    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)

  8. Development of inherent core technologies for advanced reactor

    International Nuclear Information System (INIS)

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  9. Development of inherent core technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keung Koo; Noh, J.M.; Hwang, D.H. [and others

    1999-03-01

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  10. Influence of air flow parameters on nanosecond repetitively pulsed discharges in a pin-annular electrode configuration

    KAUST Repository

    Heitz, Sylvain A

    2016-03-16

    The effect of various air flow parameters on the plasma regimes of nanosecond repetitively pulsed (NRP) discharges is investigated at atmospheric pressure. The two electrodes are in a pin-annular configuration, transverse to the mean flow. The voltage pulses have amplitudes up to 15 kV, a duration of 10 ns and a repetition frequency ranging from 15 to 30 kHz. The NRP corona to NRP spark (C-S) regime transition and the NRP spark to NRP corona (S-C) regime transition are investigated for different steady and harmonically oscillating flows. First, the strong effect of a transverse flow on the C-S and S-C transitions, as reported in previous studies, is verified. Second, it is shown that the azimuthal flow imparted by a swirler does not affect the regime transition voltages. Finally, the influence of low frequency harmonic oscillations of the air flow, generated by a loudspeaker, is studied. A strong effect of frequency and amplitude of the incoming flow modulation on the NRP plasma regime is observed. Results are interpreted based on the cumulative effect of the NRP discharges and an analysis of the residence times of fluid particles in the inter-electrode region. © 2016 IOP Publishing Ltd.

  11. High-voltage isolation transformer for sub-nanosecond rise time pulses constructed with annular parallel-strip transmission lines.

    Science.gov (United States)

    Homma, Akira

    2011-07-01

    A novel annular parallel-strip transmission line was devised to construct high-voltage high-speed pulse isolation transformers. The transmission lines can easily realize stable high-voltage operation and good impedance matching between primary and secondary circuits. The time constant for the step response of the transformer was calculated by introducing a simple low-frequency equivalent circuit model. Results show that the relation between the time constant and low-cut-off frequency of the transformer conforms to the theory of the general first-order linear time-invariant system. Results also show that the test transformer composed of the new transmission lines can transmit about 600 ps rise time pulses across the dc potential difference of more than 150 kV with insertion loss of -2.5 dB. The measured effective time constant of 12 ns agreed exactly with the theoretically predicted value. For practical applications involving the delivery of synchronized trigger signals to a dc high-voltage electron gun station, the transformer described in this paper exhibited advantages over methods using fiber optic cables for the signal transfer system. This transformer has no jitter or breakdown problems that invariably occur in active circuit components.

  12. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    Science.gov (United States)

    Clement, J. D.; Rust, J. H.

    1977-01-01

    Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

  13. Method of controlling the heterogeneous reactor core in FBR type reactors

    International Nuclear Information System (INIS)

    Purpose: To maintain the power distribution of fuel assemblies constant all over the reactor operation period by operating the control rods depending on the power change in blanket fuels. Method: Blanket fuels (internal blanket) are loaded at a central region of a reactor core comprising plutonium enriched region. Further, control rods for the start-up and shutdown of a reactor and fuel compensation and back-up control rods are arranged within the reactor core. The reactor core is surrounded with an axial blanket and a neutron shielding body. 21 fuel compensating control rods are present in the reactor core and 18 rods out of them are arranged at the outer region of the inner blanket. At the initial stage of the reactor operation, the control rods are divided into three blocks and they are inserted into the reactor core by 0%, 21% and 20% respectively required for the compensation of the burning reactivity at the initial stage of the reactor operation and inserted by 2%, 18% and 15% respectively at the initial balanced stage of the reactor core. (Horiuchi, T.)

  14. Assumed mode approach to fast reactor core seismic analysis

    International Nuclear Information System (INIS)

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

  15. Power instability and stochastic dynamics of periodic pulsed reactors

    International Nuclear Information System (INIS)

    This paper reports that physicists dealing with conventional reactor dynamics recognize two types of instability and reactor behavior beyond the stability region: asymptotic excursions and nonlinear periodic oscillations. A periodically pulsed reactor (PPR) has another peculiar instability: Under certain conditions, its power tends to oscillate at a frequency just twice less than the reactor pulsation frequency. The PPR dynamics far beyond the stability region are analyzed by using a discrete nonlinear model. A PPR with a negative temperature reactivity effect inevitably shows the chaotic power pulse energy behavior known as deterministic chaos. The way by which a reactor goes to chaos is defined by the time dependence of the feedback and by the kind of dynamics model used

  16. 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).

  17. Optimization of the core of a 600 MV HTGR reactor

    International Nuclear Information System (INIS)

    Through a thermal analysis, several reactor core parameters are considered, viz.: cooling channel diameter, juel channel diameter, distance between two channels power generated for lenght unit, etc. Using several criteria, the best solution or solutions are chosen

  18. Development of core design and analyses technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

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

    International Nuclear Information System (INIS)

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

  20. Core Physics of Pebble Bed High Temperature Nuclear Reactors

    NARCIS (Netherlands)

    Auwerda, G.J.

    2014-01-01

    To more accurately predict the temperature distribution inside the reactor core of pebble bed type high temperature reactors, in this thesis we investigated the stochastic properties of randomly stacked beds and the effects of the non-homogeneity of these beds on the neutronics and thermal-hydraulic

  1. Hollow-core fibers for high power pulse delivery

    DEFF Research Database (Denmark)

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

    2016-01-01

    picosecond pulses. A novel fiber with 7 tubes and a core of 30 mu m was fabricated and it is here described and characterized, showing remarkable low loss, low bend loss, and good mode quality. Its optical properties are compared to both a 10 mu m and a 18 mu m core diameter photonic band gap hollow......-core fiber. The three fibers are characterized experimentally for the delivery of 22 picosecond pulses at 1032nm. We demonstrate flexible, diffraction limited beam delivery with output average powers in excess of 70W. (C) 2016 Optical Society of America...

  2. Review of neutronic assessments applied to small reactor core physics

    International Nuclear Information System (INIS)

    In its design division for material test reactors and research reactors, AREVA TA has to characterize these manufactured cores. This step is sequential with neutronics benchmarks associated with validation (standard Verification and Validation approach). The previous two points are embedded in core projects and can be run separately especially when experimental tests are foreseen for validation database enrichment. Methodological standard is given in order to match validation and benchmark process illustrated alongside with two specific items on critical research reactors (AZUR - JHR) and subcritical mock up (AZUR). (author)

  3. Neutron spectrometric methods for core inventory verification in research reactors

    CERN Document Server

    Ellinger, A; Hansen, W; Knorr, J; Schneider, R

    2002-01-01

    In consequence of the Non-Proliferation Treaty safeguards, inspections are periodically made in nuclear facilities by the IAEA and the EURATOM Safeguards Directorate. The inspection methods are permanently improved. Therefore, the Core Inventory Verification method is being developed as an indirect method for the verification of the core inventory and to check the declared operation of research reactors.

  4. Core neutronics of a swimming pool research reactor

    International Nuclear Information System (INIS)

    The initial cores of the 5 MW swimming pool research reactor of the Nuclear Research Centre, Tehran have been analyzed using the computer codes METHUSELAH and EQUIPOISE. The effective multiplication factor, critical mass, moderator temperature and void coefficients of the core have been calculated and compared with vendor's values. Calculated values agree reasonably well with the vendor's results. (author)

  5. Monte Carlo modelling of VR-1 reactor core

    International Nuclear Information System (INIS)

    The possibilities of reactor core analysis by precise Monte Carlo codes are gradually increasing along with the accessibility of computing power. In the case of zero power research reactors, where temperature and burn-up effects remain negligible, model can approximate the reality to a very high degree. In such a case, most of calculation uncertainty can be caused by uncertainties in technical specifications of fuel and reactor internals. Thus performance of the modelling and its predictive power can be significantly improved via comparison with a large set of experimental data that can be acquired during reactor operation and via subtle tuning and improving the calculation model. The paper describes the case for neutronics calculations of VR-1 zero power reactor core. (author)

  6. Characteristics of fast reactor core designs and closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N. [State Scientific Center of the Russian Federation, Institute for Physics and Power Engineering (IPPE), Obninsk, Kaluga Region (Russian Federation)

    2007-07-01

    On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. COMSORS: A light water reactor chemical core catcher

    International Nuclear Information System (INIS)

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B2O3) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation

  10. Annular pancreas

    Science.gov (United States)

    ... page: //medlineplus.gov/ency/article/001142.htm Annular pancreas To use the sharing features on this page, please enable JavaScript. An annular pancreas is a ring of pancreatic tissue that encircles ...

  11. Radiation Characterization Summary: ACRR Central Cavity Free-Field Environment with the 32-Inch Pedestal at the Core Centerline (ACRR-FF-CC-32-cl).

    Energy Technology Data Exchange (ETDEWEB)

    Vega, Richard Manuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Parma, Edward J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Naranjo, Gerald E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lippert, Lance L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vehar, David W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Griffin, Patrick J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    This document presents the facilit y - recommended characteri zation o f the neutron, prompt gamma - ray, and delayed gamma - ray radiation fields in the Annular Core Research Reactor ( ACRR ) for the cen tral cavity free - field environment with the 32 - inch pedestal at the core centerline. The designation for this environmen t is ACRR - FF - CC - 32 - cl. The neutron, prompt gamma - ray , and delayed gamma - ray energy spectra , uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma - ray fluence profiles within the experiment area of the cavity . Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples . Acknowledgements The authors wish to th ank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work . Also thanks to David Ames for his assistance in running MCNP on the Sandia parallel machines.

  12. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  13. Advanced eddy current technique for measurement of annular gap between pressure tube and calandria tube in Indian Pressurized Heavy Water Reactors (PHWRs)

    International Nuclear Information System (INIS)

    In Indian Pressurised Heavy Water Reactors (PHWRs), the PT (pressure tube) is designed to be nominally concentric with the encircling CT (calandria tube). Due to various factors PT becomes eccentric with respect to CT over the life of reactor. If this becomes excessive, hot PT will come in contact with cold CT. Such a cold spot could act as potential location for initiating blister formation and premature failure of PT. Hence it is important to periodically measure annular gap between PT and CT. An advanced eddy current technique has been successfully developed and incorporated in BARCIS (BARC Channel Inspection System) for measurement of PT-CT gap. (author). 4 refs., 3 figs

  14. A remote maintenance robot system for a pulsed nuclear reactor

    International Nuclear Information System (INIS)

    This paper presents a remote maintenance robot system for use in a hazardous environment. The system consists of turntable, robot and hoist subsystems which operate under the control of a supervisory computer to perform coordinated programmed maintenance operations on a pulsed nuclear reactor. The system is operational

  15. Aspects of long pulse commercial tokamak reactor design

    International Nuclear Information System (INIS)

    Illustrative design parameters have been developed for an Ultra-Long Pulse Tokamak Reactor (ULTR) with a pulse length of 24 hours. The principles developed for the 24 hour pulse length ULTR design have also been used in a scoping study for machines with shorter pulses. Parameters for a machine with a 4 hour pulse length are given. The ULTR design has an ohmic transformer which is optimized to satisfy the conflicting requirements of maximized OH drive and for reaction of the inward forces of the TF coil. The toroidal field magnet is completely modularized, with each coil in its own cryostat and shearing panels at room temperature. Analysis of effects of thermal fatigue on a representative first wall design has been performed. Windows of allowed operation have been found, limited by primary stress, swelling and fatigue. The analysis has been carried out for first walls made of stainless steel or vanadium

  16. Assessment of HCDA energetics in the CRBRP heterogeneous reactor core

    International Nuclear Information System (INIS)

    The results of hypothetical core disruptive event analyses for the CRBRP heterogeneous reactor core are reported. The analytical results cover a large number of parametric cases including variations in design parameters and phenomenological assumptions. Reactor core configurations at the beginning of cycle one and end of cycle four are evaluated. The energetic consequences are evaluated based upon both fuel expansion thermodynamic work potential and a relative probability assignment. It is concluded that the structural loads, which result from 101 megajoules of available expansion work at sodium slug impact on the reactor closure head (equivalent to 661 megajoules of fuel expansion work to one atmosphere), is an adequate energetic consequence envelope for use in specifying the Structural Margins Beyond the Design Basis

  17. Assessment of HCDA energetics in the CRBRP heterogeneous reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Rhow, S K; Switick, D M; McElroy, J L; Joe, B W; Elawar, Z J

    1981-03-27

    The results of hypothetical core disruptive event analyses for the CRBRP heterogeneous reactor core are reported. The analytical results cover a large number of parametric cases including variations in design parameters and phenomenological assumptions. Reactor core configurations at the beginning of cycle one and end of cycle four are evaluated. The energetic consequences are evaluated based upon both fuel expansion thermodynamic work potential and a relative probability assignment. It is concluded that the structural loads, which result from 101 megajoules of available expansion work at sodium slug impact on the reactor closure head (equivalent to 661 megajoules of fuel expansion work to one atmosphere), is an adequate energetic consequence envelope for use in specifying the Structural Margins Beyond the Design Basis.

  18. Studying the effects of dynamical parameters on reactor core temperature

    Directory of Open Access Journals (Sweden)

    R Khodabakhsh

    2015-01-01

    Full Text Available In order to increase productivity, reduce depreciation, and avoid possible accidents in a system such as fuel rods' melting and overpressure, control of temperature changes in the reactor core is an important factor. There are several methods for solving and analysing the stability of point kinetics equations. In most previous analyses, the effects of various factors on the temperature of the reactor core have been ignored. In this work, the effects of various dynamical parameters on the temperature of the reactor core and stability of the system in the presence of temperature feedback reactivity with external reactivity step, ramp and sinusoidal for six groups of delayed neutrons were studied using the method of Lyapunov exponent. The results proved to be in good agreement with other works

  19. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

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

  20. Investigation of the core melt accident in light water reactors

    International Nuclear Information System (INIS)

    In the thesis the core melt accident, heating up and collapsing of the reactor core were investigated. The most important parameters of influence were found and their effect on the development of the accident were shown. A causal diagram was developed representing the great number of events occurring in the course of the core melt accident as well as their mutual dependences. Models were developed and applied for a detailed description of the collapse process, melting of materials, heat and material transport at flow-off of the melted mass and for taking into account steam blocking in the destroyed core sections. (orig.)

  1. Site Investigation for Detection of KIJANG Reactor Core Center

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Tae-Hyun; Kim, Jun Yeon; Kim, Jeeyoung [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    It was planned for the end of March 2017 and extended to April 2018 according to the government budget adjustment. The KJRR project is intended for filling the self-sufficiency of RI demand including Mo-99, increasing the NTD capacity and developing technologies related to the research reactor. In project, site investigation is the first activity that defines seismologic and related geologic aspects of the site. Site investigation was carried out from Oct. 2012 to Jan. 2014 and this study is intended to describe detail procedures in locating the reactor core center. The location of the reactor core center was determined by collectively reviewing not only geological information but also information from architects engineering. EL 50m was selected as ground level by levering construction cost. Four recommended locations (R-1a - R-1d) are displayed for the reactor core center. R-1a was found optimal in consideration of medium rock contour, portion of medium rock covering reactor buildings, construction cost, physical protection and electrical resistivity. It is noted that engineering properties of the medium rock is TCR/RQD 100/53, elastic modulus 7,710 - 8,720MPa, permeability coefficient 2.92E-06cm/s, and S-wave velocity 1,380m/s, sound for foundations of reactor buildings.

  2. Design and development of small and medium integral reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Chang, M. H.; Lee, C. C.; Song, J. S.; Cho, B. O.; Kim, K. Y.; Kim, S. J.; Park, S. Y.; Lee, K. B.; Lee, C. H.; Chun, T. H.; Oh, D. S.; In, W. K.; Kim, H. K.; Lee, C. B.; Kang, H. S.; Song, K. N.

    1997-07-01

    Recently, the role of small and medium size integral reactors is remarkable in the heat applications rather than the electrical generations. Such a range of possible applications requires extensive used of inherent safety features and passive safety systems. It also requires ultra-longer cycle operations for better plant economy. Innovative and evolutionary designs such as boron-free operations and related reactor control methods that are necessary for simple reactor system design are demanded for the small and medium reactor (SMR) design, which are harder for engineers to implement in the current large size nuclear power plants. The goals of this study are to establish preliminary design criteria, to perform the preliminary conceptual design and to develop core specific technology for the core design and analysis for System-integrated Modular Advanced ReacTor (SMART) of 330 MWt power. Based on the design criteria of the commercial PWR`s, preliminary design criteria will be set up. Preliminary core design concept is going to be developed for the ultra-longer cycle and boron-free operation and core analysis code system is constructed for SMART. (author). 100 refs., 40 tabs., 92 figs.

  3. Uranium droplet nuclear reactor core with MHD generator

    Science.gov (United States)

    Anghaie, Samim; Kumar, Ratan

    An innovative concept employing liquid uranium droplets as fuel in an ultrahigh-temperature vapor core reactor (UTVR) magnetohydrodynamic (MHD) generator power system for space power generation has been studied. Metallic vapor in superheated form acts as a working fluid for a closed-Rankine-type thermodynamic cycle. Usage of fuel and working fluid in this form assures certain advantages. The major technical issues emerging as a result involve a method for droplet generation, droplet transport in the reactor core, heat generation in the fuel and transport to the metallic vapor, and materials compatibility. A qualitative and quantitative attempt to resolve these issues has indicated the promise and tentative feasibility of the system.

  4. Vibration tests on some models of PEC reactor core elements

    International Nuclear Information System (INIS)

    This paper describes the aims of the experimental tests carried out at ISMES, within an agreement with the Department of Fast Reactors of ENEA, on some models of the elements of PEC Fast Nuclear Reactor Core in the frame of the activities for the seismic verification of the PEC core. The seismic verification is briefly described with particular attention to the problems arising from the shocks among the various elements during an earthquake, as well as the computer code used, the purpose and the techniques used to perform tests, some results and the first comparison between the theory and the experimental data

  5. Modification of the Core Cooling System of TRIGA 2000 Reactor

    Science.gov (United States)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  6. Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

    Science.gov (United States)

    Butler, C.; Albright, D.

    2007-01-01

    Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

  7. 78 FR 56174 - In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2013-09-12

    ... the Federal Register (FR) on May 23, 2012 (77 FR 30435). The petitioner requested that the NRC amend...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Parts 50 and 52 In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission. ACTION: Petition for...

  8. 77 FR 30435 - In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2012-05-23

    ... assigned Docket ID PRM-50-84 (73 FR 71564; November 25, 2008). In addition, the petition states that the... COMMISSION 10 CFR Part 50 In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core... ``require all holders of operating licenses for nuclear power plants (``NPP'') to operate NPPs with...

  9. Influence of nuclear data covariance on reactor core calculations

    International Nuclear Information System (INIS)

    The influence of nuclear data uncertainties on reactor core calculations were investigated systematically using the sampling based uncertainty and sensitivity software XSUSA developed at GRS. Varied nuclear data are generated randomly corresponding to the uncertainty information from the covariance matrices. After performing a large number of calculations with these data, the results are statistically evaluated; this can be done not only for integral, but also for local output quantities like the assembly power distribution of a reactor core. The method is applied to multi-group Monte Carlo calculations stationary states of the PWR MOX/UO2 core transient benchmark, and to corresponding nodal diffusion calculations. Unexpectedly large uncertainties result for the radial power distribution. The uncertainties in the nodal results agree very well with those in the Monte Carlo reference results; thus, it is possible to apply the random sampling method to determine the influence of nuclear data uncertainties on transient core calculations. (author)

  10. Validation of eureka-2/rr code for analysis of pulsing parameters of triga mark ii research reactor in bangladesh

    International Nuclear Information System (INIS)

    Some parametric studies on pulsing mode for fresh core of TRIGA Mark II research reactor in AERE, Savar, have been carried out with coupled thermal-hydraulics code EUREKA-2/RR in association with neutronics code SRAC. At the beginning, role of some important parameters in pulsing like delayed neutron fraction (beta eff) and reactivity insertion have been studied keeping prompt neutron life time (lp) fixed at 33.4 micro-sec. After a series of experiments, we found that the pulsing peak that is consistent with the Safety Analysis Report (SAR) is for the delayed neutron fraction (beta eff) of 0.007 and reactivity insertion of 2. Study has determined the pulsing peak of the fresh core for this particular condition to be 857.86 MW which is 852 MW according to SAR. Experiment also shows the pulsing peak increases with the increase of reactivity insertion whereas decreases with increase of delayed neutron fraction. With the utilization of the particular values of these parameters, pulsing parameters like prompt energy released, reactor period, pulse width at half maxima, alongwith safety parameters including peak power and clad maximum temperature, have been analyzed. The clad maximum temperature for fresh core is simulated to be 144.54 MW, which is much less than the SAR Value, ensuring the validity of codes and the safety of pulsing in that particular condition. (author)

  11. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  12. Reactor core calculations incorporating subassembly thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Lynas, S.W. [Applied Modelling and Computation Group Imperial Coll. Centre for Environmental Technology Royal School of Mines Prince Consort Road London (United Kingdom); Jones, J.R.

    1997-12-31

    Three dimensional reactor physics calculations performed in parallel with subassembly thermal hydraulic analysis can be used to examine local reactivity effects and increase modelling accuracy. Coupling together codes for coarse mesh neutronics and subassembly thermal hydraulics aids fault studies (fuel clad integrity, safety margin indication etc) and the examination of the interaction between physics and thermal hydraulics during transient events such as LOCA, boron dilution and control rod ejection. Local heating of the coolant decreases reactivity and the fission power peaking factor. Doppler feedback is stronger in the hot region of the fuel, also reducing peak power and reactivity. These thermal hydraulic feedback effects can play an important role in decelerating power excursions and their representation is described in this paper. (author)

  13. Reactor core calculations incorporating subassembly thermal hydraulics

    International Nuclear Information System (INIS)

    Three dimensional reactor physics calculations performed in parallel with subassembly thermal hydraulic analysis can be used to examine local reactivity effects and increase modelling accuracy. Coupling together codes for coarse mesh neutronics and subassembly thermal hydraulics aids fault studies (fuel clad integrity, safety margin indication etc) and the examination of the interaction between physics and thermal hydraulics during transient events such as LOCA, boron dilution and control rod ejection. Local heating of the coolant decreases reactivity and the fission power peaking factor. Doppler feedback is stronger in the hot region of the fuel, also reducing peak power and reactivity. These thermal hydraulic feedback effects can play an important role in decelerating power excursions and their representation is described in this paper. (author)

  14. Researches on a reactor core in heavy ion inertial fusion

    CERN Document Server

    Kondo, S; Iinuma, T; Kubo, K; Kato, H; Kawata, S; Ogoyski, A I

    2016-01-01

    In this paper a study on a fusion reactor core is presented in heavy ion inertial fusion (HIF), including the heavy ion beam (HIB) transport in a fusion reactor, a HIB interaction with a background gas, reactor cavity gas dynamics, the reactor gas backflow to the beam lines, and a HIB fusion reactor design. The HIB has remarkable preferable features to release the fusion energy in inertial fusion: in particle accelerators HIBs are generated with a high driver efficiency of ~30-40%, and the HIB ions deposit their energy inside of materials. Therefore, a requirement for the fusion target energy gain is relatively low, that would be ~50 to operate a HIF fusion reactor with a standard energy output of 1GW of electricity. In a fusion reactor the HIB charge neutralization is needed for a ballistic HIB transport. Multiple mechanical shutters would be installed at each HIB port at the reactor wall to stop the blast waves and the chamber gas backflow, so that the accelerator final elements would be protected from the ...

  15. Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Bess; T. L. Maddock; M. A. Marshall

    2011-09-01

    The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

  16. Methodology for reactor core physics analysis - part 2

    International Nuclear Information System (INIS)

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

  17. NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

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

    International Nuclear Information System (INIS)

    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)

  19. Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Mckerley, Bill [Los Alamos National Laboratory; Bustamante, Jacqueline M [Los Alamos National Laboratory; Costa, David A [Los Alamos National Laboratory; Drypolcher, Anthony F [Los Alamos National Laboratory; Hickey, Joseph [Los Alamos National Laboratory

    2010-01-01

    We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is in support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts

  20. Preliminary Core Analysis of a Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  1. The need to address the larger universe of HEU-fueled reactors, including critical assemblies, pulsed reactors and propulsion reactors

    International Nuclear Information System (INIS)

    Full text: The RERTR program has focused thus far primarily on ending shipments of HEU fuel to research reactors. This has resulted in giving highest priority to reactors with steady thermal powers of 1 megawatt or more, because they require regular refuelling. Critical facilities and pulsed reactors can also of serious concern, because some of them contain very large amounts of barely-irradiated HEU and plutonium. They could be costly to convert - and conversion to LEU may be impractical for fast-neutron critical assemblies. An assessment should be carried out first, therefore, as to which are still needed. Critical assemblies are required today primarily to benchmark Monte Carlo neutron-transport codes. Perhaps the world nuclear community could share a few instead of each reactor-design institute having its own. There is also a whole universe of HEU-fuelled pressurized-water reactors used to power submarines and other types of nuclear-powered ships. These reactors collectively require much more HEU fuel each year than research reactors. The risk of HEU diversion from their fuel cycles is not zero but it is difficult for outsiders to discuss conversion because of the fuel designs are classified. This makes the conversion of Russia's civilian icebreaker reactors of particular interest because issues of classified fuel design are less problematic and these reactors load annually fuel containing about 400 kg of U-235. Another reason for interest in developing LEU fuel for these reactors is that the KLT-40 icebreaker reactor is being adapted for a floating nuclear power plant. Finally, the research-reactor community is, in any case, faced with developing fuels that can operate at power-reactor-fuel temperatures because there are a few high-powered research reactors that operate in this temperature range. (author)

  2. Design of air-core transformer for pulsed current system

    International Nuclear Information System (INIS)

    In this paper, a strip air-core pulse transformer is designed to convert the current. And, how it works and the process of making is elaborated in detail. The transformer contains leads, insulator, copper strip and supporting core. Under the conditions of the charging voltage 2500 V, 5.52 kA and 1.48 kA peak current of primary and secondary windings are obtained, and correspondingly, the current rising rate is 37 A/μs and 138 A/μs. It is supported by analysis and experiment that the transformer can reduce the requirement of the rising rate of the switching current effectively. So, the thyristor can be used in the pulse current system which has a high current rising rate and improve its performance on repetition and stability. (authors)

  3. Pulsed Magnetic Welding for Advanced Core and Cladding Steel

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Guoping [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong [Univ. of Florida, Gainesville, FL (United States)

    2013-12-19

    To investigate a solid-state joining method, pulsed magnetic welding (PMW), for welding the advanced core and cladding steels to be used in Generation IV systems, with a specific application for fuel pin end-plug welding. As another alternative solid state welding technique, pulsed magnetic welding (PMW) has not been extensively explored on the advanced steels. The resultant weld can be free from microstructure defects (pores, non-metallic inclusions, segregation of alloying elements). More specifically, the following objectives are to be achieved: 1. To design a suitable welding apparatus fixture, and optimize welding parameters for repeatable and acceptable joining of the fuel pin end-plug. The welding will be evaluated using tensile tests for lap joint weldments and helium leak tests for the fuel pin end-plug; 2 Investigate the microstructural and mechanical properties changes in PMW weldments of proposed advanced core and cladding alloys; 3. Simulate the irradiation effects on the PWM weldments using ion irradiation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-08-01

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

  5. System Study: Reactor Core Isolation Cooling 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  6. System Study: Reactor Core Isolation Cooling 1998–2012

    Energy Technology Data Exchange (ETDEWEB)

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  7. System Study: Reactor Core Isolation Cooling 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, L.L. (comp.)

    1977-09-01

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

  9. Numerical investigation on the enhancement capability of annular chimney towards natural convective heat transfer in the interior zone of scaled down FBR core catcher

    International Nuclear Information System (INIS)

    Full text of publication follows: A numerical study has been carried out to determine the influence of annular cylindrical chimney on buoyancy-induced flow in the dished end cavity of scaled down Fast Breeder Reactor. Results are presented for (i) cylindrical chimney configuration and (ii) annular chimney configuration occupying the center of the circular plate. Two dimensional laminar simulations are obtained by solving the fully elliptical governing equations of flow and energy. The fluid is Newtonian and incompressible and satisfies the Boussinesq approximation. Results for the upward facing isothermal circular plate with chimney configurations in confined enclosure are analyzed. The velocity fields and isotherms are studied extensively to assess the impact of both geometries on the flow structure, dynamics and overall heat transfer characteristics in the cavity, towards enhancement of natural convective heat transfer. The predicted results for the cylindrical chimney are compared with known experimental results. The results are of interest to post accident heat removal in fast breeder reactors (FBR). (authors)

  10. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors AGENCY... Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance... emergency core cooling systems (ECCSs) of pressurized water reactors (PWRs). This RG also describes...

  11. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs)....

  12. Operation trial at rated power and measurement of xenon poison on Xi'an pulsed reactor

    International Nuclear Information System (INIS)

    The author introduces the general situation of 72 hours continuous operation trial at rated power on Xi'an Pulsed Reactor (XAPR) steady-state core. The experimental results of environmental irradiation dose testing, measurement of equilibrium Xenon poison and Iodine pit reactivity depth while operating at rated full power are also given. The experimental results show that the main body and process systems of XAPR are working order, and that the synthetic performance has reach the design requirements, and that the fuel temperature has more safety margin

  13. Power spectral analysis for a subcritical reactor system driven by a pulsed spallation neutron source

    International Nuclear Information System (INIS)

    A series of power spectral analyses for a thermal subcritical reactor system driven by a pulsed spallation neutron source was carried out at Kyoto University Critical Assembly (KUCA), to determine the prompt-neutron decay constant of the Accelerator-Driven System (ADS). High-energy protons (100 MeV) obtained from the fixed field alternating gradient accelerator were injected onto a lead-bismuth target, whereby the spallation neutrons were generated. In the cross-power spectral density between time-sequence signal data of two neutron detectors, many delta-function-like peaks at the integral multiple of pulse repetition frequency could be observed. However, no continuous reactor-noise component could be measured. This is because these detectors have too high count-rate to be placed closely to the core. From the point data of these delta-function-like peaks, the prompt-neutron decay constant could be determined. At a slightly subcritical state, the decay constant was consistent with that obtained by a previous power spectral analysis for a pulsed 14 MeV neutron source and by a pulsed neutron experiment. At another deeply subcritical state, however, the present analysis leads to an underestimate of the decay constant. (author)

  14. Reactor Core Internals Replacement of Ikata Units 1 and 2

    International Nuclear Information System (INIS)

    Ikata Units 1 and 2 have been in operation for a very long time. Unit 1, in particular, is one of the longest operating PWRs in Japan. In view of this history, preventive and proactive strategy has been adopted for the maintenance of major primary system components. Both units successfully completed the replacement of steam generators and reactor vessel heads approximately ten years ago. With regard to the reactor core internals, baffle former bolts (BFBs) were found to have been damaged by stress corrosion cracking (SCC) in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in other European and U.S. plants, resulting in the replacement of failed BFBs. The BFB issue can be dealt with either by replacing bolts when damage is found or by replacing the entire core internals with those of a new design. Ikata Units 1 and 2 chose the latter and carried it out in 2004 and 2005, respectively.

  15. Partial oxidation of methane in the pulsed compression reactor: experiments and simulation

    NARCIS (Netherlands)

    Roestenberg, Timo; Glushenkov, Maxim; Kronberg, Alexander; Verbeek, Anton A.; Meer, van der Theo H.

    2010-01-01

    The Pulsed Compression Reactor promises to be a compact, economical and energy efficient alternative to conventional chemical reactors. In this article, the production of synthesis gas using the Pulsed Compression Reactor is investigated. This is done experimentally as well as with simulations. The

  16. Small ex-core heat pipe thermionic reactor concept (SEHPTR)

    International Nuclear Information System (INIS)

    The Idaho National Engineering Laboratory (INEL) has developed an innovative space nuclear power concept with unique features and significant advantages for both Defense and Civilian space missions. The Small Ex-core Heat Pipe Thermionic Reactor (SEHPTR) concept was developed in response to Air Force needs for space nuclear power in the range of 10 to 40 kilowatts. This paper describes the SEHPTR concept and discusses the key technical issues and advantages of such a system

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

    International Nuclear Information System (INIS)

    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)

  18. Development of an automated core model for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R.D.

    1998-12-31

    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.

  19. Gas core reactors for actinide transmutation. [uranium hexafluoride

    Science.gov (United States)

    Clement, J. D.; Rust, J. H.; Wan, P. T.; Chow, S.

    1979-01-01

    The preliminary design of a uranium hexafluoride actinide transmutation reactor to convert long-lived actinide wastes to shorter-lived fission product wastes was analyzed. It is shown that externally moderated gas core reactors are ideal radiators. They provide an abundant supply of thermal neutrons and are insensitive to composition changes in the blanket. For the present reactor, an initial load of 6 metric tons of actinides is loaded. This is equivalent to the quantity produced by 300 LWR-years of operation. At the beginning, the core produces 2000 MWt while the blanket generates only 239 MWt. After four years of irradiation, the actinide mass is reduced to 3.9 metric tonnes. During this time, the blanket is becoming more fissile and its power rapidly approaches 1600 MWt. At the end of four years, continuous refueling of actinides is carried out and the actinide mass is held constant. Equilibrium is essentially achieved at the end of eight years. At equilibrium, the core is producing 1400 MWt and the blanket 1600 MWt. At this power level, the actinide destruction rate is equal to the production rate from 32 LWRs.

  20. Reactor core materials research and integrated material database establishment

    International Nuclear Information System (INIS)

    Mainly two research areas were covered in this project. One is to establish the integrated database of nuclear materials, and the other is to study the behavior of reactor core materials, which are usually under the most severe condition in the operating plants. During the stage I of the project (for three years since 1999) in- and out of reactor properties of stainless steel, the major structural material for the core structures of PWR (Pressurized Water Reactor), were evaluated and specification of nuclear grade material was established. And the damaged core components from domestic power plants, e.g. orifice of CVCS, support pin of CRGT, etc. were investigated and the causes were revealed. To acquire more resistant materials to the nuclear environments, development of the alternative alloys was also conducted. For the integrated DB establishment, a task force team was set up including director of nuclear materials technology team, and projector leaders and relevant members from each project. The DB is now opened in public through the Internet

  1. A compact high-voltage pulse generator based on pulse transformer with closed magnetic core.

    Science.gov (United States)

    Zhang, Yu; Liu, Jinliang; Cheng, Xinbing; Bai, Guoqiang; Zhang, Hongbo; Feng, Jiahuai; Liang, Bo

    2010-03-01

    A compact high-voltage nanosecond pulse generator, based on a pulse transformer with a closed magnetic core, is presented in this paper. The pulse generator consists of a miniaturized pulse transformer, a curled parallel strip pulse forming line (PFL), a spark gap, and a matched load. The innovative design is characterized by the compact structure of the transformer and the curled strip PFL. A new structure of transformer windings was designed to keep good insulation and decrease distributed capacitance between turns of windings. A three-copper-strip structure was adopted to avoid asymmetric coupling of the curled strip PFL. When the 31 microF primary capacitor is charged to 2 kV, the pulse transformer can charge the PFL to 165 kV, and the 3.5 ohm matched load can deliver a high-voltage pulse with a duration of 9 ns, amplitude of 84 kV, and rise time of 5.1 ns. When the load is changed to 50 ohms, the output peak voltage of the generator can be 165 kV, the full width at half maximum is 68 ns, and the rise time is 6.5 ns.

  2. Determination of oxygen to uranium plus plutonium atom ratio in high density annular mixed oxide fuel pellets for fast reactor

    International Nuclear Information System (INIS)

    This paper highlights the encountered difficulties and applied modifications in the analytical steps for the determination of [O/(U+Pu)] in high density annular (NatU0.335233U0.37 Pu0.295)O2 pellets, manufactured for irradiation in FBTR and discusses the results. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Samim Anghaie

    2002-08-13

    core. Still there are problems of containment since many of the proposed vessel materials such as W or Mo have high neutron cross sections making the design of a critical system difficult. There is also the possibility for a GCR to remain in a subcritical state, and by the use of a shockwave mechanism, increase the pressure and temperature inside the core to achieve criticality. This type of GCR is referred to as a shockwave-driven pulsed gas core reactor. These two basic designs were evaluated as advance concepts for space power and propulsion.

  4. Tritium control and activation in the Pulse*Star reactor

    International Nuclear Information System (INIS)

    Pulse*Star is an inertial fusion reactor that uses LiPb coolant in a pool type geometry. LiPb does not release great quantities of chemical energy in a fire, and the pool geometry reduces the difficulty of safely transporting the extremely dense fluid. The compact geometry and good neutronics qualities of LiPb lead to a thermal-to-fusion energy ratio of 1.26, a tritium breeding ratio of 1.22, and a net electric power density 29 times higher than in a fission reactor containment building. The afterheat of the coolant and steel is low enough that emergency cooling systems will be either simple or not required. The gamma dose rate of the bell jar or screen is high enough to require remote maintenance of these components. The steam generators and pumps are on the borderline between limited hands-on and remote maintenance. With additional design attention, limited hands-on maintenance could be feasible for these components. The biological hazard potential indicates that only 10-7 to 10-6 of the reactor central region can be vaporized and released; these are values typical of other fusion reactor designs

  5. Measurements with a Pulsed and Modulated Source in a Reactor

    International Nuclear Information System (INIS)

    A generator with a neutron level variable in terms of any time factor has been developed by Philips Research Laboratories. Its practical use. in reactor physics has been demonstrated through a series of measurements carried out in the BRO2 reactor when subcritical. The stability of this generator, and the possibility of introducing sharp variations in the neutron intensity and of pulsing the flux or modulating it sinusoidally, makes it a very versatile instrument. It enables reactivity (ρ = Δk/β) and neutron lifetime (ℓ/β) to be determined by different independent methods. An exact comparison can be made of these methods since they can be employed without changing the conditions under which measurements are carried out. The following were determined: (1) ρ based on delayed neutrons, by a sudden reduction of neutron level, (2) ρ based on prompt neutrons by neutron pulses, (3) (ℓ/β) by a combination of (1) and (2) for 0.5$ < ρ < 2$; and (4) ℓ/β based on the transfer function of the reactor for a modulated source. The transfer functions for a reactivity oscillator and for a sinusoidally modulated source are discussed. It is shown that the measurement of ℓ/β is possible for 0.1 $ < ρ < 10 $ by using a modulated source. The same method also gives the reactivity on the basis of the ratio of prompt neutrons to delayed neutrons for an optimal frequency, practically independently of the data for delayed neutrons and of the value of ℓ/β. By accumulating a large number of cycles in the multi-channel analyser, better statistics for each method can be obtained. Since the neutron level from the generator is in fact sinusoidal, the response of the reactor may be integrated over each quarter of a period, as the measurement sequence is controlled by the generator; measurement time is then minimal. Observations recorded on a perforated tape are analysed by a digital computer

  6. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm2.s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm2.s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U02-12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at 30 k

  7. Aspects of cell calculations in deterministic reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-02-15

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

  8. Optimization of ultra-long cycle fast reactor core

    International Nuclear Information System (INIS)

    An optimization of an ultra-long cycle fast reactor (UCFR) design with a power rate of 1000 MW (electric), UCFR-1000, has been performed to increase the safety of UCFR. Firstly, geometric optimization has been performed to decrease its peaking factors so that the peak temperatures measured by thermal hydraulic feedback are within the limit of design basis event (DBE). Secondly, fuel composition optimization has been performed by adopting Pressurized Water Reactor (PWR) spent fuel as a blanket material instead of natural uranium. Lastly, a small-size UCFR with a power rate of 100 MWe, UCFR-100, has been proposed for developing a short term deployable nuclear reactor. The major optimization process for UCFR-100 is decreasing maximum neutron flux and fast neutron fluence. The optimized UCFR-1000 has been enlarged radially and shortened axially from the initial UCFR design and this modification makes the burning speed of active core movement slower. It has been confirmed that a full-power operation of 60 years without refueling is feasible for both UCFR-1000 and UCFR-100 core designs by a breed-and-burn strategy. By the design optimization study, the reductions of maximum neutron flux, fast neutron fluence, and axial power peaking have been achieved, which are favorable for the safety of the UCFR. (author)

  9. Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Zhujie [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Bartels, David [Univ. of Notre Dame, IN (United States)

    2015-04-02

    This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that the effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.

  10. Cronos 2: a neutronic simulation software for reactor core calculations

    International Nuclear Information System (INIS)

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  11. Excitation of neutron flux waves in reactor core transients

    International Nuclear Information System (INIS)

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

  12. Heat transfer evaluation in a plasma core reactor

    International Nuclear Information System (INIS)

    Numerical evaluations of heat transfer in a fissioning uranium plasma core reactor cavity, operating with seeded hydrogen propellant, have been performed. A two-dimensional analysis is based on an assumed flow pattern and cavity wall heat exchange rate. Various iterative schemes were required by the nature of the radiative field and by the solid seed vaporization. Approximate formulations of the radiative heat flux are generally used, due to the complexity of the solution of a rigorously formulated problem. The present work analyzes the sensitivity of the results with respect to approximations of the radiative field, geometry, seed vaporization coefficients and flow pattern. The results present temperature, heat flux, density and optical depth distributions in the reactor cavity, acceptable simplifying assumptions, and iterative schemes

  13. Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

    International Nuclear Information System (INIS)

    Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

  14. Theory and experiment of Fourier-Bessel field calculation and tuning of a pulsed wave annular array

    DEFF Research Database (Denmark)

    Fox, Paul D.; Jiqi, Cheng; Jian-yu, Lu

    2003-01-01

    and tuning the propagated field by linking the quantized surface pressure profile to a known set of limited diffraction Bessel beams propagating into the medium. This enables derivation of an analytic expression for the field at any point in space and time in terms of the transducer surface pressure profile....... Tuning of the field then also follows by formulating a least-squares design for the transducer surface pressure with respect to a given desired field in space and time. Simulated and experimental results for both field computation and tuning are presented in the context of a 10-ring annular array...

  15. A supercomputing application for reactors core design and optimization

    International Nuclear Information System (INIS)

    Advanced nuclear reactor designs are often intuition-driven processes where designers first develop or use simplified simulation tools for each physical phenomenon involved. Through the project development, complexity in each discipline increases and implementation of chaining/coupling capabilities adapted to supercomputing optimization process are often postponed to a further step so that task gets increasingly challenging. In the context of renewal in reactor designs, project of first realization are often run in parallel with advanced design although very dependant on final options. As a consequence, the development of tools to globally assess/optimize reactor core features, with the on-going design methods accuracy, is needed. This should be possible within reasonable simulation time and without advanced computer skills needed at project management scale. Also, these tools should be ready to easily cope with modeling progresses in each discipline through project life-time. An early stage development of multi-physics package adapted to supercomputing is presented. The URANIE platform, developed at CEA and based on the Data Analysis Framework ROOT, is very well adapted to this approach. It allows diversified sampling techniques (SRS, LHS, qMC), fitting tools (neuronal networks...) and optimization techniques (genetic algorithm). Also data-base management and visualization are made very easy. In this paper, we'll present the various implementing steps of this core physics tool where neutronics, thermo-hydraulics, and fuel mechanics codes are run simultaneously. A relevant example of optimization of nuclear reactor safety characteristics will be presented. Also, flexibility of URANIE tool will be illustrated with the presentation of several approaches to improve Pareto front quality. (author)

  16. Matching a (sub)nanosecond pulse source to a corona plasma reactor

    Science.gov (United States)

    Huiskamp, T.; Beckers, F. J. C. M.; Hoeben, W. F. L. M.; van Heesch, E. J. M.; Pemen, A. J. M.

    2016-10-01

    In this paper we investigate the energy transfer from the pulses of a (sub)nanosecond pulse source to the plasma in a corona-plasma reactor. This energy transfer (or ‘matching’) should be as high as possible. We studied the effect of multiple parameters on matching, such as the reactor configuration, the pulse duration and amplitude and the energy density. The pulse reflection on the reactor interface has a significant influence on matching, and should be as low as possible to transfer the most energy into the reactor. We developed a multiple-wire inner conductor for the reactor which decreases the vacuum impedance of the reactor to decrease the pulse reflection on the reactor interface while maintaining a high electric field on the wire. The results were very encouraging and showed an energy transfer efficiency of over 90 percent. The matching results further show that there is only a small effect on the matching between different wire diameters. In addition, a long reactor and a long pulse result in the best matching due to the more intense plasma that is generated in these conditions. Finally, even without the multiple-wire reactor, we are able to achieve a very good matching (over 80 percent) between our pulse source and the reactor.

  17. VIPRE modeling of VVER-1000 reactor core for DNB analyses

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

    1995-09-01

    Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

  18. The JASON reactor: from core removal to fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Beeley, P.; Williams, A.; Lockwood, R. [Defence College of Electromechanical Engineering, Nuclear Dept., HMS SULTAN (United Kingdom); Raymond, B.; Spyrou, N. [Surrey Univ., Dept. of Physical and Electronic Sciences (United Kingdom); Auziere, P. [AREVA NC, Treatment Business Unit, 78 - Velizy (France)

    2007-07-01

    The 10 kW JASON Argonaut reactor was operated at the Royal Naval College, Greenwich, London, between 1962 and 1996. After initial cooling in the core, the MTR type fuel (80% enriched U{sup 235}) was dry stored on site before transport in 1998 to BNFL, Sellafield for interim wet storage. Arrangements for reprocessing of the fuel at AREVA NC, La Hague are now in progress and this paper will describe various aspects of the storage, transfer, monitoring, and the treatment at La Hague plant. The radioactive waste resulting from the processing of these used fuels will be conditioned into a suitable package for return to UK.

  19. Impedance change measurements of a superconducting shielded-core reactor

    International Nuclear Information System (INIS)

    A device was constructed using a stack of superconducting rings surrounding a ferrite rod, with the assembly inserted in a high turns count solenoid. Superconducting end pieces were also placed at either end of the rod to minimize flux leakage to the ferrite rod. The superconducting rings act as a magnetic shield to the ferrite, effectively eliminating the low reluctance path the ferrite offers. At a specific field the superconductor will be fully penetrated, placing the ferrite in the magnetic circuit and reducing the reactance offered by the solenoidal winding. In this mode of operation the shielded core reactor can be applied as a current limiting device. Results included in this paper, indicate that in the best design achieved leakage to the ferrite core could not be eliminated. The superconducting current induced by this leakage eliminated the low reluctance path of the ferrite by producing a counter-flux in the core exactly opposing the applied field. Shielding currents set up by penetration of the externally applied field were found to be minimal compared to the induced currents caused by leakage flux in the ferrite core

  20. Benchmark problems of start-up core physics of High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    The experimental data of the HTTRs start-up core physics are useful to verify design codes of commercial HTGRs due to the similarities in the core size and excess reactivity. Form these viewpoints, it is significant to carry out the bench mark tests of design codes by using data of start-up core physics experiments planned for the HTTR. The evaluations of the first criticality, excess reactivity of annular cores, etc., are proposed for the benchmark problem. It was found from our precalculations that diffusion calculations provide larger excess reactivity and small number of fuel columns for the first criticality than Monte Carlo calculations. 19 refs

  1. Machine learning of the reactor core loading pattern critical parameters

    International Nuclear Information System (INIS)

    The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employed a recently introduced machine learning technique, Support Vector Regression (SVR), which has a strong theoretical background in statistical learning theory. Superior empirical performance of the method has been reported on difficult regression problems in different fields of science and technology. SVR is a data driven, kernel based, nonlinear modelling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modelling. The starting set of experimental data for training and testing of the machine learning algorithm was obtained using a two-dimensional diffusion theory reactor physics computer code. We illustrate the performance of the solution and discuss its applicability, i.e., complexity, speed and accuracy, with a projection to a more realistic scenario involving machine learning from the results of more accurate and time consuming three-dimensional core modelling code. (author)

  2. Thermohydraulics of emergency core cooling in light water reactors

    International Nuclear Information System (INIS)

    This report, by a group of experts of the OECD-NEA Committee on the Safety of Nuclear Installations, reviews the current state-of-knowledge in the field of emergency core cooling (ECC) for design-basis, loss-of-coolant accidents (LOCA) and core uncover transients in pressurized- and boiling-water reactors. An overview of the LOCA scenarios and ECC phenomenology is provided for each type of reactor, together with a brief description of their ECC systems. Separate-effects and integral-test facilities, which contribute to understanding and assessing the phenomenology, are reviewed together with similarity and scaling compromises. All relevant LOCA phenomena are then brought together in the form of tables. Each phenomenon is weighted in terms of its importance to the course of a LOCA, and appraised for the adequacy of its data base and analytical modelling. This qualitative procedure focusses attention on the modelling requirements of dominant LOCA phenomena and the current capabilities of the two-fluid models in two-phase flows. This leads into the key issue with ECC: quantitative code assessment and the application of system codes to predict with a well defined uncertainty the behaviour of a nuclear power plant. This issue, the methodologies being developed for code assessment and the question of how good is good enough are discussed in detail. Some general conclusions and recommendations for future research activities are provided

  3. A neural network to predict reactor core behaviors

    International Nuclear Information System (INIS)

    The global fuel management problem in BWRs (Boiling Water Reactors) can be understood as a very complex optimization problem, where the variables represent design decisions and the quality assessment of each solution is done through a complex and computational expensive simulation. This last aspect is the major impediment to perform an extensive exploration of the design space, mainly due to the time lost evaluating non promising solutions. In this work, we show how we can train a Multi-Layer Perceptron (MLP) to predict the reactor behavior for a given configuration. The trained MLP is able to evaluate the configurations immediately, thus allowing performing an exhaustive evaluation of the possible configurations derived from a stock of fuel lattices, fuel reload patterns and control rods patterns. For our particular problem, the number of configurations is approximately 7.7 x 1010; the evaluation with the core simulator would need above 200 years, while only 100 hours were required with our approach to discern between bad and good configurations. The later were then evaluated by the simulator and we confirm the MLP usefulness. The good core configurations reached the energy requirements, satisfied the safety parameter constrains and they could reduce uranium enrichment costs. (authors)

  4. Calculation of Kinetic Parameters of TRIGA Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka; Kavcic, Andrej; Zerovnik, Gasper; Ravnik, Matjaz [' Jozef Stefan' Institute, Jamova 39, SI-1000 Ljubljana (Slovenia)

    2008-10-29

    Modern Monte Carlo transport codes in combination of fast computer clusters enable very accurate calculations of the most important reactor kinetic parameters. Such are the effective delayed neutron fraction, {beta}{sub eff}, and mean neutron generation time, {lambda}. We calculated the {beta}{sub eff} and {lambda} for various realistic and hypothetical annular TRIGA Mark II cores with different types and amount of fuel. It can be observed that the effective delayed neutron fraction strongly depends on the number of fuel elements in the core or on the core size. E.g., for 12 wt. % uranium standard fuel with 20 % enrichment, {beta}{sub eff} varies from 0.0080 for a small core (43 fuel rods) to 0.0075 for a full core (90 fuel rods). It is interesting to note that calculated value of {beta}{sub eff} strongly depends also on the delayed neutron nuclear data set used in calculations. The prompt neutron life-time mainly depends on the amount (due to either content or enrichment) of {sup 235}U in the fuel as it is approximately inversely proportional to the average absorption cross-section of the fuel. E.g., it varies from 28 {mu}s for 30 wt. % uranium content fuelled core to 48 {mu}s for 8.5 wt. % uranium content LEU fuelled core. The results are especially important for pulse mode operation and analysis of the pulses. (authors)

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

    International Nuclear Information System (INIS)

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

  6. A comparison of measurements of atmospheric ammonia by filter packs, transition-flow reactors, simple and annular denuders and fourier transform infrared spectroscopy

    Science.gov (United States)

    Wiebe, H. A.; Anlauf, K. G.; Tuazon, E. C.; Winer, A. M.; Biermann, H. W.; Appel, B. R.; Solomon, P. A.; Cass, G. R.; Ellestad, T. G.; Knapp, K. T.; Peake, E.; Spicer, C. W.; Lawson, D. R.

    Using data obtained during the 1985 Nitrogen Species Methods Comparison Study (1988, Atmospheric Environment22, 1517), several measurement methods for sampling ambient NH 3 are compared. Eight days of continuous measurements at Pomona College, a smog receptor site in Los Angeles, provided an extensive data base for comparing the following methods: Fourier transform i.r. spectroscopy (FTIR), three filter pack configurations, a simple and an annular denuder, and the transition flow reactor. FTIR was defined as the reference method and it reported hourly NH 3 concentrations ranging from > 60 to 2280 nmol m -3 (1.5-57ppb) during the course of the study, the highest values coming from the influence of nearby livestock operations. Although only limited quality assurance procedures were carried out, the following conclusions can, nevertheless, be drawn: most of the methods correlated highly with the FTIR method (correlation coefficient r > 0.96); generally, the linear regression slopes were close to unity and the intercepts were insignificantly different from zero at the 95% confidence level); relative to the FTIR average values, (1) for 4-6 h sampling periods, the averages of the three filter packs from three research groups were 83-130% and the annular denuder average was 87%, and (2) for 10-12 h sampling periods, the simple denuder averaged 90% and the two transition flow reactors were 77-98%. Possible reasons for the reported systematic biases are presented, but these are not able to fully explain the large range of differences reported by the various methods.

  7. New results from pulse tests in the CABRI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, F.; Papin, J.; Haessler, M. [Institut de Proterction et de Surete Nucleaire, Saint Paul Lez Durance (France)] [and others

    1996-03-01

    At the 21st and 22nd WRSM (1,2), the motivation and objectives of the French program on the behaviour of high burnup PWR fuel under RIA conditions in the CABRI test reactor has been presented. The major results of the three first tests of the test matrix were presented and in particular REP-Na1, which failed at an unexpected low level of fuel enthalpy, was exposed to the community of nuclear safety research. At this time, no final understanding was reached for the origin of the failure. This objective is reached now. Two further tests, REP-Na4 and 5, have been performed in 1995, they demonstrated a satisfactory and safe behaviour by resisting to the early phase of severe loading during the RIA pulse test. Further examination work and analytical testing is in progress and the next tests with MOX fuel are being prepared.

  8. Real time simulation research in 200 MW low temperature nuclear heating reactor core

    International Nuclear Information System (INIS)

    200 MW low temperature nuclear heating reactor is an important new-type reactor. Natural circulation is adopted in the flowage of reactor core. High precise models are built and selected, which are low temperature reactor power model, residual power releasing model, heat conductivity model in reactor core, thermo-hydraulic model, subcooling boiling model, CHF calculation model and so on. These models are solved using Gear arithmetic and Adams arithmetic, which are testified each other. Using appropriate arithmetic, the real time simulation of thermo-hydraulic process in the core is truly fulfilled. (authors)

  9. In-reactor experiments in fast breeder test reactor for assessment of core structural materials

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India is a sodium cooled reactor with neutron flux level of the order of 1015 n/cm2/s and temperature of coolant in the range of 650-790K (380-520oC). This reactor is being used as a test bed for the development of fuel and structural materials required for Indian Fast Reactor Programme. FBTR is also used as a test facility to carry out accelerated irradiation tests on thermal reactor structural materials. In-reactor experiments on core structural materials are being carried out by subjecting prefabricated specimens to desired conditions of temperature and neutron fluence levels in FBTR. Non-instrumented irradiation capsules that can be loaded at any location of FBTR core are used for the experiments. Pressurised capsules of zirconium alloys have been developed and subjected to irradiation in FBTR to determine the irradiation creep rate of indigenously developed zirconium alloys (Zircaloy-2 and Zr-2.5%Nb alloy) for life assessment of pressure tubes of Indian Pressurised Heavy Water Reactors (PHWRs). Technology development of pressurised capsules was carried out at IGCAR. These pressurised capsules were filled with argon and a small fraction of helium at a high pressure (5.0-6.5 MPa at room temperature) in such a way that the target stresses were attained in the walls of the pressurised capsules at the desired temperature of irradiation in the reactor. FBTR was operated at a low power of 8 MWt during this irradiation campaign to have an inlet temperature of about 579 K (306oC) which was close to the temperature of pressure tubes at full power in PHWR. Irradiation of thirty pressurised capsules was carried out in FBTR using six irradiation capsules for different durations (upto 79 days). The fluence levels attained by the pressurised capsules were up to 1.1 x 1021 n/cm2 (E> 1 MeV) at temperatures of 579 to 592 K. Post-irradiation increase in diameter of the pressurised

  10. Core conversion of the Portuguese research reactor to LEU fuel

    International Nuclear Information System (INIS)

    Core conversion of the Portuguese Research Reactor (RPI) to LEU fuel is being performed within IAEA's Technical Cooperation project POR/4/016, with financial support from the US and Portugal. CERCA was selected as manufacturer of the LEU assemblies by the IAEA after an international call for bids. CERCA provided a comprehensive package to the RPI which included the mechanical verification of the design of the assemblies, their manufacture and arrangements for a joint inspection of the finished assemblies. The LEU fuel assemblies were manufactured within 8 months upon final approval of the design. The safety analyses for the core conversion to LEU fuel were made with the assistance of the RERTR program and were submitted for review by the IAEA and by Portuguese authorities in January 2007. Revised documents were submitted in June 2007 addressing the issues raised during review. Regulatory approval was received in early August and core conversion was done in early September. All measured safety parameters are within the defined acceptance limits. Operation at full power is expected by the end of October. (author)

  11. PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

    Science.gov (United States)

    Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

    2014-06-01

    The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

  12. Fast reactor core management in Japan: twenty years of evolution at JOYO

    International Nuclear Information System (INIS)

    Twenty years of operations at the experimental fast reactor JOYO provide a wealth of experience with core and fuel management. This experience has been applied to several core modifications to upgrade JOYO's irradiation capability. Core physics tests and Post Irradiation Examination (PIE) results have been used to confirm the accuracy of neutron diffusion theory calculations. These experiences and accumulated data will be useful for the core design in future fast reactors in Japan's development. (author)

  13. Machine Learning of the Reactor Core Loading Pattern Critical Parameters

    International Nuclear Information System (INIS)

    The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm, and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper, we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employ a recently introduced machine learning technique, support vector regression (SVR), which is a data driven, kernel based, nonlinear modeling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modeling. We illustrate the performance of the solution and discuss its applicability, that is, complexity, speed, and accuracy

  14. Device for assisting the operation and administration of reactor cores

    International Nuclear Information System (INIS)

    Purpose: To enable even unskilled persons to select adequate control rod planning in the same manner as done by the skilled designers. Constitution: Information showing the state of the reactor core before the control rod operation, for example, the control rod pattern and the power distribution, and the control rod alteration pattern after the control rod operation are inputted into an input device, while data base previously prepared based on the considerations of skilled designers are stored in the data base memory device. The control rod change pattern and the power distribution are inputted by the input device to the adequacy judging device for the control rod relative position and the stored data base are read out to determine the adequacy for the control relative position. The result is outputted to the judging device to display the adequacy. (Sekiya, K.)

  15. Recriticality in a BWR [boiling water reactor] following a core damage event

    International Nuclear Information System (INIS)

    This report describes the results of a study conducted by Pacific Northwest Laboratory to assist the US Nuclear Regulatory Commission in evaluating the potential for recriticality in boiling water reactors (BWRs) during certain low probability severe accidents. Based on a conservative bounding analysis, this report concludes that there is a potential for recriticality in BWRs if core reflood occurs after control blade melting has begun but prior to significant fuel rod melting. However, a recriticality event will most likely not generate a pressure pulse significant enough to fail the vessel. Instead, a quasi-steady power level would result and the containment pressure and temperature would increase until the containment failure pressure is reached, unless actions are taken to terminate the event. Two strategies are identified that would aid in regaining control of the reactor and terminate the recriticality event before containment failure pressures are reached. The first strategy involves initiating boration injection at or before the time of core reflood if the potential for control blade melting exists. The second strategy involves initiating residual heat removal suppression pool cooling to remove the heat load generated by the recriticality event and thus extend the time available for boration. 31 figs., 17 tabs

  16. LMFBR type reactor core and its fuel exchange method

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-08-20

    Upon initial loading, two kinds of fuel assemblies including first fuel assemblies having a highest enrichment degree and second fuel assemblies having a lowest enrichment degree are loaded. The average fuel enrichment degree of an upper region of the first fuel assembly is made greater than that of the lower region. The reactivity of the lower region of the first fuel assembly is made lower than that of the upper portion to reduce power peak. Upon transfer from a first cycle to a second cycle, at least one of the second fuel assemblies is exchanged by the same number of the third fuel assemblies. In this case, an average fuel enrichment degree of the upper region of the third fuel assembly is made greater than that of the lower region to suppress the reactivity in the lower region of the third fuel assembly lower than the reactivity in the upper region thereby reducing the power peak. Thus, the upper power peak over the entire reactor core is moderated thereby capable of ensuring the reactor shut down margin without deteriorating the same. (N.H.)

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

    International Nuclear Information System (INIS)

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

  18. Langmuir probe diagnostic studies of pulsed hydrogen plasmas in planar microwave reactors

    OpenAIRE

    A. Rousseau(MSSL, Surrey, United Kingdom); Teboul, E.; Lang, N.; M. Hannemann; Röpcke, J.

    2002-01-01

    Langmuir probe techniques have been used to study time and spatially resolved electron densities and electron temperatures in pulse-modulated hydrogen discharges in two different planar microwave reactors (fmicrowave= 2.45 GHz, tpulse= 1 ms). The reactors are (i) a standing-wave radiative slotted waveguide reactor and (ii) a modified travelling-wave radiative slotted waveguide reactor, which generate relatively large plasmas over areas from about 350 cm^2 to 500 cm^2. The plasma properties of...

  19. Nuclear reactor pulse calibration using a CdZnTe electro-optic radiation detector

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, Kyle A., E-mail: knelson1@ksu.edu [S.M.A.R.T. Laboratory, Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States); Geuther, Jeffrey A. [TRIGA Mark II Nuclear Reactor, Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States); Neihart, James L.; Riedel, Todd A. [S.M.A.R.T. Laboratory, Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States); Rojeski, Ronald A. [Nanometrics, Inc., 1550 Buckeye Drive, Milpitas, CA 95035 (United States); Saddler, Jeffrey L. [TRIGA Mark II Nuclear Reactor, Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States); Schmidt, Aaron J.; McGregor, Douglas S. [S.M.A.R.T. Laboratory, Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States)

    2012-07-15

    A CdZnTe electro-optic radiation detector was used to calibrate nuclear reactor pulses. The standard configuration of the Pockels cell has collimated light passing through an optically transparent CdZnTe crystal located between crossed polarizers. The transmitted light was focused onto an IR sensitive photodiode. Calibrations of reactor pulses were performed using the CdZnTe Pockels cell by measuring the change in the photodiode current, repeated 10 times for each set of reactor pulses, set between 1.00 and 2.50 dollars in 0.50 increments of reactivity. - Highlights: Black-Right-Pointing-Pointer We demonstrated the first use of an electro-optic device to trace reactor pulses in real-time. Black-Right-Pointing-Pointer We examined the changes in photodiode current for different reactivity insertions. Black-Right-Pointing-Pointer Created a linear best fit line from the data set to predict peak pulse powers.

  20. Compression of realistic laser pulses in hollow-core photonic bandgap fibers

    DEFF Research Database (Denmark)

    Lægsgaard, Jesper; Roberts, John

    2009-01-01

    Dispersive compression of chirped few-picosecond pulses at the microjoule level in a hollow-core photonic bandgap fiber is studied numerically. The performance of ideal parabolic input pulses is compared to pulses from a narrowband picosecond oscillator broadened by self-phase modulation during...... amplification. It is shown that the parabolic pulses are superior for compression of high-quality femtosecond pulses up to the few-megawatts level. With peak powers of 5-10 MW or higher, there is no significant difference in power scaling and pulse quality between the two pulse types for comparable values...

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

    DEFF Research Database (Denmark)

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

    2015-01-01

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

  2. Investigation of activity release during light water reactor core meltdown

    International Nuclear Information System (INIS)

    A test facility was developed for the determination of activity release and of aerosol characteristics under realistic light water reactor core melting conditions. It is composed of a high-frequency induction furnace, a ThO2 crucible system, and a collection apparatus consisting of membrane and particulate filters. Thirty-gram samples of a representative core material mixture (corium) were melted under air, argon, or steam at 0.8 to 2.2 bar. In air at 27000C, for example, the relative release was 0.4 to 0.7% for iron, chromium, and cobalt and 4 to 11% for tin, antimony, and manganese. Higher release values of 20 to 40% at lower temperatures (21500C, air) were found for selenium, cadmium, tellurium, and cesium. The size distribution of the aerosol particles was trimodal with maxima at diameters of 0.17, 0.30, and 0.73 μm. The result of a qualitative x-ray microanalysis was that the main elements of the melt were contained in each aerosol particle. Further investigations will include larger melt masses and the additional influence of concrete on the release and aerosol behavior

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. The influence of reactor core parameters on effective breeding coefficient Keff

    Institute of Scientific and Technical Information of China (English)

    Liu Li-Po; Liu Yi-Bao; Wang Juan; Yang Bo; Zhang Tao

    2008-01-01

    The values of effective breeding coefficient Keff in a reactor core of nuclear power plant are calculated for different values of parameters (core structure, fuel assembly component) by using the Monte Carlo method. The obtained values of Keff are compared and analysed, which can provide theoretical basis for reactor design.

  6. Study of magnetic particles pulse-injected into an annular SPLITT-like channel inside a quadrupole magnetic field.

    Science.gov (United States)

    Hoyos, M; Moore, L R; McCloskey, K E; Margel, S; Zuberi, M; Chalmers, J J; Zborowski, M

    2000-12-01

    Advantages of the continuous magnetic flow sorting for biomedical applications over current, batch-wise magnetic separations include high throughput and a potential for scale-up operations. A continuous magnetic sorting process has been developed based on the quadrupole magnetic field centered on an annular flow channel. The performance of the sorter has been described using the conceptual framework of split-flow thin (SPLITT) fractionation, a derivative of field-flow fractionation (FFF). To eliminate the variability inherent in working with a heterogenous cell population, we developed a set of monodisperse magnetic microspheres of a characteristic magnetization, and a magnetophoretic mobility, similar to those of the cells labeled with a magnetic colloid. The theory of the magnetic sorting process has been tested by injecting a suspension of the magnetic beads into the carrier fluid flowing through the sorter and by comparing the theoretical and experimental recovery versus total flow-rate profiles. The position of the recovery maxima along the total flow-rate axis was a function of the average bead magnetophoretic mobility and the magnetic field intensity. The theory has correctly predicted the position of the peak maxima on the total flow-rate axis and the dependence on the bead mobility and the field intensity, but has not correctly predicted the peak heights. The differences between the calculated and the measured peak heights were a function of the total flow-rate through the system, indicating a fluid-mechanical origin of the deviations from the theory (such as expected of the lift force effects in the system). The well-controlled elution studies using the monodisperse magnetic beads, and the SPLITT theory, provided us with a firm basis for the future sorter evaluation using cell mixtures. PMID:11153960

  7. Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    EL-Khawlani, Afrah [Physics Department, Sana' a (Yemen); Aziz, Moustafa [Nuclear and radiological regulatory authority, Cairo (Egypt); Ismail, Mahmud Yehia; Ellithi, Ali Yehia [Cairo Univ. (Egypt). Faculty of Science

    2015-03-15

    The MCNPX (Monte Carlo N-Particle Transport Code System) code has been used for modeling and simulation of a half core of CANDU (CANada Deuterium-Uranium) reactor, both homogenous and heterogeneous model for the reactor core are designed. The fuel is burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor for homogeneous and heterogeneous reactor core is calculated and compared during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the model. The flux and power distributions through channels are calculated.

  8. Design and fabrication of hollow-core photonic crystal fibers for high-power ultrashort pulse transportation and pulse compression.

    Science.gov (United States)

    Wang, Y Y; Peng, Xiang; Alharbi, M; Dutin, C Fourcade; Bradley, T D; Gérôme, F; Mielke, Michael; Booth, Timothy; Benabid, F

    2012-08-01

    We report on the recent design and fabrication of kagome-type hollow-core photonic crystal fibers for the purpose of high-power ultrashort pulse transportation. The fabricated seven-cell three-ring hypocycloid-shaped large core fiber exhibits an up-to-date lowest attenuation (among all kagome fibers) of 40 dB/km over a broadband transmission centered at 1500 nm. We show that the large core size, low attenuation, broadband transmission, single-mode guidance, and low dispersion make it an ideal host for high-power laser beam transportation. By filling the fiber with helium gas, a 74 μJ, 850 fs, and 40 kHz repetition rate ultrashort pulse at 1550 nm has been faithfully delivered at the fiber output with little propagation pulse distortion. Compression of a 105 μJ laser pulse from 850 fs down to 300 fs has been achieved by operating the fiber in ambient air.

  9. Pulsed Neutron Measurements on a Heavy Water Power Reactor (MZFR) at Zero Energy

    International Nuclear Information System (INIS)

    The pulsed neutron method was used for zero-power measurements in the core of a heavy water reactor. Various methods were used for the evaluation of the pulsed measurements. The so-called ''integral'' evaluation methods are based on theories published by Sjöstrand and Gozani; so far they have been applied mainly to light water reactors. These methods use not only the prompt neutron decay constant but also the information contained in the delayed neutron tails to determine the reactivity. For measurements on the heavy water reactor, however, the methods had to be modified so as to adequately take into account the time dependence of the delayed neutrons. The fraction of the delayed neutrons was calculated using a reasonable assumption for its time dependence. All the information needed could be obtained from the measurements. These methods are well suited for hand calculations to yield the reactivity with proper accuracy. An analytical procedure was applied to check the results of the integral methods. This essentially involves the exact calculation of the time dependence of the delayed neutron fraction by an iteration procedure. The results of the different evaluation methods mentioned above are compared by plotting them as functions of the D2O level and of the boron concentration. Due to the inclined control rods the flux distribution is distorted in a rather complicated manner when the rods are inserted. Therefore the time dependence of this distribution was measured for different positions of the pulsed neutron source. It was possible to find one position for which the influence of higher modes on the measurements of the shutdown reactivity was sufficiently small. Finally it is shown that the values of (δρ(H, ci)/δ(l/H2)) H = Hi and (δρ(Hi, c)/δc) c = ci (ρ reactivity, Hi critical D2O level for boron concentration c1) obtained by period measurements in the slightly supercritical state and pulsed measurements in the subcritical state are in excellent

  10. Emergency reactor core cooling water injection device for light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Junro.

    1994-05-13

    A reactor pressure vessel is immersed in pool water of a reactor container. A control valve is interposed to a water supplying pipelines connecting pool water and a pressure vessel. A valve actuation means for opening/closing the control valve comprises a lifting tank. The inner side of the lifting tank and the inner side of the pressure vessel are connected by a communication pipeline (a syphon pipe) at upper and lower two portions. The lifting tank and the control valve are connected by a link mechanism. When a water level in the pressure vessel is lowered, the water level in the lifting tank is lowered to the same level as that in the pressure vessel. This reduces the weight of the lifting tank, the lifting tank is raised, to open the control valve by way of a link mechanism. As a result, liquid phase in the pressure vessel is in communication with the pool water, and the pool water flows down into the pressure vessel to maintain the reactor core in a flooded state. (I.N.).

  11. Behaviour of steel pipe exposed to fouling by heavy oil during core-annular flow; Comportamento de tubo de aco exposto a sujeira de oleo pesado durante escoamento nucleo-anular

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Adriana; Bannwart, Antonio C. [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia Mecanica. Dept. de Engenharia de Petroleo

    2004-07-01

    The use of water-assisted technologies such as core-annular flow to the pipelines of viscous oils has been proposed as an attractive alternative for production and transportation of heavy crudes in both onshore and offshore scenarios. Usually, core-annular flow can be created by injecting a relatively small water flow rate laterally in the pipe, so as to form a thin water annulus surrounding the viscous oil, which is pumped through the center. The reduction in friction losses obtained thanks to lubrication by water is significant, since the pressure drop in a steady state core flow becomes comparable to water flow only. For a complete assessment of core flow technology, however, unwanted effects associated with possible oil adhesion onto the pipe wall should be investigated, since these may cause severe fouling of the wall and pressure drop increase. It has been observed that oil adhesion on metallic surfaces may occur for certain types of crude and oilphilic pipe materials. In this work we present results of pressure drop monitoring during 35 hour-operation of a heavy oil-water core annular flow in a 26.08 mm. i.d. horizontal steel pipe. The oil used is described in terms of its main components and the results of static wet ability tests are also presented for comparison (author)

  12. Modeling TRIGA reactor pulses using the STAR 3D nodal kinetics and WIMS-D4 codes

    International Nuclear Information System (INIS)

    A detailed three-dimensional (3D) time-dependent STAR nodal kinetics model coupled to a one-dimensional (1D) thermal-hydraulics WIGL model has been developed to describe and benchmark the peak power and pulse behavior of the Penn State University (PSU) Breazeale TRIGA reactor. Different core loading patterns were used for several TRIGA pulse tests with different reactivity insertion worths (1.5 dollar, 2.0 dollar, 2.5 dollar). The STAR nodal kinetics code and TRIGA model adequately simulates TRIGA pulses when group constants are generated from physics codes (i.e., WIMS-D4) that can accurately model the TRIGA uranium-zirconium-hydride fuel

  13. [Granuloma annulare].

    Science.gov (United States)

    Butsch, F; Weidenthaler-Barth, B; von Stebut, E

    2015-11-01

    Granuloma annulare is a benign, chronic inflammatory skin disease. Its pathogenesis is still unclear, but reports on infections as a trigger can be found. In addition, some authors reported an association with other systemic disease, e.g., cancer, trauma, and diabetes mellitus; however, these have not been verified. The clinical picture of granuloma annulare ranges from the localized form predominantly at the extremities to disseminated, subcutaneous, or perforating forms. Diagnosis is based on the typical clinical presentation which may be confirmed by a biopsy. Histologically, necrobiotic areas within granulomatous inflammation are typical. The prognosis of the disease is good with spontaneous resolution being frequently observed, especially in localized forms. Disseminated manifestations tend to persist longer, and recurrences are reported. When choosing between different therapeutic options, the benign disease character versus the individual degree of suffering and the potential therapy side effects must be considered. For local treatment, topical application of corticosteroids is most common. Disseminated forms can be treated systemically with corticosteroids for several weeks; alternatively, dapsone, hydroxychloroquine, retinoids, fumaric acid, cyclosporine, and anti-TNFα appear to be effective. PMID:26487494

  14. Measurement of the vacuum reactivity coefficient of the RP-0 reactor 7A4 core

    International Nuclear Information System (INIS)

    Estimate results of the vacuum reactivity coefficient of the RP-0 reactor 7A4 core through the inverse kinetics and neutronic noise are presented. For this effect, a compensated ionization chamber was used at the position E2 of the core. Experience was carried out at 0,47 W power which was monitored by the same measurement equipment. Aluminum blades were used as vacuum in different configurations within the reactor core. Results were assessed through perturbation theory to an energy group

  15. FARM: a new tool for optimizing the core performance and safety characteristics of gas cooled fast reactor cores

    International Nuclear Information System (INIS)

    Designing and optimising a reactor core is rather complex as it involves neutronics, thermal-hydraulics and thermomechanics. In order to tentatively overcome these difficulties, a new approach based on simplified models, is being developed aiming in optimising both core performance (core volume, in-cycle Pu inventory..) and core safety characteristics (neutronics coefficients, core pressure drop, transient response..) of a Fast Neutron Reactor. This new approach, called FARM (Fast Reactor Methodology) is currently used for studying a Helium-Cooled Fast Reactor core with carbide fuel pins, and a SiC-based CMC (Ceramic Matrix Composite) cladding. This method has demonstrated that, for a given initial set of specifications (thermal power, inlet coolant temperature, He pressure), 10 optimization variables are sufficient to estimate fair core design features. All simplified models are built from reference CEA codes (ERANOS for neutronics, METEOR for fuel thermomechanics) by way of polynomial interpolations derived from physical analytical considerations. Some safety aspects are also considered in the analysis using analytical descriptions (decay heat removal by natural convection, thermal inertia of the core, etc...). With a multi-criterion genetic algorithm, the 10 optimization variables are then searched for improving both neutronics and safety characteristics. This new methodology allows less accurate, but optimized, core design features to be obtained and proves they are the best that fulfil all the requirements. The first series of studies justify several safety trends already considered in the conventional method (minimisation of pressure drop). Current results confirm that such an approach is possible, and leads to new core designs, similar to the reference core, but with better performance (at least, supply pumping power reduced by 30%, for the same core performance). (authors)

  16. Multi-group nodal expansion method for reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byung Oh; Joo, Han Gyu; Park, Sang Yoon; Zee, Sung Quun; Kim, Ha Yong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    MASTER-2.0 is a nuclear design code based on the two group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. The response matrix based NEM has been extended for multi-group neutron diffusion theory in order to increase the computational accuracy for rectangular geometry. Coarse mesh rebalancing scheme is used to accelerate the convergence of iteration process. The transverse leakage profile involved in NEM is approximated by a parabola. Its coefficients are determined by using the continuity condition at interfaces or the intra-nodal flux shape including node vertices. For the verification of the multi-group NEM routine of MASTER-2.0, the combinations of the transverse leakage approximation with NEM were tested using two benchmark problems in order to check the sound operation of the routine. Comparisons made reveal that the accuracy of the NEM for the prediction of eigenvalue and power distribution is quite good and the four-group cross sections generated by CASMO-3 work properly in the MASTER code system. 11 refs., 7 figs., 4 tabs. (Author)

  17. A computer program to determine the specific power of prismatic-core reactors

    International Nuclear Information System (INIS)

    A computer program has been developed to determine the maximum specific power for prismatic-core reactors as a function of maximum allowable fuel temperature, core pressure drop, and coolant velocity. The prismatic-core reactors consist of hexagonally shaped fuel elements grouped together to form a cylindrically shaped core. A gas coolant flows axially through circular channels within the elements, and the fuel is dispersed within the solid element material either as a composite or in the form of coated pellets. Different coolant, fuel, coating, and element materials can be selected to represent different prismatic-core concepts. The computer program allows the user to divide the core into any arbitrary number of axial levels to account for different axial power shapes. An option in the program allows the automatic determination of the core height that results in the maximum specific power. The results of parametric specific power calculations using this program are presented for various reactor concepts

  18. Shape optimization of a Sodium Fast Reactor core

    Directory of Open Access Journals (Sweden)

    Dombre Emmanuel

    2013-01-01

    Full Text Available We apply in this paper a geometrical shape optimization method for the design of the core of a SFR (Sodium-cooled Fast Reactor in order to minimize a thermal counter-reaction known as the sodium void effect. In this kind of reactors, by increasing the temperature, the core may become liable to a strong increase of reactivity, a key-parameter governing the chain-reaction at quasi-static states. We first use the one group energy diffusion model and give the generalization to the two groups energy equation. We then give some numerical results in the case of the one group energy equation. Note that the application of our method leads to some designs whose interfaces can be parametrized by very smooth curves which can stand very far from realistic designs. We don’t explain here the method that it would be possible to use for recovering an operational design but there exists several penalization methods (see [2] that could be employed to this end. On applique dans cet article une méthode d’optimisation géométrique dans le cadre de la conception d’un cœur de réacteur SFR (Sodium-cooled Fast Reactor, i.e. réacteur à neutron rapide refroidi au sodium dans le but de minimiser une contre réaction thermique connue sous le nom d’effet de vidange sodium. Lorsqu’une augmentation de température survient, ce type de réacteur peut être sujet à une forte augmentation de réactivité, un paramètre clé dans le contrôle de la réaction en chaîne en régime quasi-statique. On a recours à l’équation de diffusion à un groupe puis on donne la généralisation du modèle d’optimisation pour l’équation de la diffusion à deux groupes d’énergie. On présente ensuite quelques résultats numériques obtenus dans le cas de l’équation à un groupe d’énergie. On note que l’application de cette méthode conduit à des designs de cœur présentant des interfaces très régulières qui sont loin d’un design de cœur faisable sur le

  19. Simple Model for Gas Holdup and Liquid Velocity of Annular Photocatalytic External-Loop Airlift Reactor Under both Bubble and Developing Slug Flow

    Institute of Scientific and Technical Information of China (English)

    王一平; 陈为强; 黄群武; 冯加和; 崔勇

    2016-01-01

    Based on the momentum conservation approach, a theoretical model was developed to predict the su-perficial liquid velocity, and a correlation equation was established to calculate the gas holdup of an annular exter-nal-loop airlift reactor(AELAR)in the bubble flow and developing slug flow pattern. Experiments were performed by using tap-water and silicone oil with the viscosity of 2.0 mm2/s(2cs-SiO)and 5.0 mm2/s(5cs-SiO)as liquid phases. The effects of liquid viscosity and flow pattern on the AELAR performance were investigated. The predic-tions of the proposed model were in good agreement with the experimental results of the AELAR. In addition, the comparison of the experimental results shows that the proposed model has good accuracy and could be used to pre-dict the gas holdup and liquid velocity of an AELAR operating in bubble and developing flow pattern.

  20. Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

    International Nuclear Information System (INIS)

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm2 sand length of cycle is 57 days is the most optimal of IRR. (author)

  1. Dynamic functional characterization and phylogenetic changes due to Long Chain Fatty Acids pulses in biogas reactors

    DEFF Research Database (Denmark)

    Kougias, Panagiotis; Treu, Laura; Campanaro, Stefano;

    2016-01-01

    the dynamics of the microbial community during an inhibitory shock load induced by single pulses of unsaturated LCFA at two different concentrations (i.e. 2 g/L-reactor and 3 g/L-reactor). The metagenomic analysis showed that only the microbes associated with LCFA degradation could encode proteins related...

  2. About a fuel for burnup reactor of periodical pulsed nuclear pumped laser

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, A.I.; Lukin, A.V.; Magda, L.E.; Magda, E.P.; Pogrebov, I.S.; Putnikov, I.S.; Khmelnitsky, D.V.; Scherbakov, A.P. [Russian Federal Nuclear Center, Snezhinsk (Russian Federation)

    1998-07-01

    A physical scheme of burnup reactor for a Periodic Pulsed Nuclear Pumped Laser was supposed. Calculations of its neutron physical parameters were made. The general layout and construction of basic elements of the reactor are discussed. The requirements for the fuel and fuel elements are established. (author)

  3. Heat transfer and core neutronics considerations of the heat pipe cooled thermionic reactor

    Science.gov (United States)

    Determan, W. R.; Lewis, Brian

    1991-01-01

    The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power conversion, heat transport, and heat rejection functions are all combined into a single modular unit. The reactor/converter assembly uses UN fuel pins to obtain a critical core configuration with in-core safety rods and reflector controls added to complete the subassembly. By thermally bonding the core heat transfer assemblies during the reactor core is coupled neutronically, thermally, and electrically into a modular assembly of individual power sources with cross-tied architecture. A forward-facing heat pipe radiator assembly extends from the reactor head in the shape of a frustum of a cone on the opposite side of the power system from the payload. Important virtues of the concept are the absence of any single-point failures and the ability of the core to effectively transfer the TFE waste heat load laterally to other in-core heat transfer assemblies in the event of multiple failures in either in-core and radiator heat pipes.

  4. Burnup concept for a long-life fast reactor core using MCNPX.

    Energy Technology Data Exchange (ETDEWEB)

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  5. Test and application of thermal neutron radiography facility at Xi'an pulsed reactor

    CERN Document Server

    Yang Jun; Zhao Xiang Feng; Wang Dao Hua

    2002-01-01

    A thermal neutron radiography facility at Xi'an Pulsed Reactor is described as well as its characteristics and application. The experiment results show the inherent unsharpness of BAS ND is 0.15 mm. The efficient thermal neutron n/gamma ratio is lower in not only steady state configuration but also pulsing state configuration and it is improved using Pb filter

  6. MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Yan; Gohar, Yousry

    2015-11-01

    In a periodic pulsed reactor, the reactor state varies periodically from slightly subcritical to slightly prompt supercritical for producing periodic power pulses. Such periodic state change is accomplished by a periodic movement of specific reactor parts, such as control rods or reflector sections. The analysis of such reactor is difficult to perform with the current reactor physics computer programs. Based on past experience, the utilization of the point kinetics approximations gives considerable errors in predicting the magnitude and the shape of the power pulse if the reactor has significantly different neutron life times in different zones. To accurately simulate the dynamics of this type of reactor, a Monte Carlo procedure using the transfer function TRCL/TR of the MCNP/MCNPX computer programs is utilized to model the movable reactor parts. In this paper, two algorithms simulating the geometry part movements during a neutron history tracking have been developed. Several test cases have been developed to evaluate these procedures. The numerical test cases have shown that the developed algorithms can be utilized to simulate the reactor dynamics with movable geometry parts.

  7. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    Science.gov (United States)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

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

  8. Evaluation of the shielding design around the reactor core in the prototype FBR Monju

    Energy Technology Data Exchange (ETDEWEB)

    Shiraki, Takako; Tada, Keiko [Advanced Reactor Technology Co., Ltd., Tokyo (Japan); Usami, Shin; Sasaki, Kenji [Japan Nuclear Cycle Development Institute, Tsuruga, Fukui (Japan); Tabayashi, Masao [Japan Atomic Power Co., Tokyo (Japan)

    2000-03-01

    This paper describes shielding evaluation of the measurements around the reactor core of Monju. The measurements were performed during the system start-up tests at different power levels between 0% and 45%. The measured reaction rates have been obtained radially from the core to the in-vessel storage rack and axially to the reactor vessel upper plenum. The measured values (E) were compared with the calculated values (C) obtained with the FBR shielding analysis methods. Based upon these results, the design margins around the reactor core have been re-examined and re-confirmed. (author)

  9. Evaluation of fluid effects on the dynamic response of a fast reactor core

    International Nuclear Information System (INIS)

    The results of dynamic experiments on shaking tables, carried out in water (simulating sodium) on both single and coupled core element prototypes and core simplified mock-up configurations of the Italian PEC fast reactor test facility, with excitation gradually increasing up to above Safe Shutdown Earthquake, have been analysed by use of the one-dimensional computer program CORALIE and the two-dimensional program CLASH. The study confirmed the conservative nature of the PEC core design calculations, provided the natural frequency and damping values to be used in the calculations for the Final Safety Report, and allowed the fluid-structure interaction model to be assessed for the PEC core seismic analysis. It also contributed to the validation of the above-mentioned computer codes for their general use for the fast reactor core analysis as well as to a better understanding of fluid-structure interaction problems concerning the fast reactor core

  10. Steady Thermal Field Simulation of Forced Air-cooled Column-type Air-core Reactor

    Institute of Scientific and Technical Information of China (English)

    DENG Qiu; LI Zhenbiao; YIN Xiaogen; YUAN Zhao

    2013-01-01

    Modeling the steady thermal field of the column-type air-core reactor,and further analyzing its distribution regularity,will help optimizing reactor design as well as improving its quality.The operation mechanism and inner insulation structure of a novel current limiting column-type air-core reactor is introduced in this paper.The finite element model of five encapsulation forced air-cooled column type air-core reactor is constructed using Fluent.Most importantly,this paper present a new method that,the steady thermal field of reactor working under forced air-cooled condition is simulated without arbitrarily defining the convection heat transfer coefficient for the initial condition; The result of the thermal field distribution shows that,the maximum steady temperature rise of forced air-cooled columntype air-core reactor happens approximately 5% to its top.The law of temperature distribution indicates:In the 1/3part of the reactor to its bottom,the temperature will rise rapidly to the increasing of height,yet the gradient rate is gradually decreasing; In the 5 % part of the reactor to its top,the temperature will drop rapidly to the increasing of height; In the part between,the temperature will rise slowly to the increasing of height.The conclusion draws that more thermal withstand capacity should be considered at the 5 % part of the reactor to its top to achieve optimal design solution.

  11. Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W. [Oak Ridge National Lab., TN (United States); Kenton, M.A. [Dames and Moore, Westmont, IL (United States)

    1996-09-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations.

  12. Enhancement of peak intensity in a filament core with spatiotemporally focused femtosecond laser pulses

    Energy Technology Data Exchange (ETDEWEB)

    Zeng Bin; Chu Wei; Li Guihua; Zhang Haisu; Ni Jielei [State Key Laboratory of High Field Laser Physics, Shanghai Institute of Optics and Fine Mechanics, Chinese Academy of Sciences, Shanghai 201800 (China); Graduate School of Chinese Academy of Sciences, Beijing 100080 (China); Gao Hui; Liu Weiwei [Institute of Modern Optics, Nankai University, Tianjin, 300071 (China); Yao Jinping; Cheng Ya; Xu Zhizhan [State Key Laboratory of High Field Laser Physics, Shanghai Institute of Optics and Fine Mechanics, Chinese Academy of Sciences, Shanghai 201800 (China); Chin, See Leang [Center for Optics, Photonics and Laser (COPL) and Department of Physics, Engineering Physics and Optics, Universite Laval, Quebec City, QC, G1V 0A6 (Canada)

    2011-12-15

    We demonstrate that the peak intensity in the filament core, which is inherently limited by the intensity clamping effect during femtosecond laser filamentation, can be significantly enhanced using spatiotemporally focused femtosecond laser pulses. In addition, the filament length obtained by spatiotemporally focused femtosecond laser pulses is {approx}25 times shorter than that obtained by a conventional focusing scheme, resulting in improved high spatial resolution.

  13. Thermohydraulic assessment of the RP-10 reactor core to determine the maximum power

    International Nuclear Information System (INIS)

    Thermohydraulic parameters assessment of the RP-10 reactor core from the most thermally demanded (hot channel). Determination of the operation thermal maximum power considering security margins and statistical treatment of uncertainty factors

  14. Comparison of three/four equations reactor core models in the Laguna Verde simulator

    International Nuclear Information System (INIS)

    This work presents results of the simulation of three transients in the full scope Laguna Verde nuclear power plant simulator. Three and four equations reactor core models were used, and simulation results are compared with manufacturer's predictions. (Author)

  15. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

    Directory of Open Access Journals (Sweden)

    Patrícia A. L. Reis

    2015-01-01

    Full Text Available Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature.

  16. Numerical Analysis of Magnetic Force of Dry-Type Air-Core Reactor

    Institute of Scientific and Technical Information of China (English)

    LIUZhi-gang; GENGYing-san; WANGJian-hua

    2004-01-01

    This paper presents a coupled magnetic-circuit method for computing the magnetic force of air-core reactor under short-time current. The current and the magnetic flux density are computed first and then the magnetic force is obtained. Thus, the dynamic stability performance of air-core reactor can be analyzed at the design stage to reduce experimental cost and shorten the lead-time of product development.

  17. 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)

    Aiming to assure the continued utilization of fission energy, the development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed, as the best type for the Brazilian nuclear system in the future. The inherent safety characteristics are superior to current fast breeder reactors and an efficient utilization of thorium can be realized. The construction and operation of an experimental fast reactor is the first step and a basic tool for the development of FBRs technologies. A serie of core design for an 90 MW FBR is studied and the possible options and sizes of the main parameters are identified. (E.G.)

  18. Development of Core Design Model for Small-Sized Research Reactor and Establishment of Infrastructure for Reactor Export

    International Nuclear Information System (INIS)

    Within 10 years a growing world-wide demand of new research reactor construction is expected because of obsolescence. In Korea, a new research reactor is also required in order to meet domestic demand of utilization. KAERI has been devoted to develop an export-oriented research reactors for these kinds of demand. A next generation research reactor should comply with general requirements for safety, economics, environment-friendliness and non-proliferation as well as high performance requirement of high flux level. A export-tailored reactor should be developed for the demand of developing counties or under-developed countries. A new design concept is to be developed for a long cycle length core which has excellent irradiation facility with high flux

  19. Influence of high dose irradiation on core structural and fuel materials in advanced reactors

    International Nuclear Information System (INIS)

    The IAEA International Working Group on Fast Reactors (IWGFR) periodically organizes meeting to discuss and review important aspects of fast reactor technology. The fifth meeting held in Obninsk, Russian Federation, 16-19 June 1997, was devoted to the influence of high dose irradiation on the mechanical properties of reactor core structural and fuel materials. The proceedings includes the papers submitted at this meeting each with a separate abstract

  20. Irradiation capabilities of LR-0 reactor with VVER-1000 Mock-Up core.

    Science.gov (United States)

    Košťál, Michal; Rypar, Vojtěch; Svadlenková, Marie; Cvachovec, František; Jánský, Bohumil; Milčák, Ján

    2013-12-01

    Even low power reactors, such as zero power reactors, are sufficient for semiconductor radiation hardness effect investigation. This reflects the fact that fluxes necessary for affecting semiconductor electrical resistance are much lower than fluxes necessary to affect material parameters. The paper aims to describe the irradiation possibilities of the LR-0 reactor with a special core arrangement corresponding to VVER-1000 dosimetry Mock-Up.

  1. Core monitoring and surveillance of VVER-440 type reactors in the Czech Republic and Slovak Republic

    International Nuclear Information System (INIS)

    The SCORPIO-VVER reactor core monitoring system is an advanced redundant software system without actuating members falling in the BT3 class which has been installed at the four Dukovany reactor units and at two units of the Slovak Jaslovske Bohunice V2 NPP. The system is described in detail and its history and experience gained at Dukovany are highlighted. (orig.)

  2. Optimization of a Potential New Core of the TRIGA Mark II Reactor Vienna

    International Nuclear Information System (INIS)

    The TRIGA Mark II Vienna is one of the last TRIGA reactors utilizing a mixed core with High Enrich Uranium (HEU) fuel. Due to the US Fuel Return Program, the Vienna University of Technology/Atominstitut (ATI) is obliged to return its HEU fuel by 2019. There is no final decision on any further utilization of the Vienna research reactor beyond that point. However, of all possible scenarios of the future, the conversion of the current core into Low Enriched Uranium (LEU) fuel and the complete replacement of all existing 83 burned FE(s) by new fresh FE(s) are investigated herein. This paper presents detailed reactor design calculations for three different reactor cores. The core 1 employs 104-type, core 2 uses 108-type and core 3 is loaded with mixed TRIGA fuels (i.e. 104 and 108). The combination of the Monte Carlo based neutronics code MCNP5, Oak Ridge Isotope Generation and depletion code ORIGEN2 and diffusion theory based reactor physics program TRIGLAV is used for this study. On the basis of this neutronics study, the amount of fuel required for a possible future reactor operation and its cost minimization is presented in this paper. The criticality, core excess reactivity, length of initial life cycle and thermal flux density distribution is simulated for three different cores. Keeping the utilization of existing fourteen 104-type FE(s) (i.e. six burned and eight fresh FE(s)) in view, the core 3 is found the most economical, enduring and safe option for future of the TRIGA Mark II reactor in Vienna. (author)

  3. RESEARCH OF RATIONAL LENGTH OF CORE SOIL USED IN ANNULAR EXCAVATION METHOD%隧道环形开挖时核心土合理长度研究

    Institute of Scientific and Technical Information of China (English)

    周路军; 叶剑锋; 尚岳全

    2011-01-01

    为提高预留核心土工法的科学应用,对预留核心土环形开挖法中核心土合理长度问题进行研究.通过有限元模拟,分析了核心土长度变化对围岩塑性区、掌子面纵向位移及拱顶沉降的影响.%In order to improve the application of reserving core soil construction method, rational length of core soil used in reserving core soil with annular excavation method is researched. By means of finite element method, the influence of length of core soil to the plastic area in surrounding rock, longitudinal displacement of tunnel face and vault settlement are analyzed. Research shows that rational length of core soil is 2. 5 ~3.5m while surrounding rock is ranked Ⅵ.

  4. A concept of prospective sodium fast reactor with ductless fuel subassemblies in the core

    Energy Technology Data Exchange (ETDEWEB)

    Sedov, A.A.; Alekseev, P.N.; Fomichenko, P.A.; Ponomarev-Stepnoy, N.N.; Proshkin, A.A.; Ponomarev, A.S.; Stukalov, V.A. [Russian Research Center, Kurchatov Institute, Moscow (Russian Federation)

    2007-07-01

    The Kurchatov Institute studies the concept of a sodium fast reactor (SFR) with advanced core design, which is based on the following principle technique solutions: -) application of ductless fuel subassemblies with wide lattice of fuel rods of increased diameter and spaced by grids; -) the usage of dense U-Pu ceramic fuel and low-nickel steels, and -) application of cluster-type control and protection system. Preconceptual studies have shown, that SFR with advanced core design is 3 times more effective in the fuel consumption than project BN-800 reactor due to better neutron balance in the core and CBR (core breeding ratio) {approx} 1, provides getting quite high burn-up of the core fuel (Bmax {approx} 15-20 % of heavy atoms), increases fuel life up to 7-8 years at specific loading of fissile nuclides in the core less than 5 t/GW, decreases electricity demand for pumping the primary coolant (due to low hydraulic resistance of the core) and has bigger safety potential in accidents than the core with traditional liquid metal fast reactor design (due to low core reactivity margin, high level of natural circulation and subassemblies hydraulic interaction). In the paper the main results of preconceptual feasibility study of SFR with advanced core design are presented and discussed with a focus on technique and economic aspects. Some of characteristic features of core neutron physics, thermal hydraulics and fuel rod thermal mechanics behavior are displayed and discussed as well. (authors)

  5. Analysis of fuel management pattern of research reactor core of the MTR type design

    International Nuclear Information System (INIS)

    Research reactor core design needs neutronics parameter calculation use computer codes. Research reactor MTR type is very interested because can be used as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thermal neutron flux in the core is 1.0x1015 n/cm2s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of high and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronics parameter calculation is done for new U-9Mo-Al fuel with variation of densities.The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0 x 1015 n/cm2s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. (author)

  6. Sampling of reactor pressure vessel and core internals

    International Nuclear Information System (INIS)

    Decommissioning and dismantling of nuclear power plants is a growing business as a huge number of plants built in the 1970's have now reached their lifetime. It is well known that dismantling a nuclear power plant means an extraordinary expense for the owner respectively operator. Beside the dismantling works for itself, the disposal of activated components and other nuclear waste is very expensive. What comes next is the fact that final disposal facilities are not available yet in most countries meaning a need for interim storage on-site in specially built facilities. It can be concluded that a special attention is paid on producing a minimal radioactive waste volume. For this, optimized dismantling and packaging concepts have to be developed. AREVA is proud of versatile experience in successfully dismantling nuclear components like core internals and reactor pressure vessel (RPV). The basis of a well-founded and optimized dismantling and packaging concept must always be the detailed knowledge of the radiological condition of the component to be and in the best case a 3D activation- model. For keeping the necessary sampling effort as small as possible, but simultaneously as efficient as possible, representative sampling positions are defined in advance by theoretical radiological examinations. For this, a detailed 3D-CAD-model of the components to be dismantled has proven very helpful and effective. Under these aspects a sampling of RPV and its components is necessary to verify the theoretically calculated radiological data. The obtained results of activation and contamination are taken into account for the optimized dismantling and packaging strategy. The precise 3D-activation-model will reduce the necessary number and type of final disposal containers as security factors are minimized leading to a lower shielding effort, too. Besides, components or even parts of components may be subject of release measurement. In the end, costs can be reduced. In this context

  7. 100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT

    Energy Technology Data Exchange (ETDEWEB)

    HARRINGTON RA

    2010-01-15

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

  8. Study on Mechanism of Drag Reduction by Core-annular Flow in Transportation of Emulsion Matrix%乳胶基质水环输送的机理研究

    Institute of Scientific and Technical Information of China (English)

    杨佳; 刘寿康

    2012-01-01

    对目前广泛应用于乳化炸药混装车上的水环输送乳胶基质的减阻机理进行了理论分析,分别推导出层流同心水环、湍流同心水环的速度分布与流量计算公式,并得到了从层流到湍流的转捩判据.此外,优化了水环润滑装置结构参数,提出了稳定水环输送的相关措施.%The mechanism of drag reduction by core-annular flow in transportation of emulsion matrix was widely used in the existing mixing-loading truck for emulsion explosive. After theoretical analysis, formulas were deduced respectively to calculate the velocity distribution, flow rate of annular water flow in laminar and turbulent flows. The criterion of transition from laminar flow to turbulent flow was obtained. Furthermore, the optimized parameters and some stabilizing measures are also proposed for core-annular flow device.

  9. On 135Xe poisoning in the core of a thermal reactor with circulating fuel

    International Nuclear Information System (INIS)

    The derivation of simple analytical expressions for estimating 135Xe poisoning in quasistationary state of the reactor with circulating fuel in the primary circuit. It is shown that 135Xe poisoning in such reactors depends on the ratio of the time during which fuel stays inside the core to the time outside the core (t1/t2).Even at ratio t1/t2=0.1, xenon poisoning effect can the reduced by six times compared to the reactor with fixed fuel, which essentially increases fuel use efficiency

  10. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  11. Seismic responses of N-Reactor core. Independent review of Phase II work

    International Nuclear Information System (INIS)

    Seismic response of the N-Reactor core was independently analyzed to validate the results of Impell's analysis. The analysis procedure consists of two major stages: linear soil-structure interaction (SSI) analysis of the overall N-Reactor structure complex and nonlinear dynamic analysis of the reactor core. In the SSI analysis, CLASSI computer codes were used to calculate the SSI response of the structures and to generate the input motions for the nonlinear reactor core analysis. In addition, the response was compared to the response from the SASSI analysis under review. The impact of foundation modeling techniques and the effect of soil stiffness variation on SSI response were also investigated. In the core analysis, a nonlinear dynamic analysis model was developed. The stiffness representation of the model was calculated through a finite element analysis of several local core geometries. Finite element analyses were also used to study the block to block interaction characteristics. Using this nonlinear dynamic model along with the basemat time histories generated from CLASSI and SASSI, several dynamic analyses of the core were performed. A series of sensitivity studies was performed to investigate the discretization of the core, the effect of vertical acceleration, the effect of basemat rocking, and modeling assumptions. In general, our independent analysis of core response validates the order of magnitude of the displacement calculated by Impell. 11 refs., 110 figs., 12 tabs

  12. Alternative core design for the Innovative Research Reactor (RRI) from neutronics aspects

    International Nuclear Information System (INIS)

    Based on its User Requirement Document and main function, RRI shall be able to provide a maximum thermal neutron flux of 1×1015 neutron cm-2s-1. The reason is that the RRI reactor can serve targets requiring a high neutron flux. From the previous results it was obtained that RRI design using fuel of RSG-GAS type was not possible to produce that high neutron flux. One among other reasons is that the geometry dimension is the large, as the neutron flux is inversely proportional to core volume. The objective of the study is to find an alternative core for RRI which meets the high neutron flux requirement. It was chosen an alternative fuel element one like used in JMTR (Japan Material Testing Reactor) that has smaller dimension compared to that of the RSG-GAS reactor. Besides that, active core's height was also varied for 70 cm and 75 cm. Design was carried out by means of analytic codes WIMS-D5B, Batan-FUEL and Batan-3DIFF. Alternative core applied compact core configuration concept of 5×5 with 4 follower control elements. The calculations resulted 3 (three) alternative cores fulfill the requirement, including core using RSG-GAS fuel type but of 70 cm height instead of 60 cm. Through analyzing from over all aspects of core safety and efficiency as well as effectively, core using JMTR fuel type with height of 70 cm represent the best alternative core. (author)

  13. a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

    Science.gov (United States)

    Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

    2009-08-01

    This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

  14. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff...)-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors.''...

  15. Method of measuring instant negative temperature coefficient of pulsed reactor by noise techniques

    International Nuclear Information System (INIS)

    Based on the relationship of neutron noise and temperature noise in reactors, a physical model which will be used to calculate the instant negative temperature coefficient (αF) of pulsed reactor is established in frequency domain by noise techniques. The neutron dynamic equation and thermal-dynamic equation were used while constructing the physical model. According to the disturbance in formation of neutron signal and temperature signal in the stable operation situation of reactors, the power spectrum densities are get by auto-regress moving average model. The αF of the pulsed reactor is obtained by best fitting method in the frequency domain. And the results are relative to the theory values

  16. The Chlorella killed by pulsed electrical discharge in liquid with two different reactors

    Science.gov (United States)

    Gao, Z. Y.; Sun, B.; Yan, Z. Y.; Zhu, X. M.; Liu, H.; Song, Y. J.; Sato, M.

    2013-03-01

    The application of pulsed high-voltage discharge in liquid has attracted wide attention as an effective water treatment. In this paper, two different liquid high-voltage discharge systems were constructed with plate-hole-plate and needle-plate electrode structures, and the inactivation behaviors of Chlorella were studied in the two reactors. The results show that the killing rates of algae in both reactors all increased significantly with increasing discharge voltage and the killing rates were intensely related to discharge power, instantaneous power and single pulse input energy. Furthermore, the inactivation effect in needle-plate reactor was superior to that in plate-hole-plate reactor under the same experimental conditions.

  17. Conceptual design for the superconducting magnet system of a pulsed DEMO reactor

    International Nuclear Information System (INIS)

    Highlights: ► A 1D design approach of a pulsed DEMO reactor is presented. ► The main CS and TF conductor design criteria are presented. ► A typical major radius for a 2 GW DEMO is 9 m. ► A typical plasma magnetic field is 4.9 T. ► The pulse duration is 1.85 h for an aspect ratio of 3. -- Abstract: A methodology has been developed to consistently investigate, taking into account main reactor components, possible magnet solutions for a pulsed fusion reactor aiming at a large solenoid flux swing duration within the 2–3 h range. In a conceptual approach, investigations are carried out in the equatorial plane, taking into account the radial extension of the blanket-shielding zone, of the toroidal field magnet system inner leg and of the central solenoid for estimation of the pulsed swing. Design criteria are presented for the radial extension of the superconducting magnets, which is mostly driven by the structures (casings and conductor jacket). Typical available cable current densities are presented as a function of the magnetic field and of the temperature margin. The magnet design criteria have been integrated into SYCOMORE, a code for reactor modeling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modeling task force. Possible solutions are investigated for a 2 GW fusion power reactor with different aspect ratios. The final adjustment of the DEMO pulsed reactor parameters will have to be consistently done, considering all reactor components, when the final goals of the machine will be completely clarified

  18. Conceptual design for the superconducting magnet system of a pulsed DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Duchateau, J.-L., E-mail: jean-luc.duchateau@cea.fr [CEA/IRFM, 13108 St. Paul lez Durance Cedex (France); Hertout, P.; Saoutic, B.; Magaud, P.; Artaud, J.-F.; Giruzzi, G.; Bucalossi, J.; Johner, J.; Sardain, P.; Imbeaux, F.; Ané, J.-M.; Li-Puma, A. [CEA/IRFM, 13108 St. Paul lez Durance Cedex (France)

    2013-10-15

    Highlights: ► A 1D design approach of a pulsed DEMO reactor is presented. ► The main CS and TF conductor design criteria are presented. ► A typical major radius for a 2 GW DEMO is 9 m. ► A typical plasma magnetic field is 4.9 T. ► The pulse duration is 1.85 h for an aspect ratio of 3. -- Abstract: A methodology has been developed to consistently investigate, taking into account main reactor components, possible magnet solutions for a pulsed fusion reactor aiming at a large solenoid flux swing duration within the 2–3 h range. In a conceptual approach, investigations are carried out in the equatorial plane, taking into account the radial extension of the blanket-shielding zone, of the toroidal field magnet system inner leg and of the central solenoid for estimation of the pulsed swing. Design criteria are presented for the radial extension of the superconducting magnets, which is mostly driven by the structures (casings and conductor jacket). Typical available cable current densities are presented as a function of the magnetic field and of the temperature margin. The magnet design criteria have been integrated into SYCOMORE, a code for reactor modeling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modeling task force. Possible solutions are investigated for a 2 GW fusion power reactor with different aspect ratios. The final adjustment of the DEMO pulsed reactor parameters will have to be consistently done, considering all reactor components, when the final goals of the machine will be completely clarified.

  19. Demonstration of core neutronic calculation for research and training reactors via SCALE4.4

    International Nuclear Information System (INIS)

    In this work, full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 system. KENOV.a module of SCALE4.4 system is utilized for full core neutronic analysis. The ORIGEN-S module is also coupled with the KENOV.a module to perform burnup dependent core analyses. Results of control rod worths for 1st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. (author)

  20. Analysis of thorium/U-233 lattices and cores in a breeder/burner heavy water reactor

    International Nuclear Information System (INIS)

    Due to the inevitable dwindling of uranium resources, advanced fuel cycles in the current generation of reactors stand to be of great benefit in the future. Heavy water moderated reactors have much potential to make use of thorium, a currently unexploited resource. Core fuelling configurations of a Heavy Water Reactor based on the self-sufficient thorium fuel cycle were simulated using the DRAGON and DONJON reactor physics codes. Three heterogeneously fuelled reactors and one homogeneously fuelled reactor were studied. (author)

  1. Condensed matter and materials research using neutron diffraction and spectroscopy: reactor and pulsed neutron sources

    International Nuclear Information System (INIS)

    The paper provides a short, and partial view of the neutron scattering technique applied to condensed matter and materials research. Reactor and accelerator-based neutron spectrometers are discussed, together with examples of research projects that illustrate the puissance and modern applications of neutron scattering. Some examples are chosen to show the range of facilities available at the medium flux reactor operated by Casaccia ENEA, Roma and the advanced, pulsed spallation neutron source at the Rutherford Appleton Laboratory, Oxfordshire. (author)

  2. Comparative study of pulsed and steady-state tokamak reactor burn cycles

    International Nuclear Information System (INIS)

    Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles

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

    International Nuclear Information System (INIS)

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

  4. Simulation of power pulses during large break LOCAs in natural and slightly enriched cores in the Embalse NPP

    International Nuclear Information System (INIS)

    In the frame of a joint technical feasibility study between Nucleoelectrica Argentina and Atomic Energy of Canada of using slightly enriched uranium fuel (with 0.9 w% U235) in Embalse NPP, a CANDU-6, loss of coolant accidents (LOCAs) simulations were performed. The power pulse due to two large breaks were simulated: 35% of a Reactor Inlet Header (RIH) and 80% of a Reactor Outlet Header (ROH). For each break size four simulations were performed for different initial conditions o scenarios and for Natural Uranium (NU) and slightly enriched uranium (SEU) cores. The power transients have been simulated using the 3D diffusion, spatial kinetics neutronic program PUMA (developed in Argentina) and the thermal-hydraulics program CATHENA. These codes were coupled by an iterative methodology. The CATHENA thermal-hydraulic simulation results (fuel temperatures and coolant temperatures and densities) were used as input of the PUMA calculation and the time dependent power distribution calculated by PUMA was later applied as input for a new CATHENA calculation. The process was repeated up to convergence. Single channel models were developed to calculate the relevant three key safety parameters: the maximum transient fuel centerline temperature, the maximum transient sheath temperature and the maximum transient stored energy. The main results of power pulse calculation show that the behavior of the SEU core are similar to the NU one. The result of the three safety parameter values show that in the hypothetical large break LOCA occurrence the fuel channel integrity is maintained. The maximum fuel temperature values are lower than the melting temperature of UO2 , the maximum stored enthalpies are lower than the fuel break-up limit and the maximum sheath temperature are lower than Zircalloy fusion temperature. The values of these safety parameters are similar or slightly lower for the SEU core compared with the NU one. (author)

  5. Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

    Science.gov (United States)

    Cormon, S.; Fallot, M.; Bui, V.-M.; Cucoanes, A.; Estienne, M.; Lenoir, M.; Onillon, A.; Shiba, T.; Yermia, F.; Zakari-Issoufou, A.-A.

    2014-06-01

    This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (νbare) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of 235U, 239Pu and 241Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.

  6. Core power distribution methodology in the BEACON PWR [pressurized water reactor] core monitoring system

    International Nuclear Information System (INIS)

    Westinghouse has developed an advanced operational core support package called BEACON which uses a fully analytical methodology for on-line prediction of 3-D [three-dimensional] power distributions. The system provides core monitoring, core measurement reduction, core analysis and follow, and core predictions. The heart of the system is a very fast and accurate three dimensional nodal code which is used for core simulation and predictions. The system uses a new methodology with the existing core instrumentation to infer the current measured power distribution. This methodology has been qualified and yields excellent results

  7. The reactor core TRIGA Mark-III with fuels type 30/20; El nucleo del reactor TRIGA Mark-III con combustible tipo 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F., E-mail: fortunato.aguilar@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    This work describes the calculation series carried out with the program MCNP5 in order to define the configuration of the reactor core with fuels 30/20 (fuels with 30% of uranium content in the Or-Zr-H mixture and a nominal enrichment of 20%). To select the configuration of the reactor core more appropriate to the necessities and future uses of the reactor, the following criterions were taken into account: a) the excess in the reactor reactivity, b) the switch out margin and c) to have new irradiation facilities inside the reactor core. Taking into account these criterions is proceeded to know the characteristics of the components that form the reactor core (dimensions, geometry, materials, densities and positions), was elaborated a base model of the reactor core, for the MCNP5 code, with a configuration composed by 85 fuel elements, 4 control bars and the corresponding structural elements. The high reactivity excess obtained with this model, gave the rule to realize other models of the reactor core in which the reactivity excess and the switch out margin were approximate to the values established in the technical specifications of the reactor operation. Several models were realized until finding the satisfactory model; this is composite for 74 fuels, 4 control bars and 6 additional experimental positions inside the reactor core. (Author)

  8. Subcutaneous granuloma annulare: radiologic appearance

    International Nuclear Information System (INIS)

    Objective. Granuloma annulare is an uncommon benign inflammatory dermatosis characterized by the formation of dermal papules with a tendency to form rings. There are several clinically distinct forms. The subcutaneous form is the most frequently encountered by radiologists, with the lesion presenting as a superficial mass. There are only a few scattered reports of the imaging appearance of this entity in the literature. We report the radiologic appearance of five cases of subcutaneous granuloma annulare. Design and patients. The radiologic images of five patients (three male, two female) with subcutaneous granuloma annulare were retrospectively studied. Mean patient age was 6.4 years (range, 2-13 years). The lesions occurred in the lower leg (two), foot, forearm, and hand. MR images were available for all lesions, gadolinium-enhanced imaging in three cases, radiographs in four, and bone scintigraphy in one. Results. Radiographs showed unmineralized nodular masses localized to the subcutaneous adipose tissue. The size range, in greatest dimension on imaging studies, was 1-4 cm. MR images show a mass with relatively decreased signal intensity on all pulse sequences, with variable but generally relatively well defined margins. There was extensive diffuse enhancement following gadolinium administration. Conclusion. The radiologic appearance of subcutaneous granuloma annulare is characteristic, typically demonstrating a nodular soft-tissue mass involving the subcutaneous adipose tissue. MR images show a mass with relatively decreased signal intensity on all pulse sequences and variable but generally well defined margins. There is extensive diffuse enhancement following gadolinium administration. Radiographs show a soft-tissue mass or soft-tissue swelling without evidence of bone involvement or mineralization. This radiologic appearance in a young individual is highly suggestive of subcutaneous granuloma annulare. (orig.)

  9. Criteria for structural verification of fast reactor core elements

    International Nuclear Information System (INIS)

    Structural and functional criteria and relative verifications of PEC reactor fuel element are presented and discussed. Particular attention has been given to differentiate the structural verifications of low neutronic damage zones from those high neutronic damage ones. The structural verification criteria, which had already been presented at the 8th SMIRT Seminar Conference in Paris, have had some modifications during the Safety Report preparation. Finally some necessary activities are indicated for structural criteria validation, in particular for irradiated components, and for converging towards a European fast reactor code. (author). 3 refs, 6 tabs

  10. PRODUC program package for calculating correlation relations in reactor core

    International Nuclear Information System (INIS)

    To perform calculations of fission product accumulation and radionuclide activity ratio distribution in the reactor fuel assembly (FA), the PRODUC software is developed. This package allows one to obtain distributions of radionuclide activity ratios for any fuel loading of the RBMK-1000 reactor. Plutonium and cerium-144 activity ratio distribution in the FA of the fuel loading of the 4th unit of the Chernobyl NPP as of April 25, 1986 is obtained according to the PRODUC program. 6 refs.; 7 figs.; 1 tab

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

    International Nuclear Information System (INIS)

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

  12. Improvement of pulsing operation performance in the Nuclear Safety Research Reactor (NSRR)

    International Nuclear Information System (INIS)

    The Nuclear Safety Research Reactor (NSRR) is one of the TRIGA-type research reactors widely used in the world, and has mainly been used for studying reactor fuel behaviour during postulated reactivity-initiated accidents (RIAs). Its limited pulsing operation capability, however, could produce only a power burst from low power level simulating an RIA event from essentially zero power level. A computerized automatic reactor control system was developed and installed in the NSRR to simulate a wide range of abnormal events in nuclear power plants. This digitalized reactor control system requires no manipulation of the control rods by reactor operators during the course of the pulsing operation. Using this fully automated operation system, a variety of power transients such as power ramping, power bursts from high power level, and so on were made possible with excellent stability and safety. The present modification work in the NSRR and its fruitful results indicate new possibilities in the utilization of the TRIGA type research reactor

  13. Reactor dynamics and stability analysis of a burst-mode gas core reactor, Brayton cycle space power system

    International Nuclear Information System (INIS)

    Reactor dynamics and system stability studies are performed on a conceptual burst-mode gaseous core reactor space nuclear power system. This concept operates on a closed Brayton cycle in the burst mode (on the order of 100-MW output for a few thousand seconds) using a disk magnetohydrodynamic generator for energy conversion. The fuel is a gaseous mixture of UF4 or UF6 and helium. Nonlinear dynamic analysis is performed using circulating-fuel, point-reactor-kinetics equations along with thermodynamic, lumped-parameter heat transfer and one-dimensional isentropic flow equations. The gaseous nature of the fuel plus the fact that the fuel is circulating lead to dynamic behavior that is quite different from that of conventional solid-core systems. For the transients examined, Doppler fuel temperature and moderator temperature feedbacks are insignificant when compared with reactivity feedback associated with fuel gas density variations. The gaseous fuel density power coefficient of reactivity is capable of rapidly stabilizing the system, within a few seconds, even when large positive reactivity insertions are imposed; however, because of the strength of this feedback, standard external reactivity insertions alone are inadequate to bring about significant power level changes during normal reactor operation. Additional methods of reactivity control, such as changes in the gaseous of fuel mass flow rate or core inlet pressure, are required to achieve desired power level control. Finally, linear stability analysis gives results that are qualitatively in agreement with the nonlinear analysis

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-22

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

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

    Science.gov (United States)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

  17. Restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    The experimental fast reactor Joyo is the first sodium cooled fast reactor in Japan. Joyo attained initial criticality as a breeder core in April 1977 and has operated as a high performance irradiation test bed since 2003. The 15th periodic inspection of Joyo commenced in May 2007 with the Fuel Handling Machine (FHM) being set up on the Rotating Plug (R/P) for refueling in June. When the R/P was taken down, measuring the load of the Hold-Down Shaft (HDS) revealed an abnormal decrease above the in-vessel storage rack (IVS). The HDS is a cylindrical FMH device that holds down the 6 surrounding subassemblies (S/As) which are adjacent to a withdrawn S/A. In order to investigate the cause of this, an in-vessel observation was conducted using a radiation-resistant fiber scope (RRF). As a result of the observations, it was discovered that the top of the irradiation test S/A 'MARICO-2' (the material testing rig with temperature control) had bent onto the IVS as an obstacle, and had damaged the Upper Core Structure (UCS). During the investigation of this incident, the in-vessel observations using RRF etc. took place at (1) the top of the S/As and the IVS for foreign material, (2) the bottom face of the UCS for damage under the condition with the level of sodium at -50 mm below the top of the S/As. In-vessel observation techniques for a Sodium cooled Fast Reactor (SFR) are important in confirming its safety and integrity. Since an in-vessel observation for an SFR has to be conducted under severe conditions that include high temperatures (∼ 200 deg-C) and high radiation doses (∼ 400 Gy/h), and the primary sodium coolant has to be retained in the Reactor Vessel (R/V) to remove the decay heat, an in-vessel observation equipment has to be designed to not only tolerate the severe conditions but also be capable of being inserted into the sealed R/V through the fixed holes built in to the R/P and gain access to the observation areas. The in-vessel observations were successfully

  18. Nuclear reactor pulse tracing using a CdZnTe electro-optic radiation detector

    Science.gov (United States)

    Nelson, Kyle A.; Geuther, Jeffrey A.; Neihart, James L.; Riedel, Todd A.; Rojeski, Ronald A.; Ugorowski, Philip B.; McGregor, Douglas S.

    2012-07-01

    CdZnTe has previously been shown to operate as an electro-optic radiation detector by utilizing the Pockels effect to measure steady-state nuclear reactor power levels. In the present work, the detector response to reactor power excursion experiments was investigated. Peak power levels during an excursion were predicted to be between 965 MW and 1009 MW using the Fuchs-Nordheim and Fuchs-Hansen models and confirmed with experimental data from the Kansas State University TRIGA Mark II nuclear reactor. The experimental arrangement of the Pockels cell detector includes collimated laser light passing through a transparent birefringent crystal, located between crossed polarizers, and focused upon a photodiode. The birefringent crystal, CdZnTe in this case, is placed in a neutron beam emanating from a nuclear reactor beam port. After obtaining the voltage-dependent Pockels characteristic response curve with a photodiode, neutron measurements were conducted from reactor pulses with the Pockels cell set at the 1/4 and 3/4 wave bias voltages. The detector responses to nuclear reactor pulses were recorded in real-time using data logging electronics, each showing a sharp increase in photodiode current for the 1/4 wave bias, and a sharp decrease in photodiode current for the 3/4 wave bias. The polarizers were readjusted to equal angles in which the maximum light transmission occurred at 0 V bias, thereby, inverting the detector response to reactor pulses. A high sample rate oscilloscope was also used to more accurately measure the FWHM of the pulse from the electro-optic detector, 64 ms, and is compared to the experimentally obtained FWHM of 16.0 ms obtained with the 10B-lined counter.

  19. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    International Nuclear Information System (INIS)

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% 235U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  20. Operation and Licensing of Mixed Cores in Water Cooled Reactors

    International Nuclear Information System (INIS)

    Nuclear fuel is a highly complex material that is subject to continuous development and is produced by a range of manufacturers. During operation of a nuclear power plant, the nuclear fuel is subject to extreme conditions of temperature, corroding environment and irradiation, and many different designs of fuel have been manufactured with differing fuel materials, cladding materials and assembly structure to ensure these conditions. The core of an operating power plant can contain hundreds of fuel assemblies, and where there is more than a single design of a fuel assembly in the core, whether through a change of fuel vendor, introduction of an improved design or for some other reason, the core is described as a mixed core. The task of ensuring that the different assembly types do not interact in a harmful manner, causing, for example, differing flow resistance resulting in under cooling, is an important part of ensuring nuclear safety. This report has compiled the latest information on the operational experience of mixed cores and the tools and techniques that are used to analyse the core operation and demonstrate that there are no safety related problems with its operation. This publication is a result of a technical meeting in 2011 and a series of consultants meetings

  1. A compact, high-voltage pulsed charging system based on an air-core pulse transformer.

    Science.gov (United States)

    Zhang, Tianyang; Chen, Dongqun; Liu, Jinliang; Liu, Chebo; Yin, Yi

    2015-09-01

    Charging systems of pulsed power generators on mobile platforms are expected to be compact and provide high pulsed power, high voltage output, and high repetition rate. In this paper, a high-voltage pulsed charging system with the aforementioned characteristics is introduced, which can be applied to charge a high-voltage load capacitor. The operating principle of the system and the technical details of the components in the system are described in this paper. The experimental results show that a 600 nF load capacitor can be charged to 60 kV at 10 Hz by the high-voltage pulsed charging system for a burst of 0.5 s. The weight and volume of the system are 60 kg and 600 × 500 × 380 mm(3), respectively. PMID:26429466

  2. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  3. A study on reactor core failure thresholds to safety operation of LMFBR

    International Nuclear Information System (INIS)

    Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

  4. Physical start up of the Dalat nuclear research reactor with the core configuration having a central neutron trap

    International Nuclear Information System (INIS)

    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

  5. Application of neutron activation analysis system in Xi'an pulsed reactor

    CERN Document Server

    Zhang Wen Shou; Yu Qi

    2002-01-01

    Neutron Activation Analysis System in Xi'an Pulsed Reactor is consist of rabbit fast radiation system and experiment measurement system. The functions of neutron activation analysis are introduced. Based on the radiation system. A set of automatic data handling and experiment simulating system are built. The reliability of data handling and experiment simulating system had been verified by experiment

  6. Experimental study of prompt neutron decay constant α for 300# pool reactor under mixed core

    International Nuclear Information System (INIS)

    The experimental study of prompt neutron decay constant α for 300# pool reactor under mixed core was carried out through a suit of reactor power spectral density measurement system. The two channel continuous current signals of neutron in the reactor were acquired by ionization chamber DL129 which was symmetrically putted in reactor core. The power spectral density, for two channel signals, was computed using the application program of data acquirement and data process analysis. Finally, by using the non-linear least squares method, the prompt neutron decay constant α was fitted. By comparison, the experimental results well accord to the theory calculation within the error range. The deviation can meet the actual need of project. (authors)

  7. Reactivity changes in hybrid thermal-fast reactor systems during fast core flooding

    International Nuclear Information System (INIS)

    A new space-dependent kinetic model in adiabatic approximation with local feedback reactivity parameters for reactivity determination in the coupled systems is proposed in this thesis. It is applied in the accident calculation of the 'HERBE' fast-thermal reactor system and compared to usual point kinetics model with core-averaged parameters. Advantages of the new model - more realistic picture of the reactor kinetics and dynamics during local large reactivity perturbation, under the same heat transfer conditions, are underlined. Calculated reactivity parameters of the new model are verified in the experiments performed at the 'HERBE' coupled core. The model has shown that the 'HERBE' safety system can shutdown reactor safely and fast even in the case of highly set power trip and even under conditions of big partial failure of the reactor safety system (author)

  8. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  9. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  10. Determination of the neutron fluence in the welding of the 'Core shroud' of the BWR reactor core

    International Nuclear Information System (INIS)

    With the purpose of defining the inspection frequency, in function of the embrittlement of the materials that compose the welding of the 'Core Shroud' or encircling of the core of a BWR type reactor, is necessary to know the neutron fluence received for this welding. In the work the calculated values of neutron fluence accumulated maxim (E > 1 MeV) during the first 8 operation cycles of the reactor are presented. The calculations were carried out according to the NRC Regulatory Guide 1.190, making use of the DORT code, which solves the transport equation in discreet ordinate in two dimensions (xy, rΘ, and rz). The results in 3D were obtained applying the Synthesis method according to the guide before mentioned. Results are presented for the horizontal welding H3, H4, and H5, showing the corresponding curves to the fluence accumulated to the cycle 8 and a projection for the cycle 14 is presented. (Author)

  11. Decomposition of dimethyl sulfide in a wire-cylinder pulse corona reactor

    Institute of Scientific and Technical Information of China (English)

    Jian-tao YANG; Yao SHI; Jie CHEN; Qing-fa SU; Da-hui WANG; Jing CAO

    2009-01-01

    Decomposition of dimethyl sulfide (DMS) in air was investigated experimentally by using a wire-cylinder dielectric barrier discharge (DBD) reactor at room temperature and atmospheric pressure. A new type of high pulse voltage source with a thyratron switch and a Blumlein pulse-forming network (BPFN) was adopted in our experiments. The maximum power output of the pulse voltage source and the maximum peak voltage were 1 kW and 100 kV, respectively. The important parameters affecting odor decomposition, including peak voltage, pulse frequency, gas flow rate, initial concentration, and humidity, which influenced the removal efficiency, were investigated. The results showed that DMS could be treated effectively and almost a 100% removal efficiency was achieved at the conditions with an initial concentration of 832 mg/m3 and a gas flow rate of 1000 ml/min. Humidity boosts the removal efficiency and improves the energy yield (EY) greatly. The EY of 832 mg/m3 DMS was 2.87 mg/kJ when the relative humidity was above 30%. In the case of DMS removal, the ozone and nitrogen oxides were observed in the exhaust gas. The carbon and sulfur elements of DMS were mainly converted to carbon dioxide, carbon monoxide and sulfur dioxide. Moreover, sulfur was discovered in the reactor. According to the results, the optimization design for the reactor and the matching of high pulse voltage source can be reckoned.

  12. Annular Flow Distribution test

    International Nuclear Information System (INIS)

    This report documents the Babcock and Wilcox (B ampersand W) Annular Flow Distribution testing for the Savannah River Laboratory (SRL). The objective of the Annular Flow Distribution Test Program is to characterize the flow distribution between annular coolant channels for the Mark-22 fuel assembly with the bottom fitting insert (BFI) in place. Flow rate measurements for each annular channel were obtained by establishing ''hydraulic similarity'' between an instrumented fuel assembly with the BFI removed and a ''reference'' fuel assembly with the BFI installed. Empirical correlations of annular flow rates were generated for a range of boundary conditions

  13. Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    A. M. Ougouag; R. M. Ferrer

    2010-10-01

    The prismatic block version of the High Temperature Reactor (HTR) considered as a candidate Very High Temperature Reactor (VHTR)design may use burnable poison pins in locations at some corners of the fuel blocks (i.e., assembly equivalent structures). The presence of any highly absorbing materials, such as these burnable poisons, within fuel blocks for hexagonal geometry, graphite-moderated High Temperature Reactors (HTRs) causes a local inter-block flux depression that most nodal diffusion-based method have failed to properly model or otherwise represent. The location of these burnable poisons near vertices results in an asymmetry in the morphology of the assemblies (or blocks). Hence the resulting inadequacy of traditional homogenization methods, as these “spread” the actually local effect of the burnable poisons throughout the assembly. Furthermore, the actual effect of the burnable poison is primarily local with influence in its immediate vicinity, which happens to include a small region within the same assembly as well as similar regions in the adjacent assemblies. Traditional homogenization methods miss this artifact entirely. This paper presents a novel method for treating the local effect of the burnable poison explicitly in the context of a modern nodal method.

  14. Core management, operational limits and conditions and safety aspects of the Australian High Flux Reactor (HIFAR)

    Energy Technology Data Exchange (ETDEWEB)

    Town, S.L. [ANSTO, Nuclear Technology Div., Menai (Australia)

    1997-07-01

    HIFAR is a DIDO class reactor which commenced routine operation at approximately 10 MW in 1960. It is principally used for production of medical radio-isotopes, scientific research using neutron scattering facilities and irradiation of silicon ingots for the electronics industry. A detailed description of the core, including fuel types, is presented. Details are given of the current fuel management program HIFUEL and the experimental measurements associated with reactor physics analysis of HIFAR are discussed. (author)

  15. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    OpenAIRE

    Winter, Dominik

    2014-01-01

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use ...

  16. High peak-power monolithic femtosecond ytterbium fiber chirped pulse amplifier with a spliced-on hollow core fiber compressor.

    Science.gov (United States)

    Verhoef, A J; Jespersen, K; Andersen, T V; Grüner-Nielsen, L; Flöry, T; Zhu, L; Baltuška, A; Fernández, A

    2014-07-14

    We demonstrate a monolithic Yb-fiber chirped pulse amplifier that uses a dispersion matched fiber stretcher and a spliced-on hollow core photonic bandgap fiber compressor. For an output energy of 77 nJ, 220 fs pulses with 92% of the energy contained in the main pulse, can be obtained with minimal nonlinearities in the system. 135 nJ pulses are obtained with 226 fs duration and 82 percent of the energy in the main pulse. Due to the good dispersion match of the stretcher to the hollow core photonic bandgap fiber compressor, the duration of the output pulses is within 10% of the Fourier limited duration. PMID:25090494

  17. Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

    Science.gov (United States)

    Hu, Jian; Jiang, Nan; Li, Jie; Shang, Kefeng; Lu, Na; Wu, Yan; Mizuno, Akira

    2016-03-01

    The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. supported by National Natural Science Foundation of China (No. 51177007), the Joint Funds of National Natural Science Foundation of China (No. U1462105), and Dalian University of Technology Fundamental Research Fund of China (No. DUT15RC(3)030)

  18. High-Order Homogenization Method in Diffusion Theory for Reactor Core Simulation and Design Calculation

    International Nuclear Information System (INIS)

    Most modern nodal methods in use by the reactor vendors and utilities are based on the generalized equivalence theory (GET) that uses homogenized cross sections and flux discontinuity factors. These homogenized parameters, referred to as infinite medium parameters, are precomputed by performing single bundle fine-mesh calculations with zero current boundary conditions. It is known that for configurations in which the node-to-node leakage (e.g., surface current-to-flux ratio) is large the use of the infinite medium parameters could lead to large errors in the nodal solution. This would be the case for highly heterogeneous core configurations, typical of modern reactor core designs

  19. High-Order Homogenization Method in Diffusion Theory for Reactor Core Simulation and Design Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Farzad Rahnema

    2003-09-30

    Most modern nodal methods in use by the reactor vendors and utilities are based on the generalized equivalence theory (GET) that uses homogenized cross sections and flux discontinuity factors. These homogenized parameters, referred to as infinite medium parameters, are precomputed by performing single bundle fine-mesh calculations with zero current boundary conditions. It is known that for configurations in which the node-to-node leakage (e.g., surface current-to-flux ratio) is large the use of the infinite medium parameters could lead to large errors in the nodal solution. This would be the case for highly heterogeneous core configurations, typical of modern reactor core designs.

  20. Selecting a MAPLE research reactor core for 1-10 mW operation

    International Nuclear Information System (INIS)

    The MAPLE class of research reactors is designed so that a single reactor concept can satisfy a wide range of practical applications. This paper reports the results of physics studies performed on a number of potential core configurations fuelled with either 5 w/o or 8 w/o enriched UO2 or 20 w/o U3Si-Al and assesses the relative merits of each. Recommended core designs are given to maximize the neutron fluxes available for scientific application and isotope production

  1. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

    OpenAIRE

    2015-01-01

    Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed tha...

  2. University of Illinois nuclear pumped laser program. [experiments with a TRIGA pulsed reactor with a broad pulse and a low peak flux

    Science.gov (United States)

    Miley, G. H.

    1979-01-01

    The development of nuclear pumped lasers with improved efficiency, energy storage capability, and UF6 volume pumping is reviewed. Results of nuclear pumped laser experiments using a TRIGA-type pulsed reactor are outlined.

  3. Demonstration of the reactivity constraint approach on SNL's annual core research reactor

    International Nuclear Information System (INIS)

    This paper reports on the initial demonstration of the reactivity constraint approach and its implementing algorithm, the MIT-CSDL Non-Linear Digital Controller, on the annual core research reactor (ACCR) that is operated by the Sandia National Laboratories. This demonstration constituted the first use of reactivity constraints for the closed-loop, digital control of reactor power on a facility other than the Massachusetts Institute of Technology's (MIT's) research reactor (MITR-II). Also, because the ACRR and the MITR-II are of very different design, these trials established the generic nature of the reactivity constraint approach

  4. Advanced calculational methods for power reactors and LWR core design parameters

    International Nuclear Information System (INIS)

    The purpose of the Specialists Meeting on Advanced Calculational Methods for Power Reactors, held in Cadarache, France, 10-14 September 1990, was to provide a forum for reviewing and discussing selected core physics of water cooled reactors (including high convertors). New methods of advanced calculation for advanced fuels and complex geometries of next generation reactors with a high level of accuracy were discussed and the importance of supercomputing and on-line monitoring was also acknowledged. The meeting was attended by about 60 participants from 20 countries who presented 30 papers. The Technical Committee Meeting on LWR Core Design Parameters, held in Rez, former Czechoslovakia, 7-11 October 1991, provided an opportunity for participants to exchange their experience on reactor physics aspects of benchmark calculations of various lattices, methods for core parameter calculations, core monitoring and in-core fuel management. At the Workshop there were further discussions related to the benchmark problems, homogenization techniques and cross-section representations. Thirty-five papers were presented by about 43 participants from 19 countries. A separate abstract was prepared for each of the mentioned papers. Refs, figs and tabs

  5. Gas core reactors for actinide transmutation and breeder applications. Annual report

    International Nuclear Information System (INIS)

    This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions

  6. Conversion of methane to hydrogen by a pulsed plasma reactor

    International Nuclear Information System (INIS)

    A pulsed atmospheric glow discharge, employing corona as a preionization, was used to convert methane to hydrogen and higher hydrocarbons. The experimental results showed that the overall conversion and specific energy, defined as energy needed to dissociate one mole methane, was mainly dependent on E/P, banking capacitance, repetition rate and flow rate. The dependence on E/P, especially, is more pronounced. The minimum specific energy was less than 1 MJ and it is expected that it could be further lowered by choosing higher E/P, lower banking capacitance and introducing an oxidizer to enhance the conversion efficiency

  7. Thermodynamic performance of a gas-core fission reactor

    International Nuclear Information System (INIS)

    The purpose of this thesis was to investigate the thermodynamic behaviour of a critical quantity of gaseous uranium-fluorides in chemical equilibrium with a graphite wall. From the very beginning a container was considered with cooled walls. As it was evident that a nuclear reactor working with gaseous fuel should run at much higher temperatures than classical LWR or HTGR reactors, most of the investigations were performed for walls with a surface temperature of 1800 to 2000 K. It was supposed that such a surface temperature would be technologically possible for a heat load between 1 and 5 MWatt m-2. Cooling with high pressure helium-gas has to keep balance with this heat flux. The technical construction of such a wall will be a problem in itself. It is thought that the experiences with re-entry-vessels in space-technology can be used. A basic assumption in all the calculations is that the U-C-F reactor gas 'sees' a graphite wall, possibly graphite tiles supported by heat resistant materials like SiN2, SiC2 and at a lower temperature level by niobium-steel. Such a gastight compound-system is not necessarily of high-tensile strength materials. It has to be surrounded by a cooled neutron moderator-reflector which in its turn must be supported by a steel-wall at room temperature holding pressure of the order of 100 bar (10 MPa). The design of such a compound-wall is a task for the future. 116 refs.; 28 figs.; 29 tabs

  8. Radiation Characterization Summary: ACRR Cadmium-Polyethylene (CdPoly) Bucket Located in the Central Cavity on the 32-Inch Pedestal at the Core Centerline

    Energy Technology Data Exchange (ETDEWEB)

    Parma, Edward J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Naranjo, Gerald E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kaiser, Krista Irene [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Arnold, James F. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lippert, Lance L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Clovis, Ralph D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Martin, Lonnie E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Quirk, Thomas J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vehar, David W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-10-01

    This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the cadmium-polyethylene (CdPoly) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-CdPoly-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma-ray fluence profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples. Acknowledgements The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work. Also thanks to Drew Tonigan for helping field the activation experiments in ACRR, David Samuel for helping to finalize the drawings and get the parts fabricated, and Elliot Pelfrey for preparing the active dosimetry plots.

  9. Radiation Characterization Summary: ACRR Cadmium-Polyethylene (CdPoly) Bucket Located in the Central Cavity on the 32-Inch Pedestal at the Core Centerline.

    Energy Technology Data Exchange (ETDEWEB)

    Parma, Edward J.,; Naranjo, Gerald E.; Kaiser, Krista Irene; Arnold, James F.; Lippert, Lance L.; Clovis, Ralph D.; Martin, Lonnie E.; Quirk, Thomas J.; Vehar, David W.

    2016-10-01

    This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the cadmium-polyethylene (CdPoly) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-CdPoly-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma-ray fluence profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples. Acknowledgements The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work. Also thanks to Drew Tonigan for helping field the activation experiments in ACRR, David Samuel for helping to finalize the drawings and get the parts fabricated, and Elliot Pelfrey for preparing the active dosimetry plots.

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

    International Nuclear Information System (INIS)

    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)

  11. Micro-joule sub-10-fs VUV pulse generation by MW pump pulse using highly efficient chirped-four-wave mixing in hollow-core photonic crystal fibers

    OpenAIRE

    Im, Song-Jin

    2013-01-01

    We theoretically study chirped four-wave mixing for VUV pulse generation in hollow-core photonic crystal fibers. We predict the generation of sub-10-fs VUV pulses with energy of up to hundreds of microjoule by broad-band chirped idler pulses at 830 nm and MW pump pulses with narrow-band at 277 nm. MW pump could be desirable to reduce the complexity of the laser system or use a high repetition rate-laser system. The energy conversion efficiency from pump pulse to VUV pulse reaches to 30%. This...

  12. Application of mass-predictions to isotope-abundances in breeder-reactor cores

    CERN Document Server

    Kirchner, G

    1981-01-01

    The decay-heat and isotope composition of breeder reactor-cores is calculated at normal shut-down, and a core disintegration event. Using the ORIGEN-code, the influence of the most neutron-rich fission-yield nuclei is studied. Their abundances depend on the assumption about the nuclear data (mass and half-lives). The total decay-heat is not changed from any technical viewpoint. (15 refs).

  13. Application of mass-predictions to isotope-abundances in breeder-reactor cores

    International Nuclear Information System (INIS)

    The decay-heat and isotope composition of breeder reactor-cores is calculated at normal shut-down, and a core disintegration event. Using the ORIGRN-code, the influence of the most neutron-rich fission-yield nuclei is studied. Their abundances depend on the assumption about the nuclear data (mass and half-lives). The total decay-heat is not changed from any technically view-point. (orig.)

  14. Full core analysis of IRIS reactor by using MCNPX.

    Science.gov (United States)

    Amin, E A; Bashter, I I; Hassan, Nabil M; Mustafa, S S

    2016-07-01

    This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code. PMID:27135607

  15. Reactor core protection system using a 4-channel microcomputer

    International Nuclear Information System (INIS)

    A four channel microcomputer system was fitted in Grafenrheinfeld NPP for local core protection. This system performs continuous on-line monitoring of peak power density, departure from nucleate boiling ratio and fuel duty. The system implements limitation functions with more sophisticated criteria and improved accuracy. The Grafenrheinfeld system points the way to the employment of computer based limitation system, particularly in the field of programming language, demarkation of tasks, commissioning and documentation aids, streamlining of qualification and structuring of the system. (orig.)

  16. Microjoule sub-10 fs VUV pulse generation by MW pump pulses using highly efficient chirped four-wave mixing in hollow-core photonic crystal fibers

    Science.gov (United States)

    Im, Song-Jin

    2015-03-01

    We theoretically study chirped four-wave mixing for VUV pulse generation in hollow-core photonic crystal fibers. We predict the generation of sub-10 fs VUV pulses with energy of up to hundreds of µJ by broad-band chirped idler pulses at 830 nm and MW pump pulses with narrow-band at 277 nm. The MW pump could be desirable to reduce the complexity of the laser system or use a high repetition rate laser system. The energy conversion efficiency from pump pulse to VUV pulse reaches to 30% . This generation can be realized in a kagome-lattice hollow-core PCF filled with noble gas of high pressure with core diameter less than 40 µm, which would enable technically simple or highly efficient coupling to the fundamental mode of the fiber.

  17. Micro-joule sub-10-fs VUV pulse generation by MW pump pulse using highly efficient chirped-four-wave mixing in hollow-core photonic crystal fibers

    CERN Document Server

    Im, Song-Jin

    2013-01-01

    We theoretically study chirped four-wave mixing for VUV pulse generation in hollow-core photonic crystal fibers. We predict the generation of sub-10-fs VUV pulses with energy of up to hundreds of microjoule by broad-band chirped idler pulses at 830 nm and MW pump pulses with narrow-band at 277 nm. MW pump could be desirable to reduce the complexity of the laser system or use a high repetition rate-laser system. The energy conversion efficiency from pump pulse to VUV pulse reaches to 30%. This generation can be realized in kagome-lattice hollow-core PCF filled with noble gas of high pressure with core-diameter less than 40 micrometers which would enable technically simple or highly efficient coupling to fundamental mode of the fiber.

  18. 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.

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

    Science.gov (United States)

    Rabir, Mohamad Hairie; Hah, Chang Joo; Ju, Cho Sung

    2015-04-01

    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.

  20. Polarization-selective vortex-core switching by tailored orthogonal Gaussian-pulse currents

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Young-Sang; Lee, Ki-Suk; Jung, Hyunsung; Choi, Youn-Seok; Yoo, Myoung-Woo; Han, Dong-Soo; Im, Mi-Young; Fischer, Peter; Kim, Sang-Koog

    2011-05-01

    We experimentally demonstrate low-power-consumption vortex-core switching in magnetic nanodisks using tailored rotating magnetic fields produced with orthogonal and unipolar Gaussian-pulse currents. The optimal width of the orthogonal pulses and their time delay are found, from analytical and micromagnetic numerical calculations, to be determined only by the angular eigenfrequency ωD for a given vortex-state disk of polarization p, such that σ=1/ωD and Δt=π/2p/ωD. The estimated optimal pulse parameters are in good agreement with the experimental results. Finally, this work lays a foundation for energy-efficient information recording in vortex-core cross-point architecture.

  1. Pulsed power supply and coaxial reactor applied to E. coli elimination in water by pulsed dielectric barrier discharge

    Energy Technology Data Exchange (ETDEWEB)

    Quiroz V, V. E.; Lopez C, R.; Rodriguez M, B. G.; Pena E, R.; Mercado C, A.; Valencia A, R.; Hernandez A, A. N.; Barocio, S. R.; Munoz C, A. E. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); De la Piedad B, A., E-mail: regulo.lopez@inin.gob.mx [Instituto Tecnologico de Toluca, Av. Tecnologico s/n, Ex-Rancho La Virgen, 52140 Metepec, Estado de Mexico (Mexico)

    2013-07-01

    The design and instrumentation intended for ATTC8739 Escherichia coli (E. coli) bacteria elimination in water, based on non thermal plasma generation at room pressure have been carried out by means of dielectric pulsed discharges. The latter have been produced by a power supply capable of providing voltages up to the order of 45 kV, 1-500 {mu}s pulse widths and variable frequencies between 100 Hz to 2000 Hz. This supply feeds a coaxial discharge reactor of the simple dielectric barrier type. The adequate operation of the system has been tested with the elimination of E. coli at 10{sup 4} and 10{sup 6} bacteria/ml concentrations, leading to reductions up to 85.3% and 95.1%, respectively, during the first 30 min of treatment. (Author)

  2. Study of risk reduction by improving operation of reactor core isolation cooling system

    International Nuclear Information System (INIS)

    The Fukushima Daiichi nuclear power plant fell into a station blackout (SBO) due to the earthquake and tsunami in which most of the core cooling systems were disabled. In the units 2 and 3, water injection to the core was performed only by water injection system with turbine driven pumps. In particular, it is inferred from observed plant parameters that the reactor core isolation cooling system (RCIC) continued its operation much longer than it was originally expected (8 hours). Since the preparation of safety measures did not work, the reactor core damaged. With a view to reduce risk of station blackout events in a BWR by accident management, this study investigated the efficacy of operation procedures that takes advantage of RCIC which can be operated with only equipment inside reactor building and does not require an AC power source. The efficacy was assessed in this study by two steps. The first step is a thermal hydraulic analysis with the RETRAN3D code to estimate the potential extension of duration of core cooling by RCIC and the second step is the estimation of time required for recovery of off-site power from experiences at nuclear power stations under the 3.11 earthquake. This study showed that it is possible to implement more reliable measures for accident termination and to greatly reduce the risk of SBO by the installation of accident management measures with use of RCIC for extension of core cooling under SBO conditions. (author)

  3. Dynamic simulation and study of Mechanical Shim (MSHIM) core control strategy for AP1000 reactor

    International Nuclear Information System (INIS)

    Highlights: • A reactor core fast simulation program RCFSP is developed for AP1000. • A nodal core model and the MSHIM control strategy are implemented in RCFSP. • Load follow results for the original and revised MSHIM strategies are given. • Parameter sensitivity analysis and optimization of MSHIM control system are performed. - Abstract: The advanced Mechanical Shim (MSHIM) core control strategy is implemented in the AP1000 reactor by a digital rod control system. This control system comprises of two separate rod controllers that automatically control the core reactivity and axial power distribution using the gray and black M control banks and the axial offset (AO) control bank respectively. It has been demonstrated that the MSHIM control system can provide superior reactor control capabilities via automatic rod control only, which needs it to take more burdens than many other traditional core control systems during load change transients. This paper presents the dynamic simulation, and the parameter sensitivity analysis and optimization of the MSHIM control system for AP1000 reactor. A nodal core model is used to describe the dynamic behavior of the reactor core first. Then the nodal model and the original and revised MSHIM strategies are implemented in the AP1000 reactor to develop a fast simulation program in MATLAB/SIMULINK. Based on the simulation program, the MSHIM load follow and load regulation operations are simulated, the results of which demonstrate that the core reactivity and axial power distribution can be well-controlled via automatic rod control only. To show the effects of key factors on the control system behavior, the MSHIM load follow simulations with different control parameter values are performed. According to the simulation results and subsequent quantitative analysis, the mechanisms by which the key factors affect the control system behavior are illustrated and the optimum numerical ranges of these parameters are obtained. These

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

    Science.gov (United States)

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

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

  5. ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®

    Science.gov (United States)

    Damian, F.; Brun, E.

    2014-06-01

    ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.

  6. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

    Science.gov (United States)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

    1996-01-01

    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

  7. Pebble Bed Reactor Plant screening evaluation. Volume 1. Overall plant and reactor system

    International Nuclear Information System (INIS)

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW/sub t/ Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system. Core scoping studies were performed which evaluated the effects of annular and cyclindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations

  8. Langmuir probe diagnostic studies of pulsed hydrogen plasmas in planar microwave reactors

    CERN Document Server

    Rousseau, A; Lang, N; Hannemann, M; R"opcke, J

    2002-01-01

    Langmuir probe techniques have been used to study time and spatially resolved electron densities and electron temperatures in pulse-modulated hydrogen discharges in two different planar microwave reactors (fmicrowave= 2.45 GHz, tpulse= 1 ms). The reactors are (i) a standing-wave radiative slotted waveguide reactor and (ii) a modified travelling-wave radiative slotted waveguide reactor, which generate relatively large plasmas over areas from about 350 cm^2 to 500 cm^2. The plasma properties of these reactor types are of particular interest as they have been used for basic research and for plasma processing, e.g. for surface treatment and layer deposition. In the present study the pressures and microwave powers in the reactors were varied between 33 and 55 Pa and 600 and 3600 W, respectively. In regions with high electromagnetic fields shielded Langmuir probes were used to avoid disturbances of the probe characteristic. Close to the microwave windows of the reactors both the electron density and the electron te...

  9. Activities of research-reactor-technology project in FNCA from FY2005 to FY2007. Sharing neutronics calculation technique for core management and utilization of research reactors

    International Nuclear Information System (INIS)

    RRT project (Research-Reactor-Technology Project) was carried out with the theme of 'sharing neutronics calculation technique for core management and utilization of research reactors' in the framework of FNCA (Forum for Nuclear Cooperation in Asia) from FY2005 to FY2007. The objective of the project was to improve and equalize the level of neutronics calculation technique for the reactor core management among participating countries to assure the safe and stable operation of research reactors and the promotion of the effective utilization. Neutronics calculation codes, namely SRAC code system and MVP code, were adopted as common codes. Participating countries succeeded in applying the common codes to analyzing the core of each domestic research reactor. Some participating countries succeeded in applying the common codes to analyzing for utilization of own research reactors. Activities of RRT project have improved and equalized the level of neutronics calculation technique among participating countries. (author)

  10. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    International Nuclear Information System (INIS)

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used

  11. Advanced computational methods for the assessment of reactor core behaviour during reactivity initiated accidents. Final report

    International Nuclear Information System (INIS)

    The document at hand serves as the final report for the reactor safety research project RS1183 ''Advanced Computational Methods for the Assessment of Reactor Core Behavior During Reactivity-Initiated Accidents''. The work performed in the framework of this project was dedicated to the development, validation and application of advanced computational methods for the simulation of transients and accidents of nuclear installations. These simulation tools describe in particular the behavior of the reactor core (with respect to neutronics, thermal-hydraulics and thermal mechanics) at a very high level of detail. The overall goal of this project was the deployment of a modern nuclear computational chain which provides, besides advanced 3D tools for coupled neutronics/ thermal-hydraulics full core calculations, also appropriate tools for the generation of multi-group cross sections and Monte Carlo models for the verification of the individual calculational steps. This computational chain shall primarily be deployed for light water reactors (LWR), but should beyond that also be applicable for innovative reactor concepts. Thus, validation on computational benchmarks and critical experiments was of paramount importance. Finally, appropriate methods for uncertainty and sensitivity analysis were to be integrated into the computational framework, in order to assess and quantify the uncertainties due to insufficient knowledge of data, as well as due to methodological aspects.

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

    International Nuclear Information System (INIS)

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

  13. Reducing numerical costs for core wide nuclear reactor CFD simulations by the Coarse-Grid-CFD

    Science.gov (United States)

    Viellieber, Mathias; Class, Andreas G.

    2013-11-01

    Traditionally complete nuclear reactor core simulations are performed with subchannel analysis codes, that rely on experimental and empirical input. The Coarse-Grid-CFD (CGCFD) intends to replace the experimental or empirical input with CFD data. The reactor core consists of repetitive flow patterns, allowing the general approach of creating a parametrized model for one segment and composing many of those to obtain the entire reactor simulation. The method is based on a detailed and well-resolved CFD simulation of one representative segment. From this simulation we extract so-called parametrized volumetric forces which close, an otherwise strongly under resolved, coarsely-meshed model of a complete reactor setup. While the formulation so far accounts for forces created internally in the fluid others e.g. obstruction and flow deviation through spacers and wire wraps, still need to be accounted for if the geometric details are not represented in the coarse mesh. These are modelled with an Anisotropic Porosity Formulation (APF). This work focuses on the application of the CGCFD to a complete reactor core setup and the accomplishment of the parametrization of the volumetric forces.

  14. Sodium removing facility for core-constitutional elements of FBR type reactor

    International Nuclear Information System (INIS)

    Reactor core-constitutional elements as spent reactor core fuel assemblies are contained in a containing vessel. An inert gas (N2, Ar or He) is filled in the containing vessel through an inert gas supply channel. The temperature of the inert gas is raised by the remaining after heat of the reactor core-constitutional elements. The inert gas is circulated and heated through a preheating circuit by driving a recycling gas blower and returned to the containing vessel. If the inert gas is heated to a predetermined temperature, metal sodium deposited on the surface of the materials of the reactor core-constitutional elements is evaporated. Next, a vacuum pump unit of a vacuum exhaustion channel is driven to suck an inert gas entraining sodium vapor in the containing vessel, and the sodium vapor is cooled, condensed thereby separated in a sodium separator. Then, the inert gas at a low temperature is introduced to a vacuum exhaustion channel to remove and discharge remained sodium vapor by a sodium trap. (I.N.)

  15. The effects of stainless steel radial reflector on core reactivity for small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr [KINGS, 658-91, Haemaji-ro, Seosaeng-myeon, Ulju-gun, Ulsan, 689-882 (Korea, Republic of); Cho, Sung Ju, E-mail: sungju@knfc.co.kr; Seong, Ki Bong, E-mail: kbseong@knfc.co.kr [KNFC, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2016-01-22

    Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5 w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.

  16. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  17. The effects of stainless steel radial reflector on core reactivity for small modular reactor

    International Nuclear Information System (INIS)

    Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5 w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO2 fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector

  18. 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 si

  19. Design of the zero power reactor core of Instituto de Energia Atomica, SP, Brazil

    International Nuclear Information System (INIS)

    The main characteristics of a graphite moderated core of a critical assembly to be installed in the zero power reactor of the Instituto de Energia Atomica have been defined. Several simple geometric configurations have been selected and criticality studies have been made. The necessary quantity of fissile uranium has been calculated. (author)

  20. Core characteristics on a hybrid type fast reactor system combined with proton accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Kowata, Yasuki; Otsubo, Akira [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    In our study on a hybrid fast reactor system, we have investigated it from the view point of transmutation ability of trans-uranium (TRU) nuclide making the most effective use of special features (controllability, hard neutron spectrum) of the system. It is proved that a proton beam is superior in generation of neutrons compared with an electron beam. Therefore a proton accelerator using spallation reaction with a target nucleus has an advantage to transmutation of TRU than an electron one. A fast reactor is expected to primarily have a merit that the reactor can be operated for a long term without employment of highly enriched plutonium fuel by using external neutron source such as the proton accelerator. Namely, the system has a desirable characteristic of being possible to self-sustained fissile plutonium. Consequently in the present report, core characteristics of the system were roughly studied by analyses using 2D-BURN code. The possibility of self-sustained fuel was investigated from the burnup and neutronic calculation in a cylindrical core with 300w/cc of power density without considering a target material region for the accelerator. For a reference core of which the height and the radius are both 100 cm, there is a fair prospect that a long term reactor operation is possible with subsequent refueling of natural uranium, if the medium enriched (around 10wt%) uranium or plutonium fuels are fully loaded in the initial core. More precise analyses will be planed in a later fiscal year. (author)

  1. Development of source range measurement instrument in Xi'an pulsed reactor

    CERN Document Server

    Wang Li

    2002-01-01

    Source range measurement instrument in Xi'an pulsed reactor is key equipment of low-side measuring in source range. At the same time, it is also weighty component of out-of-pile neutron-flux level observation system. The authors have done some researching and renovating based on the similar type devices used in nuclear reactor to improve the meter sensitivity, measuring range, noise proof features, reliability in running and maintainability which belong to the main performance index of the instrument. The design ideas, configurations, working principle, performance indexes, technique features and effect in utilizing are introduced briefly

  2. The behaviour of weldable strain gauges under nuclear reactor core conditions

    International Nuclear Information System (INIS)

    Highlights: • Electrical resistance strain gauges can be used in nuclear reactor environments. • Several in-reactor experiments featuring the use of strain gauges are described. • Static and dynamic strain can be measured under the very hostile conditions of nuclear reactor cores. • Irradiation effects can be corrected if they are well understood and precisely quantified. -- Abstract: Electrical resistance strain gauges are a very useful tool to measure mechanical parameters such as deformation, stress, dynamic strain, vibration, etc. This paper presents our experience with strain gauges in nuclear reactor environments. The nature of nuclear applications and the desirable characteristics of nuclear strain gauges are discussed. Several in-reactor experiments featuring the use of strain gauges are described. The behaviour of weldable strain gauges under intense nuclear radiation is discussed. Experimental results and techniques for the successful applications of strain gauges in nuclear environments are presented. It is concluded that weldable stain gauges can be used successfully under the very hostile conditions of nuclear reactor cores if appropriate procedures are followed

  3. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Gehin, Jess C [ORNL

    2009-04-01

    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  4. Thermal hydraulics characterization of the core and the reactor vessel type BWR

    International Nuclear Information System (INIS)

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  5. Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

    International Nuclear Information System (INIS)

    The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

  6. Implications for accident management of adding water to a degrading reactor core

    International Nuclear Information System (INIS)

    This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents

  7. Implications for accident management of adding water to a degrading reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-02-01

    This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents.

  8. Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Kawaji, Masahiro [City College of New York, NY (United States); Valentin, Francisco I. [City College of New York, NY (United States); Artoun, Narbeh [City College of New York, NY (United States); Banerjee, Sanjoy [City College of New York, NY (United States); Sohal, Manohar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schultz, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); McEligot, Donald M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-12-21

    The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

  9. Development of the evaluation methods in reactor safety analyses and core characteristics

    International Nuclear Information System (INIS)

    In order to support the safety reviews by NRA on reactor safety design including the phenomena with multiple failures, the computer codes are developed and the safety evaluations with analyses are performed in the areas of thermal hydraulics and core characteristics evaluation. In the code preparation of safety analyses, the TRACE and RELAP5 code were prepared to conduct the safety analyses of LOCA and beyond design basis accidents with multiple failures. In the core physics code preparation, the functions of sensitivity and uncertainty analysis were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 /SIMULATE-3 was continued by using core physics data. (author)

  10. A wide range in-core neutron monitoring system for high powered TRIGA reactors

    International Nuclear Information System (INIS)

    High power movable core TRIGA reactors present unique problems of determining power levels from a neutron flux measurement because of (1) difficulty of locating detectors; (2) water thermal effects and (3) effect of experimental facilities. A solution, along with experimental results, will be described that uses a beam tube to effectively make in-core flux measurements with an out-of-core detector. The application of this new type of detector assembly to wide range linear and log power measurement will also be discussed. (author)

  11. Analysis and control of welding deformation in nuclear reactor core barrel

    International Nuclear Information System (INIS)

    Core barrel is an important part in React Vessel Internals(RVI) and the design of tolerance is quite strict. Firstly, the paper analyses the reasons of welding deformation in core barrel and with the reasons, the paper presents control measures to solve welding deformation. Then, combining the severe deformation in the welding of core barrel in Qinshan Nuclear Power Phase II extension Project Reactor no.3 unit, the paper supply control measures to the no.4 unit and the result of weld satisfy the requirements and the result satisfy design requirement. (authors)

  12. 2240-MW(th) high-temperature reactor core power density study

    International Nuclear Information System (INIS)

    This study was done to estimate the effects of reducing the design power density of a 2240-MW(t) high-temperature gas-cooled reactor. Core history and thermal hydraulics calculations were performed for average power densities of 5.8 and 7.2 W/cm3 and the use of highly enriched fuel was considered. The fuel temperature conditions for the higher power density were found to be only moderately elevated at normal operating conditions. Economic considerations associated with changes in core performance, core size, and coolant pumping requirements were assessed

  13. Nuclear Power Station Kalkar, 300 MWe Nuclear Prototype Power Station with Fast Sodium Cooled Reactor (SNR-300), Short Description of the Reactor Core Mark-Ia

    International Nuclear Information System (INIS)

    The nuclear power station Kalkar is a prototype with a sodium cooled fast reactor (SNR-300) and a thermal power of 762 MW. The initial licensing procedure in 1972 was based on the so-called Mark-I core. During the following years, this core underwent some changes, for instance the thickness of the radial blanket was reduced to lower the electricity generation costs, the design of the absorber systems had been further optimized, and it became clear, that a full core with plutonium from MAGNOX-reactors could not be realized and that fuel from light-water reactors had also to be used. In this licensing document the modified reactor core Mark-Ia is described, and the radiological consequences of the core modification are quantified to be tolerable

  14. 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... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  15. Electric drive for accelerator target with a timer for a fast pulsed reactor

    International Nuclear Information System (INIS)

    The invention refers to electric drives with digital control. The electric drive provides for cophased rotation of the target-containing particle reflector and the frequency of the accelerator functioning, the syncronization being provided by the current mains. At the same time the reflector may rotate with any given velocity in the necessary range, this being required also for the work of pulse fast neutron reactor. The drive involves a driving syncronic engine, an electromagnetic clutch, a pulse velocity indicator and a digital regulator with a velocity measuring block, a counter and a memory block

  16. The core design of ALFRED, a demonstrator for the European lead-cooled reactors

    International Nuclear Information System (INIS)

    Highlights: • The design for the lead fast reactor is conceived in a comprehensive approach. • Neutronic, thermal-hydraulic, and transient analyses show promising results. • The system is designed to withstand even design extension conditions accidents. • Activation products in lead, including polonium, are evaluated. - Abstract: The European Union has recently co-funded the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project, in the frame of which the preliminary designs of an industrial size lead-cooled reactor (1500 MWth) and of its demonstrator reactor (300 MWth) were developed. The latter is called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) and its core, as designed and characterized in the project, is presented here. The core parameters have been fixed in a comprehensive approach taking into account the main technological constraints and goals of the system from the very beginning: the limiting temperature of the clad and of the fuel, the Pu enrichment, the achievement of a burn-up of 100 GWd/t, the respect of the integrity of the system even in design extension conditions (DEC). After the general core design has been fixed, it has been characterized from the neutronic point of view by two independent codes (MCNPX and ERANOS), whose results are compared. The power deposition and the reactivity coefficient calculations have been used respectively as input for the thermal-hydraulic analysis (TRACE, CFD and ANTEO codes) and for some preliminary transient calculations (RELAP, CATHARE and SIM-LFR codes). The results of the lead activation analysis are also presented (FISPACT code). Some issues of the core design are to be reviewed and improved, uncertainties are still to be evaluated, but the verifications performed so far confirm the promising safety features of the lead-cooled fast reactors

  17. Comparing neutronics codes performance in analyzing a fresh-fuelled research reactor core

    International Nuclear Information System (INIS)

    Highlights: • Calculation of neutron fluence rate with different neutronic codes is examined. • MCNP, TRIPOLI and CITATION were used for neutron fluence rate calculations. • The recently converted core of the Portuguese Research Reactor (RPI) was used. • Fresh fuel of low enrichment in U-235 was assumed. • Thermal, epithermal and fast neutron fluence rates were computed. - Abstract: In this paper the relative performance of different simulation approaches is examined, focusing on the neutron fluence rate distribution in a nuclear reactor core. The main scope of the work is to benchmark and validate the neutronics code systems utilized in the Greek Research Reactor (GRR-1) for a high-density Low Enriched Uranium (LEU) core of compact size. For this purpose the recently converted core of the Portuguese Research Reactor (RPI), fueled with fresh, low enrichment in U-235 fuel, was simulated with the stochastic code TRIPOLI and the deterministic code system XSDRN/CITATION. RPI was selected on the basis that it is a similar to GRR-1 pool-type reactor, using same fuel and control rods type, as well as same types of coolant, moderator and reflector. The neutron fluence rate in RPI was computed using each numerical approach with changed approximations. In this frame the stochastic code TRIPOLI was tested using two different nuclear data libraries, i.e., ENDF/B-VI versus JEFF3.1, and two different ways of source definition, i.e., “point sources”, placed in the center of each fuel cell, versus a “distributed source”, where each fuel volume was considered as a neutron source. The deterministic code system XSDRN/CITATION was tested with respect to the definition of the transverse leakages associated to each one-dimensional, user-defined core zone, as analyzed by the XSDRN code in order to provide the zone equivalent cross sections. Thermal, epithermal and fast neutron fluence rates were computed and local values found in a 15 cm segment immediately below the

  18. MTR research reactor core behavior under a loss of shutdown heat removal

    International Nuclear Information System (INIS)

    Full text of publication follows: Introduction: Heat removal during operation of medium power research reactors is assumed to be safely performed by forced convection and the adequate removal of residual decay heat after reactor shutdown need to continue forced convection removal for a certain period of time when the operating power before shut-down is above a certain power level. This is among the requirement for the overall safety of research reactor operation. Objective: The purpose of the present work is: - to estimate the maximum temperature in the core and to investigate the minimum power operating level before shutdown that needs a continuation forced convection after shutdown; - to evaluate occurrence of cladding damage following a shutdown reactor without forced convection Problem: The simulation process is undertaken using the RELAP5/Mod 3.2 code system. The IAEA 10 MW benchmark core which is a representative of medium pool type MTR research reactors was chosen here in order to investigate the cladding maximum fuel temperature without forced heat removal after shutdown of the reactor that was operating at different powers up to 10 MW. Nodalization: The benchmark core consists of 25 fuels elements placed in a 5 x 5 gird placed within pool filled by 9 m of light water. The primary loop is represented by pumps, pipes and heat exchangers. Each of the 25 fuel elements is represented individually. Results: The simulation process has shown that the cladding maximum temperature did not reach the melting point for aluminum (660 deg. C) but void is expected to be produced in the hot channels. Hence, the loss of forced heat removal after reactor scram did not induce any melting of the cladding by much deeper investigation may be undertaken because presence of void in channels could enhance corrosion phenomena and may induce some fission products release in the pools following localized fuel rupture due to corrosion. (authors)

  19. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  20. Studies on influence of sodium void reactivity effect on the concept of the core and safety of advanced fast reactor

    International Nuclear Information System (INIS)

    The paper is devoted to studies on influence of sodium void reactivity effect (SVRE) on safety and technical and economical characteristics of BN-1200 type reactor. Different core options are considered as applied to this reactor. These core options differ in designs, dimensions and, hence, SVRE value. It is shown by the analysis that most flattened core with sodium plenum at the top assures reactor self-protection under beyond design basis accident conditions. Sodium plenum abandonment and core height increase causing SVRE increase deteriorate reactor self-protection, but at the same time, improve some technical and economical characteristics of the reactor. Issues of choosing optimal core design under these conditions are discussed. (author)

  1. CP ESFR: Collaborative Project for a European Sodium Fast Reactor Core studies

    International Nuclear Information System (INIS)

    • Significant progress has been made in optimizing both the oxide and carbide ESFR cores; • For the oxide core the optimisation process concentrated on the reduction of the sodium void reactivity effect and on the evaluation of MA burning performances. The CONF2 axial configuration has provided a significant overall reduction of the sodium void reactivity effect. • The carbide core had a significantly higher reactivity loss over the fuel cycle compared to the oxide one. By increasing slightly the fuel pin diameter, whilst still retaining the advantages of lower fuel temperatures of carbide fuel, and making changes in the core layout, the reactivity loss over the cycle has been reduced to a level similar to that of the oxide core. By adopting the CONF2 axial configuration initially developed for the oxide core, the sodium void reactivity of the carbide core has also been reduced appreciably. • The MA transmutation performances of the optimized ESFR oxide core have been investigated with respect to two boundary configurations. The HET2 configuration shows a low MA transmutation rate sufficient to burn the MA produced by the ESFR core without affecting the safety parameters. The HOM4 configuration (where 4%wt. MA are loaded homogeneously in each core SA) is the most challenging configuration due to its impact on safety coefficients but it shows an high MA burning rate suitable for burning also MA accumulated by a thermal reactor fleet

  2. Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity (ρex), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

  3. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    International Nuclear Information System (INIS)

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code

  4. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  5. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Science.gov (United States)

    Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-01

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  6. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    International Nuclear Information System (INIS)

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs

  7. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Budd, W.A. (ed.)

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  8. Status report on the core conversion of the ASTRA-reactor

    International Nuclear Information System (INIS)

    The core conversion from HEU-fuel to LEU-fuel of the ASTRA-Reactor, a pool type research reactor with a thermal power of 10 MW, was started in 1982. In the beginning of the conversion MEU-fuel elements with UAlx-Al fuel and LEU-fuel elements with U3O8-Al fuel were successfully tested. Since 1985 only LEU-fuel elements with U3Si2-Al fuel were used. The HEU-fuel elements in the core were successively replaced by LEU-fuel elements. At the present time there are still 5 HEU-fuel elements in the core. The conversion will be finished in 1990. (orig.)

  9. Coolability of degraded core under reflooding conditions in Nordic boiling water reactors

    International Nuclear Information System (INIS)

    Present work is part of the first phase of subproject RAK-2.1 of the new Nordic Co-operative Reactor Safety Program, NKS. The first phase comprises reflooding calculations for the boiling water reactors (BWRs) TVO I/II in Finland and Forsmark 3 in Sweden, as a continuation of earlier severe accident analyses which were made in the SIK-2 project. The objective of the core reflooding studies is to evaluate when and how the core is still coolable with water and what are the probable consequences of water cooling. In the following phase of the RAK-2.1 project, recriticality studies will be performed. Conditions for recriticality might occur if control rods have melted away with the fuel rods intact in a shape that critical conditions can be created in reflooding with insufficiently borated water. Core coolability was investigated for two reference plants, TVO I/II and Forsmark 3. The selected accident cases were anticipated station blackout with or without successful depressurization of reactor coolant system (RCS). The effects of the recovery of emergency core cooling (ECC) were studied by varying the starting time of core reflooding. The start of ECC systems were assigned to reaching a maximum cladding temperature: 1400 K, 1600 K, 1800 K and 2000 K in the core. Cases with coolant injection through the downcomer were studied for TVO I/II and both downcomer injection and core top spray were investigated for Forsmark 3. Calculations with three different computer codes: MAAP 4, MELCOR 1.8.3 and SCDA/RELAP5/MOD 3.1 for the basis for the presented reflooding studies. Presently, and experimental programme on core reflooding phenomena has been started in Kernforschungszentrum Karlsruhe in QUENCH test facility. (EG) 17 refs

  10. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  11. Measurements of the Physics Characteristics of the Fast Pulsed Reactor, VIPER

    International Nuclear Information System (INIS)

    The VIPER pulsed reactor was first made critical on 26 May 1967. The physics measurements carried out during its subsequent commissioning to prompt critical are described. These experiments include the measurement of reaction rate distributions, perturbation effects, neutron spectrum, neutron lifetime, and prompt temperature coefficients of reactivity. The measured characteristics of prompt critical transients are compared with predictions based on calculated expansion and Doppler effects. (author)

  12. Extremely Nonlinear Optics Using Shaped Pulses Spectrally Broadened in an Argon- or Sulfur Hexafluoride-Filled Hollow-Core Fiber

    OpenAIRE

    Andreas Hoffmann; Michael Zürch; Christian Spielmann

    2015-01-01

    In this contribution we present a comparison of the performance of spectrally broadened ultrashort pulses using a hollow-core fiber either filled with argon or sulfur hexafluoride (SF6) for demanding pulse-shaping experiments. The benefits of both gases for pulse-shaping are studied in the highly nonlinear process of high-harmonic generation. In this setup, temporally shaping the driving laser pulse leads to spectrally shaping of the output extreme ultraviolet (XUV) spectrum, where total yie...

  13. Validation of the neutron and gamma fields in the JSI TRIGA reactor using in-core fission and ionization chambers.

    Science.gov (United States)

    Žerovnik, Gašper; Kaiba, Tanja; Radulović, Vladimir; Jazbec, Anže; Rupnik, Sebastjan; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-02-01

    CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring. PMID:25479432

  14. Reactor physics analysis for HANARO core conversion using high density U-Mo fuel

    International Nuclear Information System (INIS)

    Currently, HANARO is using U3Si/Al fuel of 3.15 gU/cc. To enhance the utilization of HANARO, core conversion using high density U-Mo fuel is studied. Minimal core conversion considered maintains fuel shape and only changes fuel density. U7Mo/AI of 4.0/4.5 gU/cc which has been irradiated at HANARO, and U7Mo/AI of 5.0/4.3 gU/cc for the next irradiation test are considered. Important reactor physics parameters such as linear heat generation rate, neutron flux, and reactivity, are compared. A new core model for U7Mo/Al fuel offers additional 4 irradiation sites. U7Mo/ Al core give cycle length extension of 16% and 27%, but a little bit of neutron flux decrease. The increase of linear heat generation rate in a compact U7Mo/Al core is suppressed by the optimized design of fuel assembly. Reactivity effects of U7Mo/Al core are similar to the current core. Core conversion using high density U-Mo fuel give additional irradiation sites and extension of core cycle without any significant loss

  15. Radiation Characterization Summary: ACRR Polyethylene-Lead-Graphite (PLG) Bucket Located in the Central Cavity on the 32-Inch Pedestal at the Core Centerline (ACRR-PLG-CC-32-cl).

    Energy Technology Data Exchange (ETDEWEB)

    Parma, Edward J.,; Vehar, David W.; Lippert, Lance L.; Griffin, Patrick J.; Naranjo, Gerald E.; Luker, Spencer M.

    2015-06-01

    This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the polyethylene-lead-graphite (PLG) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-PLG-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma-ray fluence profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples. Acknowledgements The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work. Also thanks to David Ames for his assistance in running MCNP on the Sandia parallel machines.

  16. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

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

    International Nuclear Information System (INIS)

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

  18. Feasibility of a Pulsed Sequencing Batch reactor with anaerobic aggregated biomass for the treatment of low strength wastewaters

    OpenAIRE

    Brito, A. G.; Rodrigues, A.C.; Melo, L. F.

    1997-01-01

    This study concerns an assessment of a SBR operation that associates anaerobic aggregated biomass with a pulsed action during the reaction phase, a system named Pulsed Sequencing Batch Reactor (P-SBR). The system uses a diaphragm pump as a pulsator unit to increase the liquid-solid contact, in order to avoid dead zones and possible external mass transfer resistance. A preliminary study of the operation of the reactor was performed with a low strength synthetic wastewater with a COD near 1000 ...

  19. Treatment of Dye Wastewater by Using a Hybrid Gas/Liquid Pulsed Discharge Plasma Reactor

    Institute of Scientific and Technical Information of China (English)

    鲁娜; 李杰; 吴彦; 佐藤正之

    2012-01-01

    A hybrid gas/liquid pulsed discharge plasma reactor using a porous ceramic tube is proposed for dye wastewater treatment. High voltage pulsed discharge plasma was generated in the gas phase and simultaneously the plasma channel was permeated through the tiny holes of the ceramic tube into the water phase accompanied by gas bubbles. The porous ceramic tube not only separated the gas phase and liquid phase but also offered an effective plasma spreading channel. The effects of the peak pulse voltage, additive gas varieties, gas bubbling rate, solution conductivity and TiO2 addition were investigated. The results showed that this reactor was effective for dye wastewater treatment. The decoloration efficiency of Acid Orange II was enhanced with an increase in the power supplied. Under the studied conditions, 97% of Acid Orange II in aqueous solution was effectively decolored with additive oxygen gas, which was 51% higher than that with argon gas, and the increasing 02 bubbling rate also benefited the decoloration of dye wastewater. Water conductivity had a small effect on the level of decoloration. Catalysis of TiO2 could be induced by the pulsed discharge plasma and addition of TiO2 aided the decoloration of Acid Orange II.

  20. Gamma compensated pulsed ionization chamber wide range neutron/reactor power measurement system

    International Nuclear Information System (INIS)

    An improved method and system of pulsed mode operation of ionization chambers is described in which a single sensor system with gamma compensation is provided by sampling, squaring, automatic gate selector, and differential amplifier circuit means, employed in relation to chambers sensitized to neutron plus gamma and gamma only to subtract out the gamma component, wherein squaring functions circuits, a supplemental high performance pulse rate system, and operational and display mode selection and sampling gate circuits are utilized to provide automatic wide range linear measurement capability for neutron flux and reactor power. Neon is employed as an additive in the ionization chambers to provide independence of ionized gas kinetics temperature effects, and the pulsed mode of operation provide independence of high temperature insulator leakage effects. (auth)

  1. Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

    Science.gov (United States)

    Sager, G. T.; Wong, C. P. C.; Kapich, D. D.; McDonald, C. F.; Schleicher, R. W.

    1993-11-01

    The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankine and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The close cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed.

  2. Methods of first wall structural analysis with applications to the long pulse commercial tokamak reactor design

    International Nuclear Information System (INIS)

    Methods of analysis for fusion first wall design are developed. Several design limits have been evaluated and combined to present trade-offs in the form of design windows. These considerations include limits related to thermal fatigue, primary membrane strength, displacement under loading, ratcheting, radiation damage, and plasma-wall interactions. Special emphasis is placed on the investigation of thermal fatigue using a two dimensional treatment of a tubular first wall configuration. The work is motivated by the proposal of the Ultra Long Pulse Commercial Reactor (ULTR), a machine capable of delivering plasma burn pulses of up to 24 hours in length. The present work looks in detail at the impact of pertinent characteristics of the ULTR design such as pulse length, coolant pressure, first wall thickness and first wall lifetime on the structural effects considered. Computer programs are developed and consider several major structural effects on a cylindrical first wall element for both 316 stainless steel and vanadium alloy

  3. Nuclear design of the burst power ultrahigh temperature UF4 vapor core reactor system

    Science.gov (United States)

    Kahook, Samer D.; Dugan, Edward T.

    1991-01-01

    Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The fuel/working fluid is a mixture of UF4 and metal fluoride. Preliminary calculations indicate high overall system efficiencies (≊20%), small radiator size (≊5 m2/MWe), and high specific power (≊5 kWe/kg). Neutronic analysis has revealed a number of attractive features for this novel reactor concept. These include some unique and very effective inherent negative reactivity control mechanisms such as the vapor-fuel density power coefficient of reactivity, the direct neutronic coupling among the multiple fissioning core regions (the central vapor core and the surrounding boiler columns), and the mass flow coupling feedback between the fissioning cores.

  4. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

  7. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  8. Sodium Fast Reactor Core Definitions (version 1.2 - September 19)

    International Nuclear Information System (INIS)

    The Generation IV International Forum (GIF) has defined the key research goals for advanced sodium-cooled fast reactors (SFR): - improved safety performance, specifically a demonstration of favourable transient behaviour under accident conditions; - improved economic competitiveness; - demonstration of flexible management of nuclear materials, in particular, waste reduction through minor actinide burning. With respect to SFR safety performance, one of the foremost GIF objective is to design cores that can passively avoid damage when the control rods fail to scram in response to postulated accident initiators (e.g., inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Under the auspices of the Working Party on Reactor and System (WPRS), a mandate has been proposed to work towards a shared analysis of the feedback and transient behaviour of next generation SFR concepts. In order to achieve these goals, a step-by-step analysis approach has been proposed: 1. Compile a 'state of the art' report: review past and recent studies performed in the framework of sodium fast reactor and build a bibliographic repository which would stress core transient behaviours as a function of fuel characteristics (oxide, carbide, nitride and metal). 2. Perform a parametric study based on two different core sizes: large size core (3600 MW thermal) and medium size core (1500-2500 MW thermal). For both cores sizes three types of fuel are proposed: oxide, carbide and metal. This comparative study is aimed at identifying the advantages and drawbacks for each concept based on nominal performances and global safety parameters: - Neutronics characterisation of global parameters (k-eff, power and flux distributions, void effect, Doppler, etc.); - Feed-back coefficient evaluation, discussion and agreement on corresponding calculation

  9. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Foyto, L [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Kutikkad, K [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; McKibben, J C [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Peters, N. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  10. Subcutaneous granuloma annulare

    Directory of Open Access Journals (Sweden)

    Dhar Sandipan

    1994-01-01

    Full Text Available Two cases of subcutaneos granuloma annulare are reported. Clinical presentation was in the form of hard subcutaneous nodules; histopathology confirmed the clinical diagnosis. The cases were unique because of onset in adult hood, occurrence over unusual sites and absence of classical lesions of granuloma annulare elsewhere.

  11. Subcutaneous granuloma annulare

    Directory of Open Access Journals (Sweden)

    Dhar Sandipan

    1993-01-01

    Full Text Available Two cases of subcutaneous granuloma annulare are reported. Clinical presentation was in the form of hard subcutaneous nodules, histopathology confirmed the clinical diagnosis. The cases were unique because of onset in adult age, occurrence over unusual sites and absence of classical lesions of granuloma annulare elsewhere.

  12. Fuel performance models for high-temperature gas-cooled reactor core design

    International Nuclear Information System (INIS)

    Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10-4 fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience

  13. A Metropolis algorithm combined with Nelder-Mead Simplex applied to nuclear reactor core design

    International Nuclear Information System (INIS)

    A hybridization of the recently introduced Particle Collision Algorithm (PCA) and the Nelder-Mead Simplex algorithm is introduced and applied to a core design optimization problem which was previously attacked by other metaheuristics. 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. The new metaheuristic performs better than the genetic algorithm, particle swarm optimization, and the Metropolis algorithms PCA and the Great Deluge Algorithm, thus demonstrating its potential for other applications

  14. The effects of aging on Boiling Water Reactor core isolation cooling system

    International Nuclear Information System (INIS)

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling system in commercial Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research program sponsored by the US Nuclear Regulatory Commission. The failure data, from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failure causes

  15. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W [Westinghouse Savannah River Co., Aiken, SC (USA); Hagrman, D L [EG and G Idaho, Inc., Idaho Falls, ID (USA); McClure, P R; Leonard, M T [Science Applications International Corp., Albuquerque, NM (USA)

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  16. A Metropolis algorithm combined with Nelder-Mead Simplex applied to nuclear reactor core design

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner F. [Depto. de Modelagem Computacional, Instituto Politecnico, Universidade do Estado do Rio de Janeiro, R. Alberto Rangel, s/n, P.O. Box 972285, Nova Friburgo, RJ 28601-970 (Brazil)], E-mail: wfsacco@iprj.uerj.br; Filho, Hermes Alves; Henderson, Nelio [Depto. de Modelagem Computacional, Instituto Politecnico, Universidade do Estado do Rio de Janeiro, R. Alberto Rangel, s/n, P.O. Box 972285, Nova Friburgo, RJ 28601-970 (Brazil); Oliveira, Cassiano R.E. de [Nuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States)

    2008-05-15

    A hybridization of the recently introduced Particle Collision Algorithm (PCA) and the Nelder-Mead Simplex algorithm is introduced and applied to a core design optimization problem which was previously attacked by other metaheuristics. 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. The new metaheuristic performs better than the genetic algorithm, particle swarm optimization, and the Metropolis algorithms PCA and the Great Deluge Algorithm, thus demonstrating its potential for other applications.

  17. 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.

  18. The behavior of reactor power and flux resulting from changes in core-coolant temperature for a miniature neutron source reactor

    International Nuclear Information System (INIS)

    In this work, measurements were performed to verify the theoretical predictions of reactor power and flux parameters that result from changes in core inlet temperature (Tin) and the temperature difference between the coolant inlet and outlet (ΔT) in the Nigeria Research Reactor-1 (NIRR-1), which is a Miniature Neutron Source Reactor (MNSR). The measured data shows that there is a strong dependence of the reactor power on coolant temperature in agreement with the design of MNSR. The experimental parameters were found to be in good agreement with data obtained using a semi-empirical relationship between the reactor power, flux parameters, core inlet temperature, and the coolant temperature rise. The relationship was therefore used to predict the power level of NIRR-1 from its neutron flux parameters to which it has been found to be proportional. The variation of Tin and ΔT with the reactor power and flux was also investigated and the results obtained are hereby discussed

  19. A study on fast reactor core mechanics by an ex-reactor test and comparisons with calculations

    International Nuclear Information System (INIS)

    This paper presents and discusses the results of core bowing experiments performed with an ex-reactor rig holding a half hexagon array of 22 sub-assemblies (S/As) simulating the Japanese DFBR conditions and the comparisons of the measured results with calculations by individually developed codes--ARKAS, RAINBOW, SANBOW. The main conclusions of this study are (1) interwrapper loads and S/A displacements within the array were measured at selected positions for a series of five tests simulating the DFBR core bowing modes, (2) the overall comparison between the non-friction calculation and measurement showed good agreement for loads, displacements and their directions, and (3) validation of the friction algorithm has also been carried out and further improvement of the agreement was obtained

  20. Evaluation method for core thermohydraulics during natural circulation in fast reactors numerical predictions of inter-wrapper flow

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, H.; Kimura, N.; Miyakoshi, H. [Japan Nuclear Cycle Development Institute, Reactor Engineering Group, O-arai Engineering Center, Ibaraki (Japan); Nagasawa, K. [Nuclear Energy System Incorporation, O-arai Office, Ibaraki (Japan)

    2001-07-01

    Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. As a decay heat removal system, direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this design, dumped heat exchanger provides cold sodium and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such natural circulation phenomena in a reactor core has been developed, which models each subassembly as a rectangular duct with gap region between the subassemblies and also the upper plenum in a reactor vessel. This numerical simulation method was verified based on experimental data of a sodium test using 7- subassembly core model and also a water test which simulates IWF using the 1/12 sector model of a reactor core. We applied the estimation method to the natural circulation in a 600 MW class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. It is desired for prediction methods on the natural circulation to simulate these phenomena. (author)

  1. Responses of the biogas process to pulses of oleate in reactors treating mixtures of cattle and pig manure

    DEFF Research Database (Denmark)

    Nielsen, Henrik Bjørn; Ahring, Birgitte Kiær

    2006-01-01

    The effect of oleate on the anaerobic digestion process was investigated. Two thermophilic continuously stirred tank reactors (CSTR) were fed with mixtures of cattle and pig manure with different total solid (TS) and volatile solid (VS) content. The reactors were subjected to increasing pulses...

  2. A study on improving the performance of a research reactor's equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2013-01-01

    Full Text Available Utilizing low enriched uranium silicide fuel (U3Si2-Al of existing uranium density (3.285 g/cm3, different core configurations have been studied in search of an equilibrium core with an improved performance for the Pakistan Research Reactor-1. Furthermore, we have extended our analysis to the performance of higher density silicide fuels with a uranium density of 4.0 and 4.8 U g/cm3. The criterion used in selecting the best performing core was that of “unit flux time cycle length per 235U mass per cycle”. In order to analyze core performance by improving neutron moderation, utilizing higher-density fuel, the effect of the coolant channel width was also studied by reducing the number of plates in the standard/control fuel element. Calculations employing computer codes WIMSD/4 and CITATION were performed. A ten energy group structure for fission neutrons was used for the generation of microscopic cross-sections through WIMSD/4. To search the equilibrium core, two-dimensional core modelling was performed in CITATION. Performance indicators have shown that the higher-density uranium silicide-fuelled core (U density 4.8 g/cm3 without any changes in standard/control fuel elements, comprising of 15 standard and 4 control fuel elements, is the best performing of all analyzed cores.

  3. Characteristic differences of LEU and HEU cores at the German FRJ-2 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nabbi, R.; Wolters, J.; Damm, G. [Central Research Reactor Division, Forschungszentrum Juelich, 52425 Juelich (Germany)

    2002-07-01

    As a sophisticated computational method for reactor physics analysis and fuel management an MCNP model in very high fidelity was developed and coupled with a depletion code and applied to the HEU-LEU core conversion study. The analysis show that as a consequence of the high amount of U-238, the amount of U-235 in the LEU core is about 14% higher than in the HEU core. The reduction of the thermal flux varies between 16% (core) and 5% in the reflector zone. The rate of U-235 burnup in the LEU core is approx. 11.5% lower which allows an extension of irradiation time. Due to the effect of neutron spectrum the worth of the absorber system decreases in an LEU core by 17% resulting in a decrease of shutdown and excess reactivity. The kinetic parameters of the core are slightly reduced causing changes in the reactivity values and transient behavior of the core. The moderator coefficient is decreased by 18% and the Doppler coefficient is increased by 63%. Due to shortening of the absorption length of the fission neutrons the prompt neutron lifetime is reduced by 7%. (author)

  4. Analyzing the thermionic reactor critical experiments. [thermal spectrum of uranium 235 core

    Science.gov (United States)

    Niederauer, G. F.

    1973-01-01

    The Thermionic Reactor Critical Experiments (TRCE) consisted of fast spectrum highly enriched U-235 cores reflected by different thicknesses of beryllium or beryllium oxide with a transition zone of stainless steel between the core and reflector. The mixed fast-thermal spectrum at the core reflector interface region poses a difficult neutron transport calculation. Calculations of TRCE using ENDF/B fast spectrum data and GATHER library thermal spectrum data agreed within about 1 percent for the multiplication factor and within 6 to 8 percent for the power peaks. Use of GAM library fast spectrum data yielded larger deviations. The results were obtained from DOT R Theta calculations with leakage cross sections, by region and by group, extracted from DOT RZ calculations. Delineation of the power peaks required extraordinarily fine mesh size at the core reflector interface.

  5. Assessment of CANDU reactor physics effects using a simplified whole-core MCNP model

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

    A whole-core Monte Carlo n-particle (MCNP) model of a simplified CANDU reactor was developed and used to study core configurations and reactor physics phenomena of interest in CANDU safety analysis. The resulting reactivity data were compared with values derived from corresponding WIMS-AECL/RFSP, two-neutron-energy-group diffusion theory core simulations, thereby extending the range of CANDU-related code-to-code benchmark comparisons to include whole-core representations. These comparisons show a systematic discrepancy of about 6 mk between the respective absolute k{sub eff} values, but very good agreement to within about -0.15 {+-} 0.06 mk for the reactivity perturbation induced by G-core checkerboard coolant voiding. These findings are generally consistent with the results of much simpler uniform-lattice comparisons involving only WIMS-AECL and MCNP. In addition, MCNP fission-energy tallies were used to evaluate other core-wide properties, such as fuel bundle and total-channel power distributions, as well as intra-bundle details, such as outer-fuel-ring relative power densities and outer-ring fuel element azimuthal power variations, which cannot be determined directly from WIMS-AECL/RFSP core calculations. The average MCNP values for the ratio of outer fuel element to average fuel element power density agreed well with corresponding values derived from WIMS-AECL lattice-cell cases, showing a small systematic discrepancy of about 0.5 %, independent of fuel bum-up. For fuel bundles containing the highest-power fuel elements, the maximum peak-to-average outer-element azimuthal power variation was about 2.5% for cases where a statistically significant trend was observed, while much larger peak-to-average outer-element azimuthal power variations of up to around 42% were observed in low-power fuel bundles at the core/radial-neutron-reflector interface. (author)

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

    International Nuclear Information System (INIS)

    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

  7. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

  8. 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.

  9. Modular high-temperature gas-cooled reactor core heatup accident simulations

    International Nuclear Information System (INIS)

    The design features of the modular high-temperature gas-cooled reactor (HTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. Simulations of long-term loss-of-forced-convection (LOFC) accidents, both with and without depressurization of the primary coolant and with only passive cooling available to remove afterheat, have shown that maximum core temperatures stay below the point at which fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. 4 refs., 5 figs

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

    International Nuclear Information System (INIS)

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

  11. Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

    International Nuclear Information System (INIS)

    Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

  12. Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard Schultz

    2012-04-01

    Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

  13. Core design calculations of IRIS reactor using modified CORD-2 code package

    International Nuclear Information System (INIS)

    Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

  14. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    International Nuclear Information System (INIS)

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  15. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Performance studies of a new core cooling monitor in a boiling water reactor

    International Nuclear Information System (INIS)

    Performance studies of a new type of core cooling monitors have been carried out in the Barsebaeck Nuclear Power Station during the operation periods 1988-10-04 to 1989-07-05, 1989-08-03 to 1990-09-05 and 1990-09-28 to 1991-07-04. The results showed that the monitors, which were placed inside the reactor core, are very sensitive to variations of the reactor operating conditions, and that 34 months of irradiation did not influence the signals from the monitors. Experiments were also carried out in a 160 bar loop, where sudden uncovers of the monitors were achieved by decreasing the liquid level of the coolant surrounding the monitors. The experiments included the pressures of 5, 20, 50, 70 and 155 bar, and the responses to uncover were in the ranges between 11 and 82 mV/sec or a total step change of 2 V at typical BWR conditions. This is of the order of two decades higher than the responses from monitors based on thermocouple readings. The monitors can be operated in two modes, the core cooling mode and the temperature mode. In the former mode the electrical current is 3-4 A, and in the latter mode, where the monitor actually serves as a thermometer, the current is in the order of 50-100 mA. In the laboratory the monitors have been studied for temperatures up to 1265 deg. C, which is very useful in case of a severe reactor accident. Thus, during such events the temperatures in the reactor core could be followed up to this level and the monitors could also be used to activate certain safety equipment. The function as well as the design of the instrument is verified in laboratory experiments, computer calculations and reactor tests and is now ready for implementation in the BWR instrumentation. In summary: 1. The proposed monitor can operate in two modes; the core cooling mode and the temperature mode. 2. Laboratory studies have shown that the responses to uncover are two decades higher than signals from monitors based on thermocouple readings. 3. No effects of

  17. The status of studies on fast reactor core thermal hydraulics at PNC

    International Nuclear Information System (INIS)

    An outlook was addressed on investigative activities of the fast reactor core thermal-hydraulics at Power Reactor and Nuclear Fuel Development Corporation. Firstly, a computational modeling to predict flow field under natural circulation decay heat removal condition using multi-dimensional codes and its validation were presented. The validation was carried out through calculations of sodium experiments on an inter-subassembly heat transfer, a transient from forced to natural circulation and an inter-wrapper flow. Secondly, experimental and computational studies were expressed on local blockage with porous media in a fuel subassembly. Lastly, information was presented on an advanced computational code based on a subchannel analysis code. The code is under the development and extended to perform whole core simulation. (author)

  18. A study of the advancement of a reactor core design environment

    International Nuclear Information System (INIS)

    Full text: During the years from 2002 to 2004 a joint project has been performed by IFE, Halden and Yonden Engineering Corporation, Japan, to develop an advanced reactor core design environment based on a communication method for controlling a reactor core code system efficiently from PCs in a distributed network. The advanced reactor core design environment is realized by using Microsoft Visual Basic and communication software based on the IFE product SoftwareBus. The project has been carried out based on the fact that a computer-aided design system has been under development at Yonden Engineering Corporation in order to perform efficiently fuel replacement calculation by Yonden's reactor design code system. In this system, the structure is such that the physics calculation code system runs on UNIX workstations (in parallel) performing the calculations, while the Man-Machine Interface for controlling the calculation programs run on PCs in a distributed network. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general communication tool (IFE's SoftwareBus) has been used for realizing communication of the n-pair n-node between the reactor core design code system and the PC applications. Further, a method of improvement in the speed of the optimal pattern calculation has been implemented by assigning each examination pattern to two or more computers distributed in the network and assigning the next pattern calculation to the computer, where the calculation has ended or has the lowest workload. The high-speed technology of the pattern survey by network distributed processing is based on SoftwareBus. The reactor core design code system is developed in FORTRAN running on a UNIX workstation (Solaris). The PC applications have been developed by using Microsoft Visual Basic on Windows 2000 platform. The first step of the verification and validation process was carried out in March

  19. Neutronic flux stability of production uranium graphite reactor conversion core relative to high-frequency oscillations

    International Nuclear Information System (INIS)

    Preliminary methodical simplified investigation into stability of the neutron field in the conversion load of industrial uranium-graphite reactors with regard to basic characteristics of the load in transient processes was carried out. Analysis was based on the calculated research into the behaviour of simplified single-point and one-dimensional models of the reactor core in transient regimes during the interconnected description of dynamics of neutron-physical and thermal properties of the load. Fundamental assumptions on the reactor characteristics used in the calculated model. In the context of accepted approximations the obtained results preclude the possibility for the occurrence of spontaneous high frequency oscillations resulting from the positive reactivity effect on the fuel temperature in the conversion load

  20. Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J. [Serco, Rutherford House, Quedgeley, Gloucester, Gl2 4NF (United Kingdom); Seren, T.; Lipponen, M.; Kekki, T. [VTT, Technical Research Centre of Finland, Otakaari 3 K, P.O. BOX 1000, Espoo, FI-02044 (Finland)

    2011-07-01

    Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

  1. Dependence of core heating properties on heating pulse duration and intensity

    Science.gov (United States)

    Johzaki, Tomoyuki; Nagatomo, Hideo; Sunahara, Atsushi; Cai, Hongbo; Sakagami, Hitoshi; Mima, Kunioki

    2009-11-01

    In the cone-guiding fast ignition, an imploded core is heated by the energy transport of fast electrons generated by the ultra-intense short-pulse laser at the cone inner surface. The fast core heating (˜800eV) has been demonstrated at integrated experiments with GEKKO-XII+ PW laser systems. As the next step, experiments using more powerful heating laser, FIREX, have been started at ILE, Osaka university. In FIREX-I (phase-I of FIREX), our goal is the demonstration of efficient core heating (Ti ˜ 5keV) using a newly developed 10kJ LFEX laser. In the first integrated experiments, the LFEX laser is operated with low energy mode (˜0.5kJ/4ps) to validate the previous GEKKO+PW experiments. Between the two experiments, though the laser energy is similar (˜0.5kJ), the duration is different; ˜0.5ps in the PW laser and ˜ 4ps in the LFEX laser. In this paper, we evaluate the dependence of core heating properties on the heating pulse duration on the basis of integrated simulations with FI^3 (Fast Ignition Integrated Interconnecting) code system.

  2. Simulation of microwave pulsing in a radial line slot antenna etch process reactor

    Science.gov (United States)

    Upadhyay, Rochan; Ishibashi, Kiyotaka; Raja, Laxminarayan

    2013-09-01

    The radial line slot antenna reactor couples the microwave power to a process plasma through a slot antenna. This arrangement leads to efficient generation of plasma with high electron energy adjacent to the window and a lower energy near wafer surfaces. This arrangement is beneficial for low ion energy applications such as soft etching or thin film processing. With increased charge densities, charge-up damage of dielectric surfaces can be a problem that can be addressed though plasma pulsing strategies in electronegative feed gases. The periodic power-off cycle results in an afterglow where electron attachment forms large amounts of negative ions that when extracted to the wafer surface, reduces the effects of positive charge trapping on the surface. We use computational modeling to investigate the effect of microwave pulsing on the negative ion generation rates for high density HBr and CF4 plasmas. We discuss improvements to a plasma chemistry mechanism for the pulsed plasma regime. Our results verify much larger negative ion to electron density ratios compared to the continuous (un-pulsed) case for both HBr and CF4 gases. Results also indicate greater plasma uniformity due to diffusion of positive and negative ions during the power-off phase of the pulse.

  3. Comments on ``large enhancement of TLD-100 sensitivity by irradiation in a reactor core''

    Science.gov (United States)

    Lakshmanan, A. R.; Chandra, Bhuwan; Bhatt, R. C.

    1987-06-01

    The large enhancement of TLD-100 sensitivity on irradiation in a reactor core reported by Lau et al. [Nucl. Instr. and Meth. B17 (1986) 170] is false and is in complete contradiction with the results reported earlier in the literature and with our recent findings. Lau et al. have misinterpreted the TL signal from thermal neutron induced 3H betas in LiF as due to enhanced TL sensitivity because of neutron induced traps/luminescent centres.

  4. Fusion-power-core design of a Compact Reversed-Field Pinch Reactor (CRFPR)

    Science.gov (United States)

    Copenhaver, C.; Schnurr, N. M.; Krakowski, R. A.; Hagenson, R. L.; Mynard, R. C.; Cappiello, C.; Lujan, R. E.; Davidson, J. W.; Chaffee, A. D.; Battat, M. E.

    A conceptual design of a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils) based on a Reversed-Field Pinch (RFP) has been completed. After a brief statement of rationale and description of the reactor configuraton, the FPC integration is described in terms of power balance, thermal-hydraulics, and mechanical design. The engineering versatility, promise, and problems of this high-power-density approach to fusion are addressed.

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

    International Nuclear Information System (INIS)

    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

  6. Modular assembly for supporting, straining, and directing flow to a core in a nuclear reactor

    International Nuclear Information System (INIS)

    A reactor core support arrangement for supporting, straining, and providing fluid flow to the core and periphery of a nuclear reactor during normal operation is described. A plurality of removable inlet modular units are contained within permanent liners in the lower supporting plate of the reactor vessel lower internals. During normal operation (1) each inlet modular unit directs main coolant flow to a plurality of core assemblies, the latter being removably supported in receptacles in the upper portion of the modular unit and (2) each inlet modular unit may direct bypass flow to a low pressure annual region of the reactor vessel. Each inlet modular unit may include special fluid seals interposed between mating surfaces of the inlet modular units and the core assemblies and between the inlet modular units and the liners, to minimize leakage and achieve a hydraulic balance. Utilizing the hydraulic balance, the modular units are held in the liners and the assemblies are held in the modular unit receptacles by their own respective weight. Included as part of the permanent liners below the horizontal support plate are generally hexagonal axial debris barriers. The axial debris barriers collectively form a bottom boundary of a secondary high pressure plenum, the upper boundary of which is the bottom surface of the horizontal support plate. Peripheral liners include radial debris barriers which collectively form a barrier against debris entry radially. During normal operation primary coolant inlet openings in the liner, below the axial debris barriers, pass a large amount of coolant into the inlet modular units, and secondary coolant inlet openings in the portion of the liners within the secondary plenum pass a small amount of coolant into the inlet modular units. The secondary coolant inlet openings also provide alternative coolant inlet flow paths in the unlikely event of blockage of the primary inlet openings

  7. Kinetics of vinyl acetate emulsion polymerization in a pulsed tubular reactor: comparison between experimental and simulation results

    Directory of Open Access Journals (Sweden)

    Sayer C.

    2002-01-01

    Full Text Available A new reactor, the pulsed sieve plate column (PSPC, was developed to perform continuous emulsion polymerization reactions. This reactor combines the enhanced flexibility of tubular reactors with the mixing behavior provided by sieved plates and by the introduction of pulses that is important to prevent emulsion destabilization. The main objective of this work is to study the kinetics of vinyl acetate (VA emulsion polymerization reactions performed in this PSPC. Therefore, both experimental studies and reaction simulations were performed. Results showed that it is possible to obtain high conversions with rather low residence times in the PSPC.

  8. Application of assemblies of in-core instruments of the emergency process instrumentation system. The reactor pressure vessel coolant level sensor in Pressurized-Water Reactors

    International Nuclear Information System (INIS)

    Using as an example the coolant level sensor within SVRD.KNITU assembly, the report deals with a possibility of using assemblies of in-core instruments (SVRD), which ensure the application of in-core monitoring system in normal operation conditions, in the emergency process instrumentation system. Assemblies portrayed in the present report are designed and operated in the Russian-built pressurized water reactors and uranium-graphite channel reactors. However, the philosophy of their design is such that an assembly can be easily adapted to reactors of other types

  9. Analysis of three-dimensional thermo-hydraulic phenomena in the reactor core of LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Hu, S.; Lee, Y. B.; Jang, W. P.; Ha, K. S.; Jung, H. Y. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    The mismatch between power and flow under the transient condition of LMFBR (Liquid Metal cooled Fast Breeder Reactor) core results in thermal stratification in hot pool. Since the fluid of hot pool enters IHXs, the temperature distribution of hot pool can alter the overall system response, therefore three-dimensional analysis of thermo-hydraulic phenomena is necessary. In this study, the thermo-hydraulic phenomena under normal operating condition and unprotected transient condition of LMFBR is investigated using which is the three-dimensional analysis code, COMMIX-1AR/P. The basic input data is based on the design data of KALIMER-600, which is sodium-cooled fast breeder reactor developed by KAERI. COMMIX-1AR/P code has not a reactivity model and the power and core flowrate must be supplied in the input data. In this study, results of SSC-K calculation is used. The temperature and velocity distributions are calculated and compared with those of SSC-K calculation results. The UTOF(Unprotected Loss Of Flow) accident is calculated using COMMIX-1AR/P and the temperature and velocity distributions in the total reactor core are calculated and the natural circulation mode under this transient condition is investigated.

  10. Benchmark Evaluation of the Neutron Radiography (NRAD) Reactor Upgraded LEU-Fueled Core

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2001-09-01

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. The final upgraded core configuration with 64 fuel elements has been completed. Evaluated benchmark measurement data include criticality, control-rod worth measurements, shutdown margin, and excess reactivity. Dominant uncertainties in keff include the manganese content and impurities contained within the stainless steel cladding of the fuel and the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 nuclear data are approximately 1.4% greater than the benchmark model eigenvalue, supporting contemporary research regarding errors in the cross section data necessary to simulate TRIGA-type reactors. Uncertainties in reactivity effects measurements are estimated to be ~10% with calculations in agreement with benchmark experiment values within 2s. The completed benchmark evaluation de-tails are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Experiments (IRPhEP Handbook). Evaluation of the NRAD LEU cores containing 56, 60, and 62 fuel elements have also been completed, including analysis of their respective reactivity effects measurements; they are also available in the IRPhEP Handbook but will not be included in this summary paper.

  11. Neutron leakage treatment in reactor physics. Consequences for predicting core characteristics

    International Nuclear Information System (INIS)

    New generations of simulation tools responding to the challenges brought by the advanced features of both 3rd+ generation Pressurized Water Reactor (PWR) cores and 4th generation sodium fast neutron reactor (SFR) cores are taking shape. The developments of new simulations tools are also motivated by strict requirements of nuclear safety authorities. The new tools have the objective of setting new reference standards for neutronic prediction and will take advantage of innovative algorithms which have been implemented in existing CEA codes, such as ERANOS (fast reactors) and APOLLO2 (PWR); the new codes should at the same time remove remaining calculation errors. Although innovative algorithms have been filling the gaps which did exist 40 years ago between tools specifically dedicated to either thermal neutron cores or fast neutron ones, there remains a series of algorithms which deserve particular attention: the treatment of leakage in cell calculations. This paper describes methods for treating neutron leakage in self-shielding calculations with the sub-group method, and in the cell balance calculation. Applications of the MOC method of solution to treat neutron leakage are described. The application of the MOC can eliminate approximations at the cell interfaces while maintaining precise neutron leakage treatment. The new APOLLO3® code, presently under development at CEA, is candidate for hosting such algorithms. (author)

  12. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    International Nuclear Information System (INIS)

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

  13. Parametric evaluation of mixed (low and high enriched) fuel core for a swimming pool type research reactor

    International Nuclear Information System (INIS)

    A study has been carried out to evaluate the performance of a swimming pool type research reactor core comprised of mixed (low and high enriched) uranium fuel. The study includes the calculations of core reactivity, worth of control rods and core criticality at the Beginning Of Life (BOL) of the core and for two operating conditions Cold Zero Power (CZP) and Hot Full Power (HFP). Further, to ensure safe and stable operation of the core from nuclear design point of view, average power densities in the fuel region, power peaking factors, axial power distribution in the hot channel and reactivity feed back coefficients have also been calculated. Two group fluxes have also been determined at different irradiation locations. All these calculations have been carried out employing reactor lattice code WIMS-D14 and reactor analysis code CITATION The calculated results show reasonably good agreement with the quoted operational data of the previous LEU cores. (Author)

  14. Assessment of core structural materials and surveillance programme of research reactors in Egypt

    International Nuclear Information System (INIS)

    The main structural materials to be used in the reactor core, support structures are stainless steel, aluminum and zirconium alloys (zircadyne). Other materials are also used, for example such as polymers in seals and protective coating, and hafnium (HF) as absorber materials in the control rod plates. Stainless steel is used for the reactor pool. The mechanical properties of stainless steel alloys change when they are subjected to irradiation. The main phenomena observed are swelling and irradiation - induced creep. The swelling phenomenon depends on the operating temperature and neutron fluence. For the reactor facility, components will operate at temperature below 70 o C and are expected to see a lifetime fluence of approximately 1 x 1023 n.cm-2.these conditions are well below the conditions where swelling becomes significant. Stainless steels have strong resistance to corrosion over a wide range of environments and temperature. The reactor pool and primary circuit water is demineralized water with controlled low conductivity of less than 100 μ.sm-1 no failure mechanism is known under such process conditions. Aluminum alloys will be used for the constructions of some reactor internals which working in radiation environment as their properties are well understood and show predictable behavior under such conditions. Aluminum is extensively used in water - cooled research reactors because of its low cross-section for the capture of thermal neutrons, excellent corrosion resistance and thermal conductivity. Irradiation damage of polymers strongly depends on the fluence received by the materials. Irradiation effects of polymers also depend on their compositions and molecular structure. if the content of natural rubber is high, irradiation induces an increase in the tensile strength. Where the content of polypropylene is high, irradiation reduces the strength. A materials surveillance plan has been developed and will be implemented from the commencement of reactor

  15. Subchannel analysis of a small ultra-long cycle fast reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Han; Kim, Ji Hyun; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2014-04-01

    Highlights: • The UCFR-100 is small-sized one of 60 years long-life nuclear reactors without refueling. • The design safety limits of the UCFR-100 are evaluated using MATRA-LMR. • The subchannel results are below the safety limits of general SFR design criteria. - Abstract: Thermal-hydraulic evaluation of a small ultra-long cycle fast reactor (UCFR) core is performed based on existing safety regulations. The UCFR is an innovative reactor newly designed with long-life core based on the breed-and-burn strategy and has a target electric power of 100 MWe (UCFR-100). Low enriched uranium (LEU) located at the bottom region of the core play the role of igniter to operate the UCFR for 60 years without refueling. A metallic form is selected as a burning fuel region material after the LEU location. HT-9 and sodium are used as cladding and coolant materials, respectively. In the present study, MATRA-LMR, subchannel analysis code, is used for evaluating the safety design limit of the UCFR-100 in terms of fuel, cladding, and coolant temperature distributions in the core as design criteria of a general fast reactor. The start-up period (0 year of operation), the middle of operating period (30 years of operation), and the end of operating cycle (60 years of operation) are analyzed and evaluated. The maximum cladding surface temperature (MCST) at the BOC (beginning of core life) is 498 °C on average and 551 °C when considering peaking factor, while the MCST at the MOC (middle of core life) is 498 °C on average and 548 °C in the hot channel, respectively, and the MCST at the EOC (end of core life) is 499 °C on average and 538 °C in the hot channel, respectively. The maximum cladding surface temperature over the long cycle is found at the BOC due to its high peaking factor. It is found that all results including fuel rods, cladding, and coolant exit temperature are below the safety limit of general SFR design criteria.

  16. Research and development of a super fast reactor. (2) Core design improvement on local void reactivity

    International Nuclear Information System (INIS)

    A 700MWe Supercritical-pressure water-cooled fast reactor (Super Fast Reactor) was designed with negative overall void reactivity. As there is no cross flow between the fuel assemblies, the local void reactivity, defined as the reactivity change when the coolant of one assembly disappears, also need to be kept negative throughout the cycle. In this study, we found out the mechanism of the local void reactivity and improved the core design to keep the local void reactivity negative for all the seed fuel assemblies. According to the theory analysis, several core configurations, including the thickness of ZrH layer, the layout of the seed fuel assembly, the layout of the core and the loading pattern, will affect the local void reactivity distribution. Sensitivity of those configurations on the local void reactivity was analyzed. 1.15cm of ZrH layer thickness is the best choice for reducing the local void reactivity for the current core design. The assembly layout has no obvious effect on the local void reactivity. It is necessary to load more blanket assemblies in the inner region of the core in order to reduce the local void reactivity of the inner seed fuel assemblies. Loading pattern is also important for flattening the local void reactivity distribution. A hybrid loading method can be employed to make the distribution of the local void reactivity more uniform. Based on those conclusions, a Super Fast Reactor is successfully designed with satisfying all of the design criteria and design goals as well as keeping the local void reactivity of all the seed fuel assemblies less than -30pcm. (author)

  17. Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion

    Science.gov (United States)

    Sayre, Edwin D.; Ring, Peter J.; Brown, Neil; Elsner, Norbert B.; Bass, John C.

    2008-01-01

    This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter and connect to the radiator. The solid core reactor is made from HT9 alloy. The fuel is uranium oxide with 90% enrichment. The thermoelectric converter is bonded to the outside of the 1.10 inch ID heat pipe and is 30 inches long. The thermoelectric couple is Si/SiGe-Si/SiC Quantum Well with over 20% efficiency with an 890 K hot side and a 490 K cold side and produces 625 Watts. 176 converters produce 110 kWe. With less than 10% loss in controls this yields 100 kWe for use. The cylindrical thermoelectric converter is designed and fabricated by HIPing to keep brittle materials in compression and to ensure conductivity. The solid core is fabricated by machining the heatpipe tubes with 6 grooves that are diffusion bonded together by HIPing to form the fuel tubes. The maximum temperature of the heat pipes is 940 K and the return flow temperature is 890 K. The reactor core is hexagonal shaped, 61 cm. wide and 76.2 cm high with 12 rotating control drums surrounding it. There is shielding to protect components and human habitation. The radiator is daisy shaped at 45 degrees with each petal 5.5 meters long. The design life is ten years.

  18. Unconstrained pulse pressure monitoring for health management using hetero-core fiber optic sensor

    Science.gov (United States)

    Nishiyama, Michiko; Sonobe, Masako; Watanabe, Kazuhiro

    2016-01-01

    In this paper, we present a pulse pressure waveform sensor that does not constrain a wearer’s daily activity; the sensor uses hetero-core fiber optics. Hetero-core fiber sensors have been found to be sensitive to moderate bending. To detect minute pulse pressure changes from the radial artery at the wrist, we devised a fiber sensor arrangement using three-point bending supports. We analyzed and evaluated the measurement validity using wavelet transformation, which is well-suited for biological signal processing. It was confirmed that the detected pulse waveform had a fundamental mode frequency of around 1.25 Hz over the time-varying waveform. A band-pass filter with a range of frequencies from 0.85 to 1.7 Hz was used to pick up the fundamental mode. In addition, a high-pass filter with 0.85 Hz frequency eliminated arm motion artifacts; consequently, we achieved high signal-to-noise ratio. For unrestricted daily health management, it is desirable that pulse pressure monitoring can be achieved by simply placing a device on the hand without the sensor being noticed. Two types of arrangements were developed and demonstrated in which the pulse sensors were either embedded in a base, such as an armrest, or in a wearable device. A wearable device without cuff pressure using a sensitivity-enhanced fiber sensor was successfully achieved with a sensitivity of 0.07–0.3 dB with a noise floor lower than 0.01 dB for multiple subjects.

  19. Detailed analysis for a control rod worth of the gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    GTHTR300 is composed of a simplified and economical power plant based on an inherent safe 600 MWt reactor and a nearly 50% high efficiency gas turbine power conversion cycle. GTHTR300 core consist of annular fuel region, center and outer side reflectors because of cooling it effectively in depressurized accident conditions, and all control rods are located in both side reflectors of annular core. As a thermal neutron spectrum is strongly distorted in reflector regions, an accurate calculation is especially required for the control rod worth evaluation. In this study, we applied the detailed Monte Carlo calculations of a full core model, and confirmed that our design method has enough accuracy. (author)

  20. Detailed analysis for a control rod worth of the gas turbine high temperature reactor (GTHTR300)

    Energy Technology Data Exchange (ETDEWEB)

    Nakata, Tetsuo; Katanishi, Shoji; Takada, Shoji; Yan, Xing; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-11-01

    GTHTR300 is composed of a simplified and economical power plant based on an inherent safe 600 MWt reactor and a nearly 50% high efficiency gas turbine power conversion cycle. GTHTR300 core consist of annular fuel region, center and outer side reflectors because of cooling it effectively in depressurized accident conditions, and all control rods are located in both side reflectors of annular core. As a thermal neutron spectrum is strongly distorted in reflector regions, an accurate calculation is especially required for the control rod worth evaluation. In this study, we applied the detailed Monte Carlo calculations of a full core model, and confirmed that our design method has enough accuracy. (author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  3. Influence of remaining fission products in low-decontaminated fuel on reactor core characteristics

    International Nuclear Information System (INIS)

    Design study of core, fuel and related fuel cycle system with low-decontaminated fuel has been performed in the framework of the feasibility study (F/S) on commercialized fast reactor cycle systems. This report summarizes the influence on core characteristics of remaining fission products (FPs) in low-decontaminated fuel related to the reprocessing systems nominated in F/S phase I. For simple treatment of the remaining FPs in core neutronics calculation the representative nuclide method parameterized by the FP equivalent coefficient and the FP volume fraction was developed, which enabled an efficient evaluation procedure. As a result of the investigation on the sodium cooled fast reactor with MOX fuel designed in fiscal year 1999, it was found that the pyrochemical reprocessing with molten salt (the RIAR method) brought the largest influence. Nevertheless, it was still within the allowable range. Assuming an infinite-times recycling, the alternations in core characteristics were evaluated as follows: increment of burnup reactivity by 0.5%Δk/kk', decrement of breeding ratio by 0.04, increment of sodium void reactivity by 0.1x10-2Δk/kk' and decrement of Doppler constant (in absolute value) by 0.7x10-3 Tdk/dT. (author)

  4. Investigation on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin

    2013-10-22

    Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: • Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. • Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. • Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. • Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the

  5. Study of core flow distribution for small modular natural circulation lead or lead-alloy cooled fast reactors

    International Nuclear Information System (INIS)

    Highlights: • A core flow distribution calculation code for natural circulation LFRs was developed. • The comparison study between the channel method and the CFD method was conducted. • The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted. - Abstract: Small modular natural circulation lead or lead-alloy cooled fast reactor (LFR) is a potential candidate for LFR development. It has many attractive advantages such as reduced capital costs and inherent safety. The core flow distribution calculation is an important issue for nuclear reactor design, which will provide important input parameters to thermal-hydraulic analysis and safety analysis. The core flow distribution calculation of a natural circulation LFR is different from that of a forced circulation reactor. In a forced circulation reactor, the core flow distribution can be controlled and adjusted by the pump power and the flow distributor, while in a natural circulation reactor, the core flow distribution is automatically adjusted according to the relationship between the local power and the local resistance feature. In this paper, a non-uniform heated parallel channel flow distribution calculation code was developed and the comparison study between the channel method and the CFD method was carried out to assess the exactness of the developed code. The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted using the developed code. A core flow distribution optimization design scheme for a 10MW natural circulation LFR was proposed according to the optimization analysis results

  6. The study of core melting phenomena in reactor severe accident of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Jeun, Gyoo Dong; Park, Seh In; Lim, Jae Hyuck; Park, Seong Yong [Hanyang Univ., Seoul (Korea, Republic of); Bang, Kwang Hyun; Kim, Ki Yong [Korea Maritime Univ., Busan (Korea, Republic of)

    1999-03-15

    After TMI-2 accident, it has been paid much attention to severe accidents beyond the design basis accidents and the research on the progress of severe accidents and mitigation and the closure of severe accidents has been actively performed. In particular, a great deal of uncertainties yet exist in the phase of late core melt progression and thus the research on this phase of severe accident progress has a key role in obtaining confidence in severe accident mitigation and nuclear reactor safety. In the present study, physics of late core melt progression, experimental data and the major phenomenological models of computer codes are reviewed and a direction of reducing the uncertainties in the late core melt progression is proposed.

  7. Development of the evaluation methods in reactor safety analyses and core characteristics

    International Nuclear Information System (INIS)

    In order to support the safety reviews by NISA on reactor safety design, the computer codes are developed and maintained in the areas of safety analyses and core characteristics evaluation. In the code preparation of safety analyses, the TRACE code was prepared to conduct the safety analysis of LOCA for PWR. Also, the statistical safety evaluation method based on the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH/TRACE has been prepared. In the core physics code preparation, the advanced neutron data library JENDL-4.0 were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 /SIMULATE-3 was continued by using core physics data. (author)

  8. Development of the evaluation methods in reactor safety analyses and core characteristics

    International Nuclear Information System (INIS)

    In order to support the safety reviews by NISA on reactor safety design, the computer codes are developed and maintained in the areas of safety analyses and core characteristics evaluation. In the code preparation of safety analyses, the TRACE code was prepared to conduct the safety analysis of LOCA for APWR. Also, the statistical safety evaluation method based on the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH/TRACE was prepared. In the core physics code preparation, the advanced neutron data library JENDL-3.3 and expanded nuclide chains were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 was continued by using core physics data. (author)

  9. Optimisation of the Core Management Scenario to Reach High Fuel Burnup in the MYRRHA Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abderrahim, H. Ait [General Management, Belgian Nuclear Research Centre SCK CEN, Boeretang 200, BE-2400, Mol (Belgium); Baeten, P.; Eynde, G. Van Den [Advanced Nuclear Systems, Belgian Nuclear Research Centre SCK CEN, Boeretang 200, BE-2400, Mol (Belgium); Sobolev, V. [Nuclear Materials Science, Belgian Nuclear Research Centre SCK CEN, Boeretang 200, BE-2400, Mol (Belgium); Nishihara, K. [Center for Neutron Science, Japanese Agency for Atomic Energy (JAEA), Muramatsu 124-2, 319-1112 Tokai-Mura, Ibaraki-Ken (Japan)

    2011-07-01

    An innovative fast spectrum experimental facility MYRRHA has being developed by the Belgian Nuclear Research Centre SCK CEN. The MYRRHA is an accelerator-driven system with a core loaded with fast reactor MOX fuel and cooled by liquid lead-bismuth eutectic. At this stage the selection of the facility operation mode and the fuel management scenario is of great importance. In the present article two different modes of the MYRRHA core management are compared: at constant power and at constant proton beam current. The results of neutronic and thermo-mechanical modeling are presented. It is shown that safer thermomechanical conditions for the fuel elements are predicted in the case of the core reshuffling with batches of ten fuel assemblies and with the ADS operation mode at a constant proton beam current. (author)

  10. Feasibility study of core management system by data communication for boiling water reactors

    International Nuclear Information System (INIS)

    Core management system by data communication has been designed and proposed for more efficient operation of BWR plants by faster transmission and centralized management of information system comprises three kinds of computers: process computer for monitoring purpose at reactor site, center computer for administration purpose at head office and large scientific computer for planning and evaluation purpose. The process and the large computers are connected to the center computer by data transmission line. To demonstrate the feasibility of such a system, operating history evaluation system, which is one of the subsystems of the core management system has been developed along the above concept. Application to the evaluation of operating history of a commercial BWR shows a great deal of merits. Quick response and considerably large amount of reduction of manpower can be expected by data communication and minimized intervention of human labor. Visual display is also found to be very useful to understand the core characteristics

  11. Annular pancreas (image)

    Science.gov (United States)

    Annular pancreas is an abnormal ring or collar of pancreatic tissue that encircles the duodenum (the part of the ... intestine that connects to stomach). This portion of pancreas can constrict the duodenum and block or impair ...

  12. Transient behavior studies of TRIGA core for variations in reactor kinetics

    International Nuclear Information System (INIS)

    Highlights: • Fast reactivity insertion analysis was done for TRIGA reactor. • Clad temperature remains below the design limit for 2$ reactivity at full power. • Peak power seems taken higher values for low initial power level. • Reactor parameters are more sensitive to variation of βeff than lp. • Inherent safe properties of TRIGA plays vital role in reactor safety. - Abstract: This paper illustrates the transient characteristics of TRIGA core at different kinetics conditions that arise from variations of externally inserted reactivity together with variations in other kinetic parameters such as prompt neutron life time, lp and effective delayed neutron fraction, βeff provided the reactor scram system is disabled. From the concern of severity of fast reactivity accident which could lead to most dangerous consequences, the inserted reactivity considered herein was fast reactivity which was a step type within the range 1$–2$ with insertion time 0.5 s. The initial power was 100 W and full power, 3 MW whereby the values of lp and βeff had been kept fixed at their recommended values 30 μs and 0.007, respectively. The observed parameters were reactor peak power and maximum clad temperature of the hottest channel. The analysis infers that clad temperature remains within its design limit value even for the maximum inserted reactivity, 2$ at full power operation of reactor. Also, the peak power took relatively higher values for transients at low power level. For each inserted reactivity, values of lp and βeff were varied individually within certain ranges. In this case, although both power and clad temperature are more sensitive to the variation in effective delayed neutron fraction than to the variation in prompt neutron life time, however, clad temperature remained within its design limit even for the maximum value of inserted reactivity with minimum βeff value considered. Prompt negative temperature coefficient of reactivity that stems from U-ZrH fuel

  13. Linear pulse motor type control element drive mechanism for the integral reactor

    International Nuclear Information System (INIS)

    The integral reactor SMART currently under development at Korea Atomic Energy Research Institute is designed with soluble boron free operation and use of nuclear heating for reactor startup. These design features require the Control Element Drive Mechanism (CEDM) for SMART to have fine-step movement capability as well as high reliability for the fine reactivity control. In this paper, design characteristics of a new concept CEDM driven by the Linear Pulse Motor (LPM) which meets the design requirements of the integral reactor SMART are introduced. The primary dimensions of the linear pulse motor are determined by the electro-magnetic analysis and the results are also presented. In parallel with the electro-magnetic analysis, the conceptual design of the CEDM is visualized and checked for interferences among parts by assembling three dimensional (3D) models on the computer. Prototype of LPM with double air-gaps for the CEDM sub-assemblies to lift 100 kg is designed, analysed, manufactured and tested to confirm the validity of the CEDM design concept. A converter and a test facility are manufactured to verify the dynamic performance of the LPM. The mover of the LPM is welded with ferromagnetic material and non-ferromagnetic material to get the magnetic flux path between inner stator and outer stator. The thrust forces of LPM predicted by analytic model have shown good agreement with experimental results from the prototype LPM. It is found that the LPM type CEDM has high force density and simple drive mechanism to reduce volume and satisfy the reactor operating circumstances with high pressure and temperature

  14. Laser pulse heating of nuclear fuels for simulation of reactor power transients

    Indian Academy of Sciences (India)

    C S Viswanadham; K C Sahoo; T R G Kutty; K B Khan; V P Jathar; S Anantharaman; Arun Kumar; G K Dey

    2010-12-01

    It is important to study the behaviour of nuclear fuels under transient heating conditions from the point of view of nuclear safety. To simulate the transient heating conditions occurring in the known reactor accidents like loss of coolant accident (LOCA) and reactivity initiated accident (RIA), a laser pulse heating system is under development at BARC, Mumbai. As a prelude to work on irradiated nuclear fuel specimens, pilot studies on unirradiated UO2 fuel specimens were carried out. A laser pulse was used to heat specimens of UO2 held inside a chamber with an optically transparent glass window. Later, these specimens were analysed by metallography and X-ray diffraction. This paper describes the results of these studies.

  15. Combined treatment of SO2 and high resistivity fly ash using a pulse energized electron reactor

    International Nuclear Information System (INIS)

    The combined removal of SO2 and high resistivity fly ash has been demonstrated in a pulse energized electron reactor (PEER). The PEER system which was originally developed for the removal of SO2 utilizes a positive pulse streamer corona discharge in a non-uniform field geometry. In performance tests on SO2, more than 90% was removed with an advantageously small power requirement. Combined treatment performance was demonstrated by introducing high resistivity fly ash into the test gas and the PEER is significantly more efficient than a conventional electrostatic precipitator operated with a dc voltage. Observations show that the PEER agglomerates the fly ash and further that the SO2 removal efficiency is improved by the presence of fly ash. The electrode configuration and performance results make retrofit consideration attractive

  16. Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    CERN Document Server

    Karch, J; Beck, M; Eberhardt, K; Hampel, G; Heil, W; Kieser, R; Reich, T; Trautmann, N; Ziegner, M

    2013-01-01

    The performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz with a maximum peak energy of 10 MJ is described. The solid deuterium converter with a volume of V=160 cm3 (8 mol), which is exposed to a thermal neutron fluence of 4.5x10^13 n/cm2, delivers up to 550 000 UCN per pulse outside of the biological shield at the experimental area. UCN densities of ~ 10/cm3 are obtained in stainless steel bottles of V ~ 10 L resulting in a storage efficiency of ~20%. The measured UCN yields compare well with the predictions from a Monte Carlo simulation developed to model the source and to optimize its performance for the upcoming upgrade of the TRIGA Mainz into a user facility for UCN physics.

  17. Advanced nitrogen removal by pulsed sequencing batch reactors (SBR) with real-time control

    Institute of Scientific and Technical Information of China (English)

    YANG Qing; PENG Yongzhen; YANG Anming; GUO Jianhua; LI Jianfeng

    2007-01-01

    The feasibility of pH and oxidation reduction potential (ORP) as on-line control parameters to advance nitrogen removal in pulsed sequencing batch reactors (SBR)was evaluated.The pulsed SBR,a novel operational mode of SBR,was utilized to treat real municipal wastewater accompanied with adding ethanol as external carbon source.It was observed that the bending-point (apex and knee) of pH and ORP profiles can be used to control denitrification process at a low influent C/N ratio while dpH/dt can be used to control the nitrification and denitrification process at a high influent C/N ratio.The experimental results demonstrated that the effluent total nitrogen can be reduced to lower than 2 mg/L,and the average total nitrogen (TN) removal efficiency was higher than 98% by using real-time controll strategy.

  18. Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    Science.gov (United States)

    Karch, J.; Sobolev, Yu.; Beck, M.; Eberhardt, K.; Hampel, G.; Heil, W.; Kieser, R.; Reich, T.; Trautmann, N.; Ziegner, M.

    2014-04-01

    The performance of the solid deuterium ultra-cold neutron (UCN) source at the pulsed reactor TRIGA Mainz with a maximum peak energy of 10MJ is described. The solid deuterium converter with a volume of cm3 (8mol), which is exposed to a thermal neutron fluence of n/cm2, delivers up to 240000 UCN ( m/s) per pulse outside the biological shield at the experimental area. UCN densities of 10 cm3 are obtained in stainless-steel bottles of 10 L. The measured UCN yields compare well with the predictions from a Monte Carlo simulation developed to model the source and to optimize its performance for the upcoming upgrade of the TRIGA Mainz into a user facility for UCN physics.

  19. Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    The performance of the solid deuterium ultra-cold neutron (UCN) source at the pulsed reactor TRIGA Mainz with a maximum peak energy of 10MJ is described. The solid deuterium converter with a volume of V=160 cm3 (8mol), which is exposed to a thermal neutron fluence of 4.5 x 1013 n/cm2, delivers up to 240000 UCN (v ≤ 6 m/s) per pulse outside the biological shield at the experimental area. UCN densities of ∼ 10 cm3 are obtained in stainless-steel bottles of V ∼ 10 L. The measured UCN yields compare well with the predictions from a Monte Carlo simulation developed to model the source and to optimize its performance for the upcoming upgrade of the TRIGA Mainz into a user facility for UCN physics. (orig.)

  20. Cluster analysis for investigation of the dynamics of pulse energy noise at the IBR-2M reactor

    International Nuclear Information System (INIS)

    The results of study of the dynamics of the noise component of IBR-2M using cluster analysis methods are presented. It is shown that spectral density changes of pulse energy fluctuations after the reactor nominal power 2 MW have a transition region of duration ~ 3 days. During the operation of the reactor the noise structure is divided into four stable structures, three of them describing the noise transition region. The fourth stable structure is independent of the reactor's noise level and operation time. The noise transition region is caused by the vibration of moving reflectors in the process of heating after increasing of the reactor power.

  1. Effects of conversion ratio change on the core performances in medium to large TRU burning reactors

    International Nuclear Information System (INIS)

    Conceptual fast reactor core designs with sodium coolant are developed at 1,500, 3,000 and 4,500 MWt which are configured to transmute recycled transuranics (TRU) elements with external feeds consisting of LWR spent fuel. Even at each pre-determined power level, the performance parameters, reactivity coefficients and their implications on the safety analysis can be different when the target TRU conversion ratio changes. In order to address this aspect of design, a study on TRU conversion ratio change was performed. The results indicate that it is feasible to design a TRU burner core to accommodate a wide range of conversion ratios by employing different fuel cladding thicknesses. The TRU consumption rate is found to be proportional to the core power without any significant deterioration in the core performance at higher power levels. A low conversion ratio core has an increased TRU consumption rate and much faster burnup reactivity loss, which calls for appropriate means for reactivity compensation. As for the reactivity coefficients related with the conversion ratio change, the core with a low conversion ratio has a less negative Doppler coefficient, a more negative axial expansion coefficient, a more negative control rod worth per rod, a more negative radial expansion coefficient, a less positive sodium density coefficient and a less positive sodium void worth. A slight decrease in the delayed neutron fraction is also noted, reflecting the fertile U-238 fraction reduction. (author)

  2. Results of Reactor Materials Experiments Investigating 2-D Core-Concrete Interaction and Debris Coolability

    International Nuclear Information System (INIS)

    The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor materials experiments and associated analysis to achieve the following objectives: 1) resolution of the ex-vessel debris coolability issue, and 2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants and provide the technical basis for better containment designs of future plants. With respect to the second objective, there remain uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a first step in bridging this gap, a large scale Core-Concrete Interaction experiment (CCI-1) has been conducted as part of the MCCI program. This test investigated the interaction of a 400 kg core-oxide melt with a crucible made of siliceous concrete along two walls and the base. The two remaining walls were made of non-ablative magnesium oxide. The initial phase of the test was conducted under dry conditions. After a predefined ablation depth was achieved, the cavity was flooded to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a description of the test facility and an overview of results from this test. (authors)

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

    DEFF Research Database (Denmark)

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

    2016-01-01

    setup. It provided 22ps pulses with a maximum average power of 95W, 40MHz repetition rate at 1032nm (~2.4μJ pulse energy), with M2 <1.3. We determined the facet damage threshold for a 7-cells hollow core photonic bandgap fiber and showed up to 59W average power output for a 5 meters fiber. The damage......We demonstrated robust and bend insensitive fiber delivery of high power laser with diffraction limited beam quality for two different kinds of hollow core band gap fibers. The light source for this experiment consists of ytterbium-doped double clad fiber aeroGAIN-ROD-PM85 in a high power amplifier...... threshold for a 19-cell hollow core photonic bandgap fiber exceeded the maximum power provided by the light source and up to 76W average output power was demonstrated for a 1m fiber. In both cases, no special attention was needed to mitigate bend sensitivity. The fibers were coiled on 8 centimeters radius...

  4. Selective Oxidation of Propane by Lattice Oxygen of Vanadium-Phosphorous Oxide in a Pulse Reactor

    Institute of Scientific and Technical Information of China (English)

    Rusong Zhao; Jian Wang; Qun Dong; Jianhong Liu

    2005-01-01

    Selective oxidation of propane by lattice oxygen of vanadium-phosphorus oxide (VPO) catalysts was investigated with a pulse reactor in which the oxidation of propane and the re-oxidation of catalyst were implemented alternately in the presence of water vapor. The principal products are acrylic acid (AA),acetic acid (HAc), and carbon oxides. In addition, small amounts of C1 and C2 hydrocarbons were also found, molar ratio of AA to HAc is 1.4-2.2. The active oxygen species are those adsorbed on catalyst surface firmly and/or bound to catalyst lattice, i.e. lattice oxygen; the selective oxidation of propane on VPO catalysts can be carried out in a circulating fluidized bed (CFB) riser reactor. For propane oxidation over VPO catalysts, the effects of reaction temperature in a pulse reactor were found almost the same as in a steady-state flow reactor. That is, as the reaction temperature increases, propane conversion and the amount of C1+C2 hydrocarbons in the product increase steadily, while selectivity to acrylic acid and to acetic acid increase prior to 350 ℃ then begin to drop at temperatures higher than 350 ℃, and yields of acrylic acid and of acetic acid attained maximum at about 400 ℃. The maximum yields of acrylic acid and of acetic acid for a single-pass are 7.50% and 4.59% respectively, with 38.2% propane conversion. When the amount of propane pulsed decreases or the amount of catalyst loaded increases, the conversion increases but the selectivity decreases. Increasing the flow rate of carrier gases causes the conversion pass through a minimum but selectivity and yields pass through a maximum. In a fixed bed reactor, it is hard to obtain high selectivity at a high reaction conversion due to the further degradation of acrylic acid and acetic acid even though propane was oxidized by the lattice oxygen. The catalytic performance can be improved in the presence of excess propane. Propylene can be oxidized by lattice oxygen of VPO catalyst at 250

  5. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis.

    Science.gov (United States)

    Adam, Zachary R

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 10(5)-10(6) years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

  6. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis.

    Science.gov (United States)

    Adam, Zachary R

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 10(5)-10(6) years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest. PMID:26680444

  7. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis

    Science.gov (United States)

    Adam, Zachary R.

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 105-106 years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

  8. Physical start up of the Dalat nuclear research reactor with the core configuration exempt from neutron trap

    International Nuclear Information System (INIS)

    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

  9. Theoretical methods for neutronics calculations of core-blanket and core-reflector systems in fast reactors

    International Nuclear Information System (INIS)

    The present work is a contribution to the neutronics calculational methods of fast neutron reactors. The first step is devoted to the analysis of the validity of the few-groups (of the order of 25) multigroup scheme, and of the transport-correction approximation for the treatment of the scattering anisotropy. This analysis includes both the reactor core, where the usual approximations are found to be satisfactory, and the reflector, where it turns out that the rapid variations of the neutron flux and of it's spectrum necessitate the improvement of the multigroup cross-sections' generation. Therefore, a zero-dimensional simple and accurate model for the average spectrum in the reflector is developed by the space-energy synthesis method. Finally using the Rayleigh-Ritz method, a model is developed in which the flux is spatially represented by an analytical function. This model is applied to the analysis of the sensitivity of reflector neutronics parameters to the variations of the cross sections

  10. Study on in-core physical design limit zone for lead bismuth eutectic cooled long-life cycle reactor

    International Nuclear Information System (INIS)

    Reactivity variation versus core burnup is a key parameter in neutron design for long-life cycle reactor. The factors affecting reactor core loading pattern are studied from neutronic design point of view, based on the core composed of U-Pu-Zr fuel and lead bismuth eutectic coolant. The methodology for defining in-core physical design limiting zone is given, and by analyzing the effects of key parameters such as initial plutonium content and fuel rod pitch-diameter ratio, the in-core physical design limit zone is defined. Analysis results show that the methodology is appropriate and the limiting area defined in this study satisfies the core depletion and core reactivity control requirement. (authors)

  11. Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-05-01

    It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

  12. Explicit core-follow simulations for a CANDU 6 reactor fuelled with recovered-uranium CANFLEX bundles

    International Nuclear Information System (INIS)

    Recovered uranium (RU) is a by-product of many light-water reactor (LWR) fuel recycling programs. After fission products and plutonium (Pu) have been removed from spent LWR fuel, RU is left. A fissile content in the RU of 0.9 to 1.0% makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficiently high neutron economy to use RU as fuel. Explicit core-follow simulations were run to analyse the viability of RU as a fuel for existing CANDU 6 cores. The core follow was performed with RFSP, using WIMS-AECL lattice properties. During the core follow, channel powers and bundle powers were tracked to determine the operating envelope for RU in a CANFLEX bundle. The results show that RU fits the operating criteria of a generic CANDU 6 core and is a viable fuel option in CANDU reactors. (author)

  13. Fluid structure interaction studies on acoustic load response of light water nuclear reactor core internals under blowdown condition

    International Nuclear Information System (INIS)

    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)

  14. Radiation characterization summary : ACRR 44-inch lead-boron bucket located in the central cavity on the 32-inch pedestal at the core centerline (ACRR-LB44-CC-32-cl).

    Energy Technology Data Exchange (ETDEWEB)

    Parma, Edward J.,; Quirk, Thomas J.; Lippert, Lance L.; Griffin, Patrick Joseph; Naranjo, Gerald E.; Luker, Spencer Michael

    2013-04-01

    This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the 44-inch-long lead-boron bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-LB44-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra are presented as well as radial and axial neutron and gamma-ray flux profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse and steady-state operations are presented with conversion examples.

  15. BGCore - A Comprehensive Package for Reactor Core and Fuel Storage Analysis

    International Nuclear Information System (INIS)

    Recent interest in Fast Gas Cooled Reactors requires major adaptations or evolutions of calculation tools to accommodate the innovative features of core design (new fuel and subassembly forms), fuel composition (homogeneous recycling of minor actinides). Fast neutron spectrum renders inadequate the familiar group reduction schemes and homogenization methods used in LWR analysis. In addition, the specificities of Fast Gas Cooled Reactors (materials, subassembly design, preferential direction for neutron leakage (streaming), high temperatures, particular reactivity effects, etc.) require at least an increase in the number of nuclides to be taken into account in the neutronic libraries with an extended tabulation in temperature. Enhancement of neutronic calculational tools is needed for S/A heterogeneity and anisotropy and to accurately model control elements and other non fueled regions. Several computational systems recently developed are widely used and several others are currently under development. All of the systems are based on Monte-Carlo codes for a 3 Dimensional representation of core and ORIGEN(8) code for fuel composition calculations. This abstract presents the outline and current progress of a development of a comprehensive calculational system for Fast Gas Cooled Reactors carried out in Ben-Gurion University. The flow chart of the system (temporarily) designated as BG-CORE is shown in Fig. 1. Our approach follows, in general, that adapted in other systems, i.e. interfacing the core model (MCNP) with a SARAF - an independently developed code for calculating fuel composition in-core and spent fuel emissions following discharge. Two novel features are added: temperature distributions and feedback, and fuel management. The following modules comprise the BG-CORE system: MCNP - Monte-Carlo code for 3 dimensional core representation using cross-section data sets based on JEF-2/JEF-3 and ENDFB-VI. This module provides Kerf, flux and power density distributions

  16. Measurements of the neutron energy spectra in the core of IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    This work presents the neutron spectrum measurements in the Reactor IPEN/MB-01 using very thin activation detectors in the metallic form, in reactor core, in moderator region. An articulated device allows that the foils are inserted in the central position of reactor core, ensuring that all the foils are irradiated in the same position. The activation detectors of different materials such Au197, Mg24, Ti4'8, In115, Sc45 and others, were selected to cover a large range of neutron spectrum. After the irradiation, the activation detectors were submitted to a spectrometry gamma by using a system of counting with high purity Germanium, to obtain the saturation activity per target nuclide. The saturation activity is one of the main data of input of unfolding code SANDBP, that through an iterative adjustment, modify the spectrum that better agree with the dataset of code input, composition mainly for measure reaction rate per target nuclide and a initial input spectrum, calculated for Hammer-Technion code, supplying a solution spectrum. (author)

  17. Criticality qualification of a new Monte Carlo code for reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Catsaros, N. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Gaveau, B. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Jaekel, M. [Laboratoire de Physique Theorique, Ecole Normale Superieure, 24 rue Lhomond, 75231 Paris (France); Maillard, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); CNRS-IDRIS, Bt 506, BP167, 91403 Orsay (France); CNRS-IN2P3, 3 rue Michel Ange, 75794 Paris (France); Maurel, G. [Faculte de Medecine, Universite Paris VI, 27 rue de Chaligny, 75012 Paris (France); MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Savva, P., E-mail: savvapan@ipta.demokritos.g [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Silva, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Varvayanni, M.; Zisis, Th. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece)

    2009-11-15

    In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.

  18. Extremely Nonlinear Optics Using Shaped Pulses Spectrally Broadened in an Argon- or Sulfur Hexafluoride-Filled Hollow-Core Fiber

    Directory of Open Access Journals (Sweden)

    Andreas Hoffmann

    2015-11-01

    Full Text Available In this contribution we present a comparison of the performance of spectrally broadened ultrashort pulses using a hollow-core fiber either filled with argon or sulfur hexafluoride (SF6 for demanding pulse-shaping experiments. The benefits of both gases for pulse-shaping are studied in the highly nonlinear process of high-harmonic generation. In this setup, temporally shaping the driving laser pulse leads to spectrally shaping of the output extreme ultraviolet (XUV spectrum, where total yield and spectral selectivity in the XUV are the targets of the optimization approach. The effect of using sulfur hexafluoride for pulse-shaping the XUV yield can be doubled compared to pulse compression and pulse-shaping using argon and the spectral range for selective optimization of a single harmonic can be extended. The obtained results are of interest for extending the range of ultrafast science applications drawing on tailored XUV fields.

  19. Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

  20. Chemical and physical analysis of core materials for advanced high temperature reactors with process heat applications

    International Nuclear Information System (INIS)

    Various chemical and physical methods for the analysis of structural materials have been developed in the research programmes for advanced high temperature reactors. These methods are discussed using as examples the structural materials of the reactor core - the fuel elements consisting of coated particles in a graphite matrix and the structural graphite. Emphasis is given to the methods of chemical analysis. The composition of fuel kernels is investigated using chemical analysis methods to determine the heavy metals content (uranium, plutonium, thorium and metallic impurity elements) and the amount of non-metallic constituents. The properties of the pyrocarbon and silicon carbide coatings of fuel elements are investigated using specially developed physiochemical methods. Regarding the irradiation behaviour of coated particles and fuel elements, methods have been developed for examining specimens in hot cells following exposures under reactor operating conditions, to supplement the measurements of in-reactor performance. For the structural graphite, the determination of impurities is important because certain impurities may cause pitting corrosion during irradiation. The localized analysis of very low impurity concentrations is carried out using spectrochemical d.c. arc excitation, local laser and inductively coupled plasma methods. (orig.)

  1. Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D.; Sterbentz, J. [Idaho National Engineering and Environmental Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415-3850 (United States); Meyer, M. [Argonne National Laboratory- West (United States); Lowden, R. [Oak Ridge National Laboratory (United States); Hoffman, E.; Wei, T.Y.C. [Argonne National Laboratory (United States)]. e-mail: weavkd@inel.gov

    2004-07-01

    The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO{sub 2}) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

  2. A simulation of a pebble bed reactor core by the MCNP-4C computer code

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

    Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

  3. Porosity Effect in the Core Thermal Hydraulics for Ultra High Temperature Gas-cooled Reactor

    Directory of Open Access Journals (Sweden)

    Motoo Fumizawa

    2008-12-01

    Full Text Available This study presents an experimental method of porosity evaluation and a predictive thermal-hydraulic analysis with packed spheres in a nuclear reactor core. The porosity experiments were carried out in both a fully shaken state with the closest possible packing and in a state of non-vibration. The predictive analysis considering the fixed porosity value was applied as a design condition for an Ultra High Temperature Reactor Experiment (UHTREX. The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 oC was achieved in the case of an UHTREX at Los Alamos Scientific Laboratory, which was a small scale UHTR. In the present study, the fuel was changed to a pebble type, a porous media. In order to compare the present pebble bed reactor and UHTREX, a calculation based on HTGR-GT300 was carried out in similar conditions with UHTREX; in other words, with an inlet coolant temperature of 871oC, system pressure of 3.45 MPa and power density of 1.3 w/cm3. As a result, the fuel temperature in the present pebble bed reactor showed an extremely lower value compared to that of UHTREX.

  4. Multimegawatt nuclear electric propulsion with gaseous and vapor core reactors with MHD

    Science.gov (United States)

    Knight, Travis; Anghaie, Samim; Smith, Blair; Houts, Michael

    2001-02-01

    This study investigated the development of a system concept for space power generation and nuclear electric propulsion based on a fissioning plasma core reactor (FPCR) with magnetohydrodynamic (MHD) power conversion system, coupled to a magnetoplasmadynamic (MPD) thruster. The FPCR is a liquid-vapor core reactor concept operating with metallic uranium or uranium tetrafluoride (UF4) vapor as the fissioning fuel and alkali metals or their fluorides as working fluid in a closed Rankine cycle with MHD energy conversion. Candidate working fluids include K, Li, Na, KF, LiF, NaF, etc. The system features core outlet temperatures of 3000 to 4000 K at pressures of about 1 to 10 MPa, MHD temperatures of 2000 to 3000 K, and radiator temperatures of 1200 to 2000 K. This combination of parameters offers the potential for low total system specific mass in the range of 0.4 to 0.6 kg/kWe. The MHD output could be coupled with minimal power conditioning to the variable specific impulse magnetoplasma rocket (VASIMR), MPD thrusters or other types of thruster for producing thrust at very high specific impulse (Isp=1500 to 10,000 s). .

  5. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    Energy Technology Data Exchange (ETDEWEB)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

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

    International Nuclear Information System (INIS)

    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

  7. Feasibility of core management system by data communication for boiling water reactors

    International Nuclear Information System (INIS)

    A core management system by data communication has been designed and proposed for more efficient operation of boiling water reactor (BWR) plants by faster transmission and centralized management of information. The system comprises three kinds f computers: a process computer for monitoring purposes at the reactor site, a center computer for administration purposes at the head office, and a large scientific computer for planning and evaluation purposes. The process and the large computers are connected to the center computer by a data transmission line. To demonstrate the feasibility of such a system, the operating history evaluation system, which is one of the subsystems of the core management system, has been developed along the above concept. Application to the evaluation of the operating history of a commercial BWR shows a great deal of merit. Quick response and a significant manpower reduction can be expected by data communication and minimized intervention of human labor. Visual display is also found to be very useful in understanding the core characteristics

  8. 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.

  9. Evaluation of matching between a pulsed-power and corona discharge reactor containing different thickness of soil

    International Nuclear Information System (INIS)

    Soil contamination by organic compounds has become an issue of concern around the world. Currently, non-thermal plasma, especially pulsed corona discharge, has received a great attention in environmental protection field. As a result, the matching between a pulsed-power and corona discharge reactor containing different thickness of soil was a significant aspect in optimizing the pulse corona discharge. In this paper, some methods have been adopted to achieve the matching, including choosing a suitable capacity, adjusting the frequency, providing a suitable soil thickness and comparing the energy and energy utilization efficiency. The details of the matching and optimization discussed were based upon the theories of streamer formation and experimental results as well. The results indicated that energy injected into the reactor increased with the pulse forming capacity and pulse frequency. There existed an optimal energy utilization efficiency with the change of soil thickness and pulse frequency under the pulse forming capacity of 100 pF. The SED at pulse voltage of 19 kV and pulse frequency of 70 Hz was achieved 0.11 J g−1soil at the soil thickness of 3 mm, which was only 0.064 J g−1soil at the soil thickness of 9 mm; meanwhile, with the increase of pulse frequency from 50 Hz to 90 Hz, the SED increased from 0.075 J g−1soil to 0.146 J g−1soil at 19 kV and soil thickness of 3 mm. This study is expected to provide reference for the matching between a pulsed-power and reactor containing different thickness of soil for producing corona discharge.

  10. Nuclear Power Station Kalkar, 300 MWe Nuclear Prototype Power Station with Fast Sodium Cooled Reactor (SNR-300), Safety Report Reactor Core Mark-Ia

    International Nuclear Information System (INIS)

    The nuclear power station Kalkar is a prototype with a sodium cooled fast reactor (SNR-300) and a thermal power of 762 MW. The initial licensing procedure in 1972 was based on the so-called Mark-I core. During the following years, this core underwent some changes, for instance the thickness of the radial blanket was reduced to lower the electricity generation costs, the design of the absorber systems had been further optimized, and it became clear, that a full core with plutonium from MAGNOX-reactors could not be realized and that fuel from light-water reactors had also to be used. In this licensing document the modified reactor core Mark-Ia is described together with its assemblies and their loading procedure. The content of radioactive materials and the irradiation protection measures are discussed and those accidents are describe in an enveloping manner, from which an influence of the core modification cannot be excluded. Finally, both core versions (Mark-I and Mark-Ia) are compared with each other

  11. Evaluation of surface deposits on the channel wall of trepanned reactor core graphite samples

    Science.gov (United States)

    Heard, P. J.; Payne, L.; Wootton, M. R.; Flewitt, P. E. J.

    2014-02-01

    Samples have been trepanned from the fuel and interstitial channel walls of PGA graphite reactor cores of two Magnox gas cooled power stations after a period of service. These samples have been considered explicitly for the presence of deposits on the channel facing surfaces. A combination of focused ion beam milling and imaging has been used to determine the presence of such deposits and where present to make measurements of the thickness. These thicknesses vary from a few nanometres to tens of micrometres. In addition, both the chemical composition and chemical state have been investigated using energy dispersive X-ray microanalysis in a scanning electron microscope and Raman spectroscopy respectively. EDX measurements showed that surface deposits found on the channel walls of one of the reactors contained increased concentrations of oxygen, iron, chromium and sulphur compared with the underlying material. Raman spectroscopy also suggested that the deposit had a smaller crystallite size than PGA graphite.

  12. Analysis of cracked core spray piping from the Quad Cities Unit 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R.; Gaitonde, S.M.

    1982-09-01

    The results of a metallurgical analysis of leaking cracks detected in the core spray injection piping of Commonwealth Edison Company's Quad Cities Unit 2 Boiling Water Reactor are described. The cracks were present in a welded 105/sup 0/ elbow assembly in the line, and were found to be caused by intergranular stress corrosion cracking associated with the probable presence of dissolved oxygen in the reactor cooling water and the presence of grain boundary sensitization and local residual stresses induced by welding. The failure is unusual in several respects, including the very large number of cracks (approximately 40) present in the failed component, the axial orientation of the cracks, and the fact that at least one crack completely penetrated a circumferential weld. Virtually all of the cracking occurred in forged material, and the microstructural evidence presented suggests that the orientation of the cracks was influenced by the presence of axially banded delta ferrite in the microstructure of the forged components.

  13. Research reactor in-core fuel management optimisation using the multiobjective cross-entropy method

    International Nuclear Information System (INIS)

    The in-core fuel management optimisation (ICFMO) problem has been studied for several decades. Very little research has, however, been aimed at multiobjective optimisation involving the fundamental notion of Pareto optimality. In this paper, the recently developed multiobjective optimisation using the cross-entropy method (MOO CEM) algorithm is applied to a multiobjective ICFMO problem for the first time. A derivation of the MOO CEM algorithm is presented for ICFMO, along with a constraint handling technique. The algorithm is applied to a biobjective test problem for the SAFARI-1 nuclear research reactor. The Pareto set approximated by the algorithm is compared to solutions obtained by typical operational reload strategies. The results indicate that the MOO CEM algorithm for multiobjective ICFMO is a robust and efficient method which is able to obtain a good spread of trade-off solutions. The method may therefore greatly aid in the decision making of a reactor operator tasked with designing reload configurations. (author)

  14. Design studies of interaction processes between melt reactor core material composition, coolant and construction material

    International Nuclear Information System (INIS)

    This effort presents results of the design studies performed for correct conducting of out-of-pile experiments at the National Nuclear Center of the Republic of Kazakhstan facilities to study interaction processes between the melt fuel composition and core materials, which might take place during accidents at the nuclear power plants. The design methods are considered to determine electromagnetic parameters of the facility melting unit - an induction melting furnace, to determine temperature fields in furnace construction component materials during the experiment process and at the moment of melt discharge. The methods are presented to calculate temperature fields and thermal flows in experimental device modelling a reactor vessel bottom when it contacts with the melt fuel composition. The results of thermal electric and hydraulic calculations are presented validating the experimental device operability designed to study the interaction processes between the melt fuel composition and reactor vessel bottom in simulating a residual energy release in the melt by direct current transmitting

  15. Cost-based optimization of a nuclear reactor core design: a preliminary model

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner F.; Alves Filho, Hermes [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Inst. Politecnico. Dept. de Modelagem Computacional]. E-mails: wfsacco@iprj.uerj.br; halves@iprj.uerj.br; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil). Div. de Reatores]. E-mail: cmnap@ien.gov.br

    2007-07-01

    A new formulation of a nuclear core design optimization problem is introduced in this article. Originally, the optimization problem consisted in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the radial power peaking factor in a three-enrichment zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. Here, we address the same problem using the minimization of the fuel and cladding materials costs as the objective function, and the radial power peaking factor as an operational constraint. This cost-based optimization problem is attacked by two metaheuristics, the standard genetic algorithm (SGA), and a recently introduced Metropolis algorithm called the Particle Collision Algorithm (PCA). The two algorithms are submitted to the same computational effort and their results are compared. As the formulation presented is preliminary, more elaborate models are also discussed (author)

  16. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    Science.gov (United States)

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor.

  17. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    Science.gov (United States)

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor. PMID:27552124

  18. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Science.gov (United States)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  19. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki [Nuclear Research Group, FMIPA, Bandung Institute of Technology Jl. Ganesha 10, Bandung 40132 (Indonesia); Miura, Ryosuke; Takaki, Naoyuki [Department of Nuclear Safety Engineering, Tokyo City University 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan); Sekimoto, H. [Emerritus Prof. of Research Laboratory for Nuclear Reactors, Tokyo Inst. of Technology (Japan)

    2015-09-30

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  20. Generation of few-cycle laser pulses:Comparison between atomic and molecular gases in a hollow-core fiber

    Institute of Scientific and Technical Information of China (English)

    黄志远; 戴晔; 赵睿睿; 王丁; 冷雨欣

    2016-01-01

    We numerically study the pulse compression approaches based on atomic or molecular gases in a hollow-core fiber. From the perspective of self-phase modulation (SPM), we give the extensive study of the SPM infl uence on a probe pulse with molecular phase modulation (MPM) effect. By comparing the two compression methods, we summarize their advan-tages and drawbacks to obtain the few-cycle pulses with micro-or millijoule energies. It is also shown that the double pump-probe approach can be used as a tunable dual-color source by adjusting the time delay between pump and probe pulses to proper values.