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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. Reactivity initiated accident (RIA) type tests and annular core pulse reactor (ACPR) operational experience

    International Nuclear Information System (INIS)

    This paper describes the test conducted to investigate the failure threshold of the fuel when subject to RIA, accomplished in the TRIGA ACPR Nuclear Research Institute, Pitesti. The reactor facility, the capsule used in experiments and the experimental results are presented. The failure threshold was determined at 200 cal/g for an atmospheric gap pressure comparable with similar tests. The failure threshold decreases with increasing gap pressure. The tests proved useful for a better understanding of the fuel behavior in the transient conditions. As it is known RIA is not a common accident for the CANDU reactors, but the fuel failure mechanism can be similar to other type of accidents as LOCA and PCM. The program will be continued, with better instrumentation for the fuel sample and also independent instrumentation to measure pulse characteristics with better statistics. A new project for the experimental fuel elements must be considered to eliminate fuel-endcap interactions. (author)

  3. Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source

    International Nuclear Information System (INIS)

    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) δ (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) δ (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)

  4. Characterization of Novel Calorimeters in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Hehr Brian D.

    2016-01-01

    Full Text Available 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.

  5. 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.

  6. Critical heat flux prediction for the annular core research reactor

    International Nuclear Information System (INIS)

    This paper reports on best estimate predictions of Critical Heat Flux Ratio (CHFR) obtained to support the upgrade of the Annular Core Research Reactor (ACRR) at Sandia National Laboratories for 2 to 4 MWt. The CHF productions are based on the University of New Mexico's (UNM)-CHF correlations in conjunction with the Global Conditions Hypothesis (GCH). Results indicate that for the range of inlet water temperature of 293 to 333 K, CHFR predictions range from 3.9 to 2.1, which is more than sufficient to support the proposed ACRR upgrade

  7. 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.

  8. 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.

  9. 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

  10. 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

  11. 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)

  12. A Preliminary Calculation of Annular Core Design for a High-flux Advanced Research Reactor

    International Nuclear Information System (INIS)

    Many of research reactors in operation over the world become old and the number of research reactors is expected to be reduced around 1/3 within a next decade. So it may be necessary to prepare in advance for the future demands of research reactors with a high performance. Therefore, based on the HANARO experiences through design to operation, a concept development of an improved research reactor is under doing. In this paper, 10 MW conceptual annular core is proposed and its basic characteristics were analyzed as a preliminary step

  13. 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...

  14. 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)

  15. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  16. Design and fabrication of the instrumented fuel elements for the Annular Core Research Reactor (ACRR)

    International Nuclear Information System (INIS)

    This report describes the design and fabrication techniques for the instrumented fuel elements of the Annular Core Research Reactor (ACRR). The thermocouple assemblies were designed and fabricated at Sandia Laboratories while the instrumented elements were assembled at Los Alamos Scientific Laboratory. In order to satisfy the ACRR's Technical Specifications, the thermocouples are required to measure temperature in excess of 18000C under rapid heating conditions. Because of the potentially high failure rates for thermocouples in such environments, the instrumented fuel elements are designed so that the thermocouples can be replaced easily

  17. 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.

  18. 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.

  19. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    International Nuclear Information System (INIS)

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  20. 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)

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. 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

  7. 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)

  8. 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

  9. 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)

  10. Safe operation of 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 reactor cooled by the pool water. The characteristics of the two reactors are presented. The reactor core configuration is shown as well as the original start-up core configuration. Fuel management of TRIGA steady state core allows obtaining the requested fluxes for experimental purposes in safe operation condition. One can firmly state that the present operation of the reactor and the HEU-LEU (High Enriched Uranium - Low Enriched Uranium), core conversion fully respect the provisions of the National Regulatory Body and IAEA. (authors)

  11. 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.

  12. 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.

  13. Vibration model of a pressurized water reactor which takes into account the fluid influence in the annular gap between core barrel and pressure vessel

    International Nuclear Information System (INIS)

    A theoretical vibration model of a pressurized water reactor is established and studied which takes into account the fluid-structure interaction of the coupled three-dimensional system reactor pressure vessel-core barrel (reactor cavity). Vibration differential equations are derived only for the two-dimensional movement; the eigenfrequencies and amplitude ratios of the undamped system as well as a dimensionless damping factor of cavity vibrations are calculated with the data of the WWER-440, and discussed. (orig.)

  14. 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...

  15. Startup testing of Romania dual-core test reactor

    International Nuclear Information System (INIS)

    Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

  16. 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

  17. Dual annular rotating open-quotes windowedclose quotes nuclear reflector reactor control system

    International Nuclear Information System (INIS)

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core. 4 figures

  18. 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 ...

  19. Dual-core TRIGA research and materials testing reactor

    International Nuclear Information System (INIS)

    General Atomic Company is under contract from the Romanian Institute for Nuclear Technologies to design, fabricate, and install a research reactor in support of the Romanian National Program for Power Reactor Development. The goal was to select a design concept that provided reasonably high neutron fluxes for long term testing of various fuel-cladding-coolant combinations and also provide high performance pulsing for transient testing of fuel specimens. An effective solution was achieved by the selection of a 14 MW steady-state TRIGA reactor for high flux endurance testing, and an Annular Core Pulsing Reactor (ACPR) for high performance pulsing testing, with both reactors mounted in the same reactor tank and operated independently. The fuel bundles for the steady-state reactor consist of 25 uranium-zirconium hydride rods clad in stainless steel arranged in a square 5 x 5 array. The steady-state core is provided with downflow cooling at a rate of approximately 275 gpm/bundle. Bundle flow tests will be performed with both heated and unheated models. The core will be optimized for peak thermal neutron flux and reactivity lifetime within the constraint of a peak fuel meat temperature of 7500C. The operation of the steady-state reactor at a power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position of 2.9 x 1014 n/cm2-sec. The corresponding fast neutron flux (less than 1.125 keV) will be 2.6 x 1014 nv. (U.S.)

  20. 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

  1. 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)

  2. Reactor core monitoring method

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Michitsugu [Tokyo Electric Power Co., Inc. (Japan); Kanemoto, Shigeru; Enomoto, Mitsuhiro; Ebata, Shigeo

    1998-05-06

    The present invention provides a method of monitoring the state of coolant flow in a reactor of a BWR power plant. Namely, a plurality of local power region monitors (LPRM) are disposed to the inside of the reactor core for monitoring a power distribution. Signals of at least two optional LPRM detectors situated at positions different in axial or radial positions of the reactor core are obtained. General fluctuation components which nuclear hydrothermally fluctuate in overall reactor core are removed from the components of the signals. Then, correlational functions between these signals are determined. The state of coolant flow in the reactor is monitored based on the correlational function. When the axial flowing rate and radial flow interference are monitored, the accuracy upon monitoring axial and radial local behaviors of coolants can be improved by thus previously removing the general fluctuation components from signals of LPRM detectors and extracting local void information near to LPRM detectors at high accuracy. (I.S.)

  3. Fast reactor core concepts to improve transmutation efficiency

    International Nuclear Information System (INIS)

    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

  4. 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

  5. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    In a BWR type nuclear reactor, the number of first fuel assemblies (uranium) loaded in a reactor core is smaller than that of second fuel assemblies (mixed oxide), the average burnup degree upon take-out of the first fuel assemblies is reduced to less than that of the second fuel assemblies, and the number of the kinds of the fuel rods constituting the first fuel assemblies is made smaller than that of the fuel rods constituting the second fuel assemblies. As a result, the variety of the plutonium enrichment degree is reduced to make the distribution of the axial enrichment degree uniform, thereby enabling to simplify the distribution of the enrichment degree. Then the number of molding fabrication steps for MOX fuel assemblies can be reduced, thereby enabling to reduce the cost for molding and fabrication. (N.H.)

  6. 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)

  7. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Dobrea, D.; Parvan, M.; Stefan, V. [Institute for Nuclear Research, Pitesti (Romania)

    2009-04-15

    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{sup -1} UO{sub 2}. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Ionescu, D.; Olteanu, G. [Inst. for Nuclear Research, Pitesti (Romania)

    2008-07-01

    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{sup -1} UO{sub 2}. 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)

  10. In-Reactor Densification of Dual Cooled Annular Fuel Pellet during Irradiation Test at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, Young Woo; Kim, Dong Joo; Kwon, Hyoung Mun; Kim, Keon Sik; Kim, Jong Hun; Oh, Jang Soo; Yang, Jae Ho; Koo, Yang Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    These advantages result in a considerably low pellet centerline temperature. Because of this considerably low pellet temperature, in-reactor behavior of an annular pellet, such as densification and swelling may be significantly different from that of the conventional PWR solid pellet. Since the pellet temperature of an annular fuel rod is lower than that of a PWR solid fuel rod by several hundred degrees, the in-reactor densification and swelling of a dual cooled annular fuel pellet might be considered as athermal phenomena due to a low pellet temperature. In order to investigate the in-reactor behavior of the annular UO{sub 2} pellet, HANARO irradiation test was planned and conducted for annular pellets with 5 different types. Post irradiation test is being carried out in the KAERI's PIE facility. In this study, we are going to report the preliminary results of PIE test on the inreactor densification behavior of a dual cooled annular fuel pellet. Irradiation test of dual cooled annular UO{sub 2} pellet was conducted at the OR-4 hole in HANARO by using a non-instrumented test rig. The preliminary results of PIE test on the in-reactor densification behavior showed that the irradiated pellets densified much more than expected values based on MATPRO relations of inreactor densification at low temperature in the annular pellet with low initial sintered density. It might be attributed to the higher fission rate during HANARO irradiation.

  11. 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)....

  12. Evaluation of NSRR reactor characteristics using a core transient behavior simulation code EXCURS-NSRR

    International Nuclear Information System (INIS)

    The Nuclear Safety Research Reactor(NSRR) in Japan Atomic Energy Research Institute (JAERI) is a modified TRIGA-ACPR(Annular Core Pulse Reactor) which was constructed in 1975 in order to investigate the fuel behavior mainly under reactivity initiated accident (RIA) conditions. This reactor generates very sharp pulse power with the maximum of 23GW by rapid reactivity insertion of the maximum of 4.7$, and has capability to simulate a power burst in RIAs of power reactors. Fuel failure mechanisms and the fuel failure threshold in RIAs have been investigated through irradiation of test fuel rods in the NSRR. The control system and the operation data acquisition system of the NSRR were modified in 1989. By the modification, the controlled high power operation with various power shape became possible. Also on-line data acquisition of reactor data such as reactor power, regulating rod position, and so on, became possible by the modification. Evaluation of reactor characteristics became easy and accurate by detailed comparison of the time history of operation data and calculation results. Authors have tried to evaluate some parameters or constants of reactor characteristics by using a core transient behavior simulation code EXCURS-NSRR which consists of one point reactor kinetic equations and thermal equations of driver fuel and coolant. Especially in a relatively small and slow pulse power generation with reactivity insertion of less than 1$, evaluation of feedback reactivity coefficient and relation between reactivity insertion and regulating rod position was conducted. This article presents the evaluation results by the comparison between obtained reactor data and parametric calculation results. (author)

  13. 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)

  14. Hydraulic lift-off issues for application of high performance annular fuels in pressurized water reactors

    International Nuclear Information System (INIS)

    Highlights: • Pin and assembly lift-off forces are compared between solid and annular fuel. • Annular fuel experiences much stronger uplift forces. • Much stronger hold-down forces are required by annular fuel assembly. • Engineering modifications for hold-down mechanisms are required by annular fuel. - Abstract: In the PWR core, the fuel assembly is firmly seated on the lower core plate during operation. However, if the hydraulic force exerted on the fuel assembly by coolant flow is too large and the fuel assembly is lifted-off from the lower core plate, the excessive vibration will cause fuel failure. Therefore, the hydraulic lift-off issue needs to be addressed when the advanced fuel assembly is developed. It has been shown that the advanced annular fuel design with internal cooling allows power uprating up to 50% while the peak temperature of the fuel can be reduced and the MDNBR can be maintained. However, if the coolant condition in the core is kept unchanged, increasing the core power by 50% requires the core flow rate also increase proportionally, which will give rise to the hydraulic lift-off, an important issue to be addressed. In this paper, taking the 17 × 17 solid fuel design as the reference, the hydraulic lift-off issue is investigated for proposed 12 × 12 and 13 × 13 annular fuel designs. Both the steady-state and start-up operating conditions are evaluated. It is found that the hydraulic lift-off indeed is an issue for annular fuel design which requires careful analysis. By comparison, the lift-off forces and hold-down forces required for the externally and internally cooled annular fuels (13 × 13 and 12 × 12 arrays) are several times larger than that of the referenced solid fuel (17 × 17 array). Therefore, the hold-down mechanism for annular fuel needs to be carefully designed

  15. 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

  16. Xi'an pulsed reactor

    International Nuclear Information System (INIS)

    Xi'an Pulsed Reactor (XAPR) Designed and constructed all by China is first research pulsed reactor with versatile applications. It is characterized with inherent safety, versatile application, structure simplicity and convenience for operation. It can be operated not only at stead-state but also at pulse mode as well as square wave mode. The rated power to the reactor under steady-state operation is 2 MW and the reactor is operated under pulsing state, its maximum peak power is about 4200 MW. XAPR is also equipped with many kinds of the experimental and irradiation facilities. The applications are radio-isotopes production, neutron activity analysis, neutron radiograph, monocrystalline silicon irradiation, material irradiation test, nuclear physics, neutron physics and nuclear chemistry studies, teaching and training. The XAPR has went into test operation and application for nearly two years that has shown its advantage and extensiveness

  17. 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.

  18. Description of Thermos reactor core

    International Nuclear Information System (INIS)

    A description is given of the 100 MWth Thermos reactor core (fuel, assembly, plates, casings, control rods) and its neutron thermohydraulic operation (steady state, transient state, fast stopping of a pump, primary circuit depressurization)

  19. The pulsed reactor and its application

    International Nuclear Information System (INIS)

    The situation of the first pulsed reactor in China is briefly described. The pulsed reactor has a large prompt negative temperature coefficient of reactivity provided by combination of the uranium-zirconium hydride fuel and the moderator. Therefore, its most outstanding features are 'inherent safety' and fairly high pulsed-power capacity. The pulsed reactor is now extensively used in science and technology

  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 for...... minimizing the optical power needed in reflectance pulse oximetry. To simplify packaging, backside photodiodes are made which are compatible with assembly using surface mounting technology without pre-packaging. Quantum efficiencies up to 95% and area-specific noise equivalent powers down to 30 fW Hz(-1....../2) cm(-1) are achieved. The photodiodes are incorporated into a wireless pulse oximetry sensor system embedded in an adhesive patch presented elsewhere as 'The Electronic Patch'. The annular photodiodes are fabricated using two masked diffusions of first boron and subsequently phosphor. The surface is...

  2. Reactor core monitoring device

    International Nuclear Information System (INIS)

    The device of the present invention reliably and conveniently detects an event of rapid increase of a coolant void coefficient at a portion of a channel by flow channel clogging event in a PWR-type reactor. Namely, upon flow channel clogging event, the coolant void coefficient is increased, an effective density is lowered, and a coolant shielding effect is lowered. Therefore, fast neutron fluxes at the periphery of a pressure tube are increased. The increase of the fast neutron fluxes is detected by a fast neutron flux detector disposed in a guide tube of an existent neutron flux detector. Based on the result, increase of coolant void coefficient can be detected. When an average void coefficient reaches from 30% to 100%, for example, the fast neutron fluxes are increased by about twice at a neutron permeation distance of coolants of about 10cm, thereby enabling to perform effective detection. (I.S.)

  3. 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

  4. IBR-2 - pulsed reactor for neutron investigations

    International Nuclear Information System (INIS)

    A brief theory and design are presented of IBR-2 fast neutron pulse reactor (with a periodic operation) of an average power of 3 MW constructed in Dubna and intended for investigation of structure and dynamics of liquids and solids in nuclear and neutron physics. The core of the reactor has the volume of 22 l and is filled with a liquid sodium. Fuel elements are made of sintered plutonium dioxide tablets and placed into a stainless steel cylinder 8.6 mm in diameter with wall thickness of 0.45 mm. The height of the fuel element active part is 445 mm, and the total length of the fuel element is about 780 mm. Fuel elements are separated from each other by a wire 0.5 mm in diameter wound around each element like a spiral. Fuel elements are assembled in cassets (their total number is 78), 7 elements in each casset. To achieve periodic operation and produce power pulses to be applied to the reactor two movable reflectors are used. The booster-type operation is assumed with using line induction accelerator. The peak density of thermal neutron flux in the reactor is 1016 cm-2s-1, for power pulse repetition rate 5 Hz and duration approximately μs. The scope of investigations to be performed on IBR-2 using the flight-of-time spectrometry method is also considered

  5. 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.

  6. Reactor core and fueling method

    International Nuclear Information System (INIS)

    When MOX fuel assemblies are used in a portion of fuel assembly of a BWR type nuclear reactor, neutron spectra are hardened due to the difference of the nuclear property between uranium and plutonium. As a result, the reactivity controllability of burnable poisons such as gadolinia is lowered, and the multiplication factor of the MOX fuel assembly at the initial stage of burning is increased greater than that of an uranium fuel assembly, to reduce thermal margin and reactor shutdown margin. Then, in the present invention, fresh fuel assemblies containing plutonium are disposed in a first region at the second layer from the outermost circumference of the reactor core and in a second region in adjacent with a control cell. Since the MOX fuel assemblies with increasing reactivity are disposed in the first and the second regions of small neutron importance, the power at the periphery of the reactor core and the circumference of the control cell can be kept substantially constant throughout the operation period. Further, satisfactory reactor operation can be kept without causing excess distortion of power distribution. (N.H.)

  7. Granular Dynamics in Pebble Bed Reactor Cores

    Science.gov (United States)

    Laufer, Michael Robert

    This study focused on developing a better understanding of granular dynamics in pebble bed reactor cores through experimental work and computer simulations. The work completed includes analysis of pebble motion data from three scaled experiments based on the annular core of the Pebble Bed Fluoride Salt-Cooled High- Temperature Reactor (PB-FHR). The experiments are accompanied by the development of a new discrete element simulation code, GRECO, which is designed to offer a simple user interface and simplified two-dimensional system that can be used for iterative purposes in the preliminary phases of core design. The results of this study are focused on the PB-FHR, but can easily be extended for gas-cooled reactor designs. Experimental results are presented for three Pebble Recirculation Experiments (PREX). PREX 2 and 3.0 are conventional gravity-dominated granular systems based on the annular PB-FHR core design for a 900 MWth commercial prototype plant and a 16 MWth test reactor, respectively. Detailed results are presented for the pebble velocity field, mixing at the radial zone interfaces, and pebble residence times. A new Monte Carlo algorithm was developed to study the residence time distributions of pebbles in different radial zones. These dry experiments demonstrated the basic viability of radial pebble zoning in cores with diverging geometry before pebbles reach the active core. Results are also presented from PREX 3.1, a scaled facility that uses simulant materials to evaluate the impact of coupled fluid drag forces on the granular dynamics in the PB-FHR core. PREX 3.1 was used to collect first of a kind pebble motion data in a multidimensional porous media flow field. Pebble motion data were collected for a range of axial and cross fluid flow configurations where the drag forces range from half the buoyancy force up to ten times greater than the buoyancy force. Detailed analysis is presented for the pebble velocity field, mixing behavior, and residence time

  8. Pulsed TRIGA reactor as substitute for long pulse spallation neutron source

    International Nuclear Information System (INIS)

    TRIGA reactor cores have been used to demonstrate various pulsing applications. The TRIGA reactor fuel (U-ZrHx) is very robust especially in pulsing applications. The features required to produce 50 pulses per second have been successfully demonstrated individually, including pulse tests with small diameter fuel rods. A partially optimized core has been evaluated for pulses at 50 Hz with peak pulsed power up to 100 MW and an average power up to 10 MW. Depending on the design, the full width at half power of the individual pulses can range between 2000 μsec to 3000 μsec. Until recently, the relatively long pulses (2000 μsec to 3000 μsec) from a pulsed thermal reactor or a long pulse spallation source (LPSS) have been considered unsuitable for time-of-flight measurements of neutron scattering. More recently considerable attention has been devoted to evaluating the performance of long pulse (1000 to 4000 μs) spallation sources for the same type of neutron measurements originally performed only with short pulses from spallation sources (SPSS). Adequate information is available to permit meaningful comparisons between CW, SPSS, and LPSS neutron sources. Except where extremely high resolution is required (fraction of a percent), which does require short pulses, it is demonstrated that the LPSS source with a 1000 msec or longer pulse length and a repetition rate of 50 to 60 Hz gives results comparable to those from the 60 MW ILL (CW) source. For many of these applications the shorter pulse is not necessarily a disadvantage, but it is not an advantage over the long pulse system. In one study, the conclusion is that a 5 MW 2000 μsec LPSS source improves the capability for structural biology studies of macromolecules by at least a factor of 5 over that achievable with a high flux reactor. Recent studies have identified the advantages and usefulness of long pulse neutron sources. It is evident that the multiple pulse TRIGA reactor can produce pulses comparable to those

  9. Fast neutron benchmark proposal at TRIGA-ACPR Reactor

    International Nuclear Information System (INIS)

    The development of fast neutron benchmarks is a historical aim of reactor physics. The dry experimental tube situated in the central region of the core in TRIGA Annular-Core Pulsing Reactor (ACPR) offers a suitable neutron source for fast neutron benchmark development. Our proposal consists in mounting a high-enriched uranium annular converter into the dry channel of the core. Preliminary computations and measurements are presented in this paper. Neutron flux computations in the dry channel and the uranium converter were performed using MCNP and WIMS codes. Also neutron flux spectrum measurements and fast and thermal neutron flux distribution measurements were performed using foil activation techniques. (authors)

  10. 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.

  11. 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.

  12. 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

  13. 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

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

    International Nuclear Information System (INIS)

    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

  15. 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.

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

    International Nuclear Information System (INIS)

    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 for minimizing the optical power needed in reflectance pulse oximetry. To simplify packaging, backside photodiodes are made which are compatible with assembly using surface mounting technology without pre-packaging. Quantum efficiencies up to 95% and area-specific noise equivalent powers down to 30 fW Hz-1/2 cm-1 are achieved. The photodiodes are incorporated into a wireless pulse oximetry sensor system embedded in an adhesive patch presented elsewhere as 'The Electronic Patch'. The annular photodiodes are fabricated using two masked diffusions of first boron and subsequently phosphor. The surface is passivated with a layer of silicon nitride also serving as an optical filter. As the final process, after metallization, a hole in the center of the photodiode is etched using deep reactive ion etch.

  17. Research on plasma core reactors

    Science.gov (United States)

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

    1976-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diameter by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17 cm thick by 89 cm diameter beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000 cu cm aluminum canister in the central region was fueled with UF6 gas and fission density distributions determined. These results are to be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  18. The need to address the larger universe of HEU-fueled reactors, including: Critical assemblies, pulsed reactors and propulsion reactors

    International Nuclear Information System (INIS)

    The RERTR program has focused on ending shipments of HEU fuel to research reactors. Highest priority has been given to reactors with steady thermal powers ≥ 1 megawatt. Since the cores of critical assemblies and pulsed reactors can contain huge amounts of HEU, they should be a second focus. Also, since many aging and specialized HEU-fuelled reactors may no longer be needed, more emphasis should be given to initiatives that could assist in their shutdown and decommissioning, including providing access to regional reactors with superior facilities. HEU-fuelled ship-propulsion reactors should also be addressed. Russia's civilian icebreaker reactors are of particular interest because their fuel design is considered less sensitive than that of naval reactor fuel. Moreover, Russia's KLT-40 icebreaker reactor is being adapted for a floating nuclear power plant and LEU icebreaker fuel could be used for converting Russian research reactors such as PIK and SM-3, that operate at power-reactor temperatures. (author)

  19. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Gas sealed assemblies are disposed in rows between reactor core fuel assemblies. The gas sealed assembly incorporates inflowed sodium (coolants) and sealed gas in a gas sealing cylinder and an inner hollow of a wrapper tube. A cylindrical heat generating member is disposed in the gas sealing cylinder. The sealed gas is compressed by a discharging pressure of a pump by way of sodium in the wrapper tube. During normal operation, the liquid level of the coolants is present above than a backwarding flow hole, and the temperature of the coolants is raised by the cylindrical heat generation member to raise the temperature of sodium in the backwarding flow hole. High temperature sodium is mixed with low temperature sodium from a lower flow hole at the lower portion of the backwarding flow hole, and sodium at a leak flow hole becomes sodium at a middle temperature. The temperature of the middle temperature sodium is detected by a thermometer. With such procedures, the liquid level in the gas sealed assembly can be detected and confirmed during normal operation. (I.N.)

  20. 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.

  1. 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.

  2. Seismic research on graphite reactor core

    International Nuclear Information System (INIS)

    Background: Reactors with graphite core structure include production reactor, water-cooled graphite reactor, gas-cooled reactor, high-temperature gas-cooled reactor and so on. Multi-body graphite core structure has nonlinear response under seismic excitation, which is different from the response of general civil structure, metal connection structure or bolted structure. Purpose: In order to provide references for the designing and construction of HTR-PM. This paper reviews the history of reactor seismic research evaluation from certain countries, and summarizes the research methods and research results. Methods: By comparing the methods adopted in different gas-cooled reactor cores, inspiration for our own HTR seismic research was achieved. Results and Conclusions: In this paper, the research ideas of graphite core seismic during the process of designing, constructing and operating HTR-10 are expounded. Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced. (authors)

  3. Embrittlement of reactor core materials

    International Nuclear Information System (INIS)

    Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion-generated hydrogen and neutron irradiation damage. Results are presented on the effects of fast fluence on the fracture toughness of: (1) wrought Zircaloy-2, Zircaloy-4, and hafnium; (2) Zircaloy-4 to hafnium butt welds; and (3) hydrogen-precharged beta-treated and weld-metal Zircaloy-4 for fluences up to a maximum of approximately 150 x 1024 n/m2 (>1 MeV). While Zircaloy-4 did not exhibit a decrement in KIC due to irradiation, hafnium and butt welds between hafnium and Zircaloy-4 are susceptible to embrittlement with irradiation. The embrittlement can be attributed to irradiation strengthening, which promotes void formation in the high-strain crack-tip region, and, in part, to the lower chemical potential of hydrogen in Zircaloy-4 compared to hafnium, which causes hydrogen to drift over time from the hafnium end toward the Zircaloy-4 end and to precipitate at the interface between the weld and base-metal interface. Neutron radiation apparently affects the fracture toughness of Zircaloy-2, Zircaloy-4, and hafnium in different ways. Possible explanations for these differences are suggested. It was found that Zircaloy-4 is preferred over Zircaloy-2 in hafnium-to-Zircaloy butt-weld applications due to absence of a radiation-induced reduction in KIC plus its lower hydrogen absorption characteristics compared with Zircaloy-2

  4. Stochastically fluctuations of the modernized fast pulsed reactor IBR-2

    International Nuclear Information System (INIS)

    Full Text : Stochastically fluctuations of the power of the IBR-2 reactor have been quite significant (20 percent), they affect the dynamics of the reactor, the process of regulation, starting on the work of the experimental equipment, etc. On the other hand, the presence of large fluctuations in power at the IBR-2M has had its advantages. Investigation of stochastic fluctuations has allowed to estimate some physical parameters of the nuclear reactor core, for example, the mean lifetime of prompt neutrons in the reactor, source of spontaneous neutrons, and absolute power of the reactor. The main results of the investigation impulse stochastically fluctuations of the IBR-2 periodic pulsed reactor after modernization have been presented. It has been shown that the experimental results have been close to the calculated ones

  5. Fixed bed suspended core nuclear reactor concept

    International Nuclear Information System (INIS)

    The fixed bed nuclear reactor (FBNR) is essentially a pressurized light water reactor having spherical fuel elements constituting a suspended reactor core at its lowest bed porosity. The principle features of the proposed reactor are that the concept is polyvalent, simple in design, may operate either as fixed or fluidized bed, have the core suspended contributing to inherent safety, passive cooling features of the reactor. The reactor is modular and has an integrated primary system utilizing either water, supercritical steam or helium gas as its coolant. Some of the advantages of the proposed reactor are being modular, low environmental impact, exclusion of severe accidents, short construction period, flexible adaptation to demand, excellent load following characteristics, and competitive economics. (orig.)

  6. Design study of marine reactor core

    International Nuclear Information System (INIS)

    JAERI have carried out four core designs for three different type Reactor of Trial design in FY 1983 ∼ 1986 and one core (the optimum core) and three survey cores for conceptual design Reactor in 1987. Based on these cores study results, we are now studying to design MRX CORE. On the other side, we started design study of super-miniaturized 2 MWt core concept for deep-sea submersible. This report describes the results of comparison and studies of the core specification, specific characteristics etc of these cores, and we study that more thick fuel outer diameter (9.5 mm) core was possible or not as for the MRX CORE that has now thin fuel outer diameter (7.0 mm) in consideration of rapid power change etc especially with marine reactor. As the results, it was found that 9.5 mm diameter fuel core was possible and some methods were found, therefore it will be necessary to study the 9.5 mm diameter fuel core in detail continuously. (author)

  7. Dynamic Response Control of Three-Layered Annular Plate Due to Various Parametres of Electrorheological Core

    Directory of Open Access Journals (Sweden)

    Pawlus Dorota

    2016-03-01

    Full Text Available The paper presents dynamic responses of annular plate composed of three layers. The middle layer of the plate has electrorheological properties expressed by the Bingham body model. The plate is loaded in the plane of facings with time-dependent forces. The electrorheological effect is observed in the area of supercritical plate behaviour. The influence of both material properties and geometrical dimensions of the core on plate behaviour is examined. The problem is solved analytically and numerically using the orthogonalization method and the finite difference method. Comparison of the results obtained using the finite difference and the finite element methods for a plate in critical state is shown. The numerical calculations are carried out for axisymmetric and asymmetric plate modes. The presented diagrams show the plate reaction to the changes in values of plate parameters and indicate that the supercritical control of plate work is possible.

  8. Reactor core heterogeneity effects on radionuclide inventory

    International Nuclear Information System (INIS)

    Highlights: ► Reactor core heterogeneity effects on radionuclide inventory are studied. ► A methodology for inventory estimation of individual fuel assembly is developed. ► Estimated inventory using presently developed and conventional method is compared. ► Radionuclide density peak and its location in equilibrium VVER core are investigated. - Abstract: Understanding fission product behavior is an important aspect of nuclear safety studies. A nuclear reactor core contains complex mixture of fuel elements with different levels of enrichment, power and burnup. Conventional method of core radionuclide inventory estimation is based on use of homogenized core parameters like burnup, enrichment, specific power etc. This approach does not reveal the variation in radionuclide density among different fuel elements within the core. The present work aims to bridge this knowledge gap by estimating the distribution of radionuclides in the nuclear reactor core taking into account the heterogeneity of the core explicitly. The analysis has brought out peak radionuclide density in the core which can be helpful in appropriate estimation of the radiological release in the accidental scenarios involving failures of few fuel assemblies. A quantitative comparison of total core inventory estimated based conventional core average parameters and detailed core inventory using individual fuel assembly inventory estimate has been made

  9. 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

  10. Fast Reactor Physics Parameters from a Pulsed Source

    International Nuclear Information System (INIS)

    One of the more important integral parameters in fast reactor physics analysis is the neutron spectrum of a particular composition reactor core. Various methods, such as proton recoil counters and nuclear emulsion analysis, have been used to study fast reactor spectra. With the development of high intensity short-duration pulsed neutron sources, the time-of-flight technique has become suitable for fast reactor spectrum determination. To evaluate the feasibility of measuring fast neutron spectra from a core using time-of-flight techniques, an experiment has been performed to measure the equilibrium spectmm in a large block of depleted uranium using the General Atomics Linac facilities. A ten-metric-ton block of depleted uranium was assembled to form a 81-cm cube. This block of uranium was pulsed by electron bombardment of a uranium target imbedded in the block. The spectra from various sections of the block were measured using time-of-flight techniques for a 50-m flight path. Spectral indices, such as the ratio of the fission rates of U238/U235, U233/U235, U234/U235, Np237/U235, Pu239/U235 were also measured. In addition, measurements of the U238 capture rates were obtained in various parts of the block. This paper describes the techniques used to obtain these reactor physics parameters. The experimental results such as the spectra and spectral indices are also compared with those obtained from theoretical considerations using multigroup transport theory analysis. The pulsed neutron technique is also applicable for the measurement of such parameters as: β/ℓ, where β is the effective delayed neutron fraction and ℓ is the lifetime; neutron importance; and keff. This paper concludes with a discussion on the proposed application of a pulsed neutron source for the measurement of some of these parameters on fast reactor cores constructed on ZPR-VI, the Argonne Fast Critical Facility. (author)

  11. Pulsed reactor IBR-2 in the 'nineties

    International Nuclear Information System (INIS)

    One may expect that in the 1990s a considerable part of neutron scattering experiments will be carried out on pulsed neutron sources. Pulsed reactors of the IBR-type are the founders of a family of pulsed sources of thermal neutrons. The first IBR-type reactor started operation in 1960 with a mean thermal power of 1 kW. In 1984 the IBR-2 reactor had a mean power of 2 MW, a pulsed power of 1500 MW at 5 s-1 repetition rate and a power pulse duration of 215 μs. The peak leakage current of thermal neutrons from the moderator surface reached 1x1016n.cm-2.s-1 which allows on the sample a mean flux of thermal neutrons up to 5x107n.cm-2.s-1 monochromized by the time-of-flight method with an uncertainty of Δdelta approx.= 0.02 A. The achieved parameters, being reasonably good for a wide range of experiments, are not to be considered as being a technical limit for this type of pulsed neutron source. The experience we already have allows us to expect development in the following directions: increasing of the reactor mean power, shortening of the pulse duration, installation of the liquid hydrogen moderator. The IBR-2 reactor is used in four main fields: physics of condensed matter, nuclear physics, fundamental and applied physics. The greatest activity is in condensed matter research. In addition to conventional structural and dynamic studies we intend to expand investigations to kinetics of transitions in the time range up to 10-4s (time-resolved spectroscopy), properties of short-living non-equilibrium states, properties of matter under extreme conditions created in the pulsed mode (e.g. under the influence of superstrong magnetic fields). We expect that within the next decade, neutron methods will be increasingly introduced into different areas of research and technology. (author)

  12. 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

  13. 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)

  14. Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

    International Nuclear Information System (INIS)

    Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO2SO4) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

  15. Innovative reactor core: potentialities and design

    International Nuclear Information System (INIS)

    Gen IV nuclear reactors are considered a very attractive answer for the demand of energy. Because public acceptance they have to fulfil very clearly the requirement of sustainable development. In this sense a reactor concept, having by itself a rather no significant interaction with the environment both on the front and back end ('adiabatic concept'), is vital. This goal in mind, a new way of designing such a core has to be assumed. The starting point must be the 'zero impact'. Therefore the core will be designed having as basic constraints: a) fed with only natural or depleted Uranium, and b) discharges only fission products. Meantime its potentiality as a net burner of Minor Actinide has to be carefully estimated. This activity, referred to the ELSY reactor, shows how to design such an 'adiabatic' core and states its reasonable capability of burning MA legacy in the order of 25-50 kg/GWey. (authors)

  16. 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

  17. 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)

  18. 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.)

  19. Removing the heat from fast reactor cores

    International Nuclear Information System (INIS)

    Whatever the view about the time when fast breeder reactors will reach the commercial and industrial stage, there is a growing and widespread interest in developing their technology. The reactors are called breeders because they can produce more fissile material than they use in their own cores. As part of an Agency programme related to their technology and economics a symposium on Alkali Metal Coolants - Corrosion Studies and System Operating Experience was held in Vienna from 28 November to 2 December

  20. 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.

  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. Reactor core design of Gas Turbine High Temperature Reactor 300

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been designing Japan's original gas turbine high temperature reactor, Gas Turbine High Temperature Reactor 300 (GTHTR300). The greatly simplified design based on salient features of the High Temperature Gas-cooled Reactor (HTGR) with a closed helium gas turbine enables the GTHTR300 a highly efficient and economically competitive reactor to be deployed in early 2010s. Also, the GTHTR300 fully taking advantage of various experiences accumulated in design, construction and operation of the High Temperature Engineering Test Reactor (HTTR) and existing fossil fired gas turbine systems reduces technological development concerning a reactor system and electric generation system. Original design features of this system are the reactor core design based on a newly proposed refueling scheme named sandwich shuffling, conventional steel material usage for a reactor pressure vessel (RPV), an innovative coolant flow scheme and a horizontally installed gas turbine unit. The GTHTR300 can be continuously operated without the refueling for 2 years. Due to these salient features, the capital cost of the GTHTR300 is less than a target cost of 200,000 yen (1667 US$)/kW e, and the electric generation cost is close to a target cost of 4 yen (3.3 US cents)/kW h. This paper describes the original design features focusing on the reactor core design and the in-core structure design, including the innovative coolant flow scheme for cooling the RPV. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan

  3. IRIS reactor core with thorium fuel

    International Nuclear Information System (INIS)

    This work is aimed at running the first IRIS reactor core with mixed thorium dioxide fuel (ThO2-UO2 and ThO2-PuO2). Calculations are performed by using Dragon 4.0.4 and Citation codes. The results show the multiplication factor(Keff) for central and peripheral assemblies as a function of burnup. To ensure the proliferation resistance,the value of 235U enrichment is ≤ 20%. The Keff is calculated using Dragon 4.0.4 for a single fuel rod and the model developed to fuel assembly, while the whole core was calculated using Citation code. For a fuel burnup, the use of increased enrichment fuel in the IRIS core leads to high reserve of reactivity, which is compensated with an integral fuel burnable absorber. The self-shielding of boron is in an IRIS reactor fuel. The effect of increased enrichment to the burn-up rates, and burnable poison distribution on the reactor performance, are evaluated. The equipment used in traditional light water reactors is evaluated for designing a small unit IRIS reactor. (authors)

  4. Reactor core flow rate measuring device

    International Nuclear Information System (INIS)

    Purpose: To accurately measure the reactor core flow rate of coolants compulsorily circulated to the reactor core. Constitution: The discharge flow rate from internal pumps has been measured by disposing a flow nozzle, an orifice, etc. to the suction or discharge port of the internal pump and determining the pressure difference thereof or by the pumping stroke. Although such a method enables easy measurement, it involves problems in view of accuracy and maintenance. According to the present invention, a post-like member of a definite length is disposed to the opening of the reactor core shroud support leg just before an internal pump and the vortex frequency emitted from the member is measured to thereby determine the flow velocity and thus the flow rate. the vortex frequency is in proportion with the flow velocity, not depending on the composition, density, temperature, pressure of fluid. The vortex frequency is measured by a piezoelectric sensor or a strain gage. Accordingly, it is possible to accurately measure the discharge flow rate of individual internal pumps to thereby easily control the reactor core power. (K.M.)

  5. Nuclear reactor with a reactor core composed of fuel elements

    International Nuclear Information System (INIS)

    A tube surrounding a fuel element projects above the liquid level. The tube is situated in a pot, whose upper edge lies between the top of the reactor core and the liquid level. A greater pressure is therefore produced, which ensures a reduction of the steam bubble proportion in the cooling liquid at the other fuel elements. (orig./HP)

  6. Nuclear reactor core modelling in multifunctional simulators

    International Nuclear Information System (INIS)

    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

  7. Reactor core flowing monitoring device

    International Nuclear Information System (INIS)

    In a BWR type reactor, a regression model is applied to signals of two local neutron detectors on the same string. Estimated values of signals of a detector disposed at an upper position (at downstream) obtained by inputting the signals of the detector on the lower position (at upstream) to the model do not contain a fluctuation ingredient caused by bubbles generated between the detectors. Then, if the difference of the observed values of the detector at the upper position and estimated values of the detector at the upper position obtained by the regression model is determined, time series data of the fluctuation ingredient of the bubbles generated between the detectors can be obtained. The regression model is a mathematical model estimating regression of the output by a linear linkage of the values of input signals in the past. Since the fluctuation ingredient caused by the bubbles contained in each of the detectors can be separated irrespective of frequency, the transferring time and the speed of the bubbles between the detectors can be calculated accurately. (N.H.)

  8. Neutronic Core Performance of CAREM-25 Reactor

    International Nuclear Information System (INIS)

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

  9. Fuel elements for pulsed TRIGA research reactors

    International Nuclear Information System (INIS)

    TRIGA fuel was developed around the concept of inherent safety. A core composition was sought that had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature would automatically cause the power excursion to terminate before any core damage resulted. Experiments have demonstrated that zirconium hydride possesses a basic neutron-spectrum-hardening mechanism to produce the desired characteristic. Additional advantages include the facts that ZrH has a good heat capacity, that it results in relatively small core sizes and high flux values due to the high hydrogen content, that it has excellent fission-product retentivity and high chemical inertness in water at temperatures up to 1000C, and that it can be used effectively in a rugged fuel element size. Tens of thousands of routine pulses to the range of 500 to 8000C peak fuel temperatures have been performed with TRIGA fuel, and a core was pulse-heated to peak fuel temperatures in excess of 11000C for hundreds of pulses before a few elements exceeded the conservative tolerances on dimensional change

  10. Stochastic dynamics of a periodic pulsed reactor

    International Nuclear Information System (INIS)

    A periodical pulsed reactor (PPR) dynamics far beyond stability is analyzed using the discrete nonlinear model as a basic one. It is shown that a PPR with negative temperature reactivity effect inevitably comes up to a stochastic chaos. A way by which a reactor goes to chaos is defined by a time dependence of feedback and by a kind of dynamics model. The most usual case is the Feigenbaum transition the matter of which is that the chaotic motion will appear after the PPR passes through the infinite cascade of oscillation period doubling. The transition of PPR to chaotic behaviour must be considered to be safe. 16 refs.; 7 figs

  11. 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

  12. Liquid metal reactor core material HT9

    International Nuclear Information System (INIS)

    A state-of-the art is surveyed on the liquid metal reactor core materials HT9. The purpose of this report is to give an insight for choosing and developing the materials to be applied to the KAERI prototype liquid metal reactor which is planned for the year of 2010. In-core stability of cladding materials is important to the extension of fuel burnup. Austenitic stainless steel (AISI 316) has been used as core material in the early LMR due to the good mechanical properties at high temperatures, but it has been found to show a poor swelling resistance. So many efforts have been made to solve this problem that HT9 have been developed. HT9 is 12Cr-1MoVW steel. The microstructure of HT9 consisted of tempered martensite with dispersed carbide. HT9 has superior irradiation swelling resistance as other BCC metals, and good sodium compatibility. HT9 has also a good irradiation creep properties below 500 dg C, but irradiation creep properties are degraded above 500 dg C. Researches are currently in progress to modify the HT9 in order to improve the irradiation creep properties above 500 dg C. New design studies for decreasing the core temperature below 500 dg C are needed to use HT9 as a core material. On the contrary, decrease of the thermal efficiency may occur due to lower-down of the operation temperature. (author). 51 refs., 6 tabs., 19 figs

  13. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH1.8) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  14. Tajoura reactor core conversion neutrons analysis

    International Nuclear Information System (INIS)

    This paper presents the preliminary neutronics studies and results of the Tajoura reactor core conversion calculations from currently used highly enriched (80% U235) fuel to low enriched fuel (36% U''2''3''5) by using the TAJN computer package. The compact core loading consists of 16 fuel assemblies type IRT-2M surrounded by removable and stationary beryllium reflector and ordinary water for moderation and cooling. The study was undertaken to compare results of TAJN computer package and the vendor documented results. The results of these calculations at the BOL and EOL conditions with equilibrium Xe at 10 MWt are shown. (author)

  15. 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

    A one-dimensional (1D) Fourier-Bessel series method for computing and tuning (beamforming) the linear lossless field of flat pulsed wave annular arrays is developed and supported with both numerical simulation and experimental verification. The technique represents a new method for modeling 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...

  16. PC-Reactor-core transient simulation code

    International Nuclear Information System (INIS)

    PC-REATOR, a reactor core transient simulation code has been developed for the real-time operator training on a IBM-PC microcomputer. The program presents capabilities for on-line exchange of the operating parameters during the transient simulation, by friendly keyboard instructions. The model is based on the point-kinetics approximation, with 2 delayed neutron percursors and up to 11 decay power generating groups. (author)

  17. Dynamic detection of nuclear reactor core incident

    International Nuclear Information System (INIS)

    Surveillance, safety and security of evolving systems area challenge to prevent accident. The dynamic detection of a hypothetical and theoretical blockage incident in the Phenix nuclear reactor is investigated. Such an incident is characterized by abnormal temperature rises in the neighbourhood of the concerned reactor core assembly. The data set is the output temperature map of the reactor, it is provided by the Atomic Energy and Alternative Energies Commission (CEA). A real time approach is proposed, based on a sliding temporal window, it is divided into two steps. The first one behaves like a sieve, its function is to detect simultaneous temperature evolutions in a close neighbourhood which may induce a potential incident. When such evolutions are detected, the second step computes the temperature contrast between each assembly having these evolutions and its neighbourhood. This method permits to monitor the system evolution in real time while only few observations are required. Results are validated on various noisy realistic simulated perturbations. (authors)

  18. 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

  19. 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.

  20. Monte Carlo simulations of periodic pulsed reactor with moving geometry parts

    International Nuclear Information System (INIS)

    Highlights: • In some periodic pulsed reactor, the reflector of the reactor core moves such that its state varies periodically from subcritical state to slightly prompt critical for producing periodic power pulses. • Current techniques of using the point kinetics approximations to simulate the dynamics in such a periodic pulse reactor are not very accurate when the reactor core is composed of zones with significantly different neutron lifetimes. • A Monte Carlo procedure is developed to simulate the geometrical movement of the reflector. • Two algorithms are proposed in modifying the neutron history tracks due to the movement of the reflector surfaces. • Numerical test cases have been developed to validate the developed Monte Carlo procedure. - Abstract: 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

  1. Emergency reactor core cooling system of BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides an emergency reactor core cooling system which can reduce a capacity of a power source required upon occurrence of emergency, extending an start-up time of an emergency reactor core cooling system (ECCA) to provide a plant endurable to a common factor accident and can provide time margin up to the start-up time. Namely, the system of the present invention comprises a division I equipped with an isolation condenser (IC), an after-heat removing system (low pressure system)(LPFL/RHR) and an emergency gas turbine generator (GT), a division II equipped with a diesel driving water injection system (high pressure system)(HDIS), LPFL/RHR, and GT, and a division III equipped with a reactor isolation time cooling system (high pressure system)(ARCIC), LPFL/RHR and GT. With such a constitution, since the IC, HDIS and ARCIC are used in combination as a high pressure system, an electromotive pump required to be operated upon high pressure state can be saved. In addition, if a static reactor cooling system (PCCS) is adopted and is provided with a back-up function for LPFL/RHR with respect to heat removal of the container upon occurrence of an accident, the countermeasure for occurrence of severe accidents can be enhanced. (I.S.)

  2. Condensed phase thermochemistry of reactor core debris

    International Nuclear Information System (INIS)

    This paper discusses a nonideal solution model of the metallic phases of reactor core debris. The metal phase model is based on the Kohler equation for a 37 component system. The binary subsystems are assumed to have subregular interactions. The model is parameterized by comparison to available data and by estimating subregular interactions using the methods developed by Miedama et al. The model is shown to predict phase separation in the metallic phase of core debris. The model also predicts reduced chemical activities of zirconium and tellurium in the metal phase. A model of the oxide phase of core debris is described briefly. The model treats the oxide phase as an associated solution. The chemical activities of solution components are determined by the existence and interactions of species formed from the components

  3. 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.)

  4. Core concepts for 'zero-sodium-void-worth core' in metal fuelled fast reactor

    International Nuclear Information System (INIS)

    Core design options to reduce the sodium void worth in metal fuelled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a 'pancaked' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket-zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. (author)

  5. Behavior of an heterogeneous annular FBR core during an unprotected loss of flow accident: Analysis of the primary phase with SAS-SFR

    International Nuclear Information System (INIS)

    In the framework of a substantial improvement on FBR core safety connected to the development of a new Gen IV reactor type, heterogeneous core with innovative features are being carefully analyzed in France since 2009. At EDF R and D, the main goal is to understand whether a strong reduction of the Na-void worth - possibly attempting a negative value - allows a significant improvement of the core behavior during an unprotected loss of flow accident. Also, the physical behavior of such a core is of interest, before and beyond the (possible) onset of Na boiling. Hence, a cutting-edge heterogeneous design, featuring an annular shape, a Na-plena with a B4C plate and a stepwise modulation of fissile core heights, was developed at EDF by means of the SDDS methodology, with a total Na-void worth of -1 $. The behavior of such a core during the primary phase of a severe accident, initiated by an unprotected loss of flow, is analyzed by means of the SAS-SFR code. This study is carried-out at KIT and EDF, in the framework of a scientific collaboration on innovative FBR severe accident analyses. The results show that the reduction of the Na-void worth is very effective, but is not sufficient alone to avoid Na-boiling and, hence, to prevent the core from entering into the primary phase of a severe accident. Nevertheless, the grace time up to boiling onset is greatly enhanced in comparison to a more traditional homogeneous core design, and only an extremely low fraction of the fuel (<0.1%) enters into melting at the end of this phase. A sensitivity analysis shows that, due to the inherent neutronic characteristics of such a core, the gagging scheme plays a major role on the core behavior: indeed, an improved 4-zones gagging scheme, associated with an enhanced control rod drive line expansion feed-back effect, finally prevents the core from entering into sodium boiling. This major conclusion highlights both the progress already accomplished and the need for more detailed future

  6. Sensors for use in nuclear reactor cores

    International Nuclear Information System (INIS)

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

  7. 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)

  8. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

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

  9. The neutron radiography facility designed for TRIGA reactors and its results

    International Nuclear Information System (INIS)

    The two TRIGA reactors of INR, the Steady State Reactor (SSR) having a power of 14 MW and Annular Core Pulsing Reactor (ACPR) having in steady state a power of 500 kW and being capable of a pulse to the peak power of 20000 MW, are placed in the same pool. The neutron flux ranging at the edges of those reactors cores is suitable for neutron radiography. The neutron radiography facility is placed in the pool of the TRIGA reactors. Till now as neutron source only the ACPR, in steady state or pulsing mode has been used. For the future one intends to use also the neutron flux of SSR. The aim of this facility is to achieve neutron radiographs of the nuclear fuel elements. (authors)

  10. Transient rod failure in a pulsing TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Full text: On July 7, 1970 the University of Texas at Austin TRIGA Mark I Pulsing Reactor experienced a failure of the transient control rod. Although no danger to personnel or damage to the reactor other than the pulse rod occurred, the failure was promptly reported to the USAEC regional compliance office. The first indication of an abnormal situation was unusual multiplication behavior during the first start-up of the day. As usual for steady state operation, the operator removed the transient rod and began to withdraw the shim and regulating rods. After partial withdrawal, he noticed that the count rate was not increasing as rapidly as was customary. While remaining at the console,the operator had a technician make a visual inspection of the core. The technician observed the transient drive rod was swinging freely in the pool and the poison section was detached. It was concluded, based on the indications of the.reactor instrumentation and visual inspection, that the transient control rod had broken off and remained in position in the core. The regulating and shim rods were inserted and the transient rod was manually cranked to the down position. The manual manipulation of the transient rod, instead of dropping the rod by gravity, was used so that the connecting rod could be reinserted in the control rod guide tube. The reactor core was then partially unloaded so that a critical mass was not present. The transient rod drive and connecting rod were removed from the pool. The poison section was retrieved from its position in the core by welding a tap to a long rod and tapping into the top of the poison section. Visual inspection of the poison section showed that the weld joining the male threads on the poison section to the main body of the control rod had failed. The threads remained screwed in the control rod drive shaft upon separation and the poison section remained fully inserted in the core. A new control rod was fabricated by Gulf General Atomic and shipped

  11. Approach to development of high flux research reactor with pebble-bed core

    International Nuclear Information System (INIS)

    Full text: The research nuclear reactor of a basin-type IRT with the designed power of 1 MW was put into operation in 'Sosny' settlement not far from Minsk-city in the Republic of Belarus in 1962. In 1971 after its modernization the power was increased up to 4 MW and maximum density of neutron flux in the core was: Thermal 5·1013 neutr./cm2.s Fast (E>0.8 MeV) 2·1013 neutr./cm2.s The reactor has been used for carrying out investigations in the field of solid-state physics, radiation construction materials, radiobiology, gaseous chemically reacting coolants and others. After the Chernobyl NPP accident, in the former USSR the requirements on safety of nuclear reactors have become sufficiently stricter. As to some parameters these requirements became the same as for reactors of nuclear power plants. In this connection the reactor in 'Sosny' settlement did not answer these new requirements by a number of performances such as seismicity of building, efficiency of control and protection system, corrosion in the reactor vessel and others, and it was shutdown in 1987 and its decommissioning was performed during 1988-1999. At the Joint Institute of Power and Nuclear Research -'SOSNY' have been carried out investigations on feasibility of creation of the research reactor with pebble-bed core. The concept of such reactor supposes using the following technical approaches: - Using as fuel the brought sphere micro fuel elements with the diameter of 500-750 mkm to an industrial level; - Organization of reactor operation in the regime with minimum possible fueling with 235U; - Implementation of hydraulic loading - unloading of micro fuel elements with the frequency of one or several days. Physical calculations of the core were carried out with the help of MCU-RFFI program based on the Monte-Carlo method. Two configurations of the pebble-bed core in the high flux reactor have been considered. The first configuration is the core with a neutron trap and an annular fuel layer formed

  12. Magnet design approach for pulsed tokamak reactors

    International Nuclear Information System (INIS)

    A choice of various operating modes of a tokamak reactor will have considerable impact on the fatigue lives and cost of ohmic heating (OH), equilibrium field (EF), and toroidal field (TF) coils. OH and EF coil requirements and their costs, as well as the effects of the fringing fields of the EF coils on the TF coils, have been studied under cyclic operation in the range of N = 102 to 106 cycles, spanning the range from a noninductively driven reactor (STARFIRE) to a conventional ohmically driven reactor. For a reference design of TF coils the design of the central OH solenoid has been studied as a function of its maximum field, B /SUP OH/. Increasing requirements for structural support lead to only negligible increases in voltseconds for B /SUP OH/greater than or equal to 10.0 T. Fatigue failure of the OH coil is not a concern for N less than or equal to 105; for N about 106 fatigue limits the strain to small values, resulting in small increases in structural requirements and modest decreases in volt-seconds. Should noninductive current drive be achievable the authors note that this not only eliminates the OH coil, but it also permits EF coil placement in the inboard region, which facilitates the creation of highly shaped plasma cross sections (large triangularity, or bean-shaped equilibria). They have computed the stored energy, coil configuration and fringing fields for a number of EF coil design options. For pulsed operation these EF coil fields result in three major problems. First, the induced eddy currents introduce additional heating to the TF magnets, which increases cryogenic demands compared to steady-state operation. Second, of greater concern, these EF coils impart out-of-plane forces and turning moments to the TF coil structure. In this case fatigue failure is a severe design constraint. Finally, pulsed operation of the EF coils requires large, expensive power supplies

  13. Near-field anomalous spectral behavior in diffraction of a Gaussian pulsed beam from an annular aperture.

    Science.gov (United States)

    Yang, Yuanjie; Zou, Qihui; Li, Yude

    2007-07-20

    Based on the Fresnel diffraction integral and by introducing a hard-aperture function into a finite sum of complex Gaussian functions, the approximate analytical expression for the near-field spectral intensity distribution of a space-time-dependent Gaussian pulsed beam passing through an annular aperture is derived, which permits us to study the on- and off-axis spectral anomalies that are near phase singularities of the diffracted Gaussian pulsed beam in the near-field. The expressions for a circular black screen and a circular aperture are given as special cases of the general results. The relative spectral shift of a space-time-dependent Gaussian pulsed beam versus the different values of the truncation parameters and the position parameters of observation points are also studied and illustrated with numerical calculations. It is shown that the spectral switch appears near phase singularities in the near-field, and the near-field spectral behavior depends on the truncation parameters, the pulse duration tau, and the position parameter. The results of this work have potential applications in free-space information encoding and transmission. PMID:17609713

  14. 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.

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

    International Nuclear Information System (INIS)

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

  16. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R., E-mail: rustamzia@yahoo.co [Vienna University of Technology, Atominstitute (ATI), Stadion allee 2, A-1020, Vienna (Austria); Stummer, T.; Boeck, H.; Villa, M. [Vienna University of Technology, Atominstitute (ATI), Stadion allee 2, A-1020, Vienna (Austria)

    2011-05-15

    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 ({kappa}{sub eff}) 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.

  18. 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.

  19. Nonlinear stability of oscillatory core-annular flow: A generalized Kuramoto-Sivashinsky equation with time periodic coefficients

    Science.gov (United States)

    Coward, Adrian V.; Papageorgiou, Demetrios T.; Smyrlis, Yiorgos S.

    1994-01-01

    In this paper the nonlinear stability of two-phase core-annular flow in a pipe is examined when the acting pressure gradient is modulated by time harmonic oscillations and viscosity stratification and interfacial tension is present. An exact solution of the Navier-Stokes equations is used as the background state to develop an asymptotic theory valid for thin annular layers, which leads to a novel nonlinear evolution describing the spatio-temporal evolution of the interface. The evolution equation is an extension of the equation found for constant pressure gradients and generalizes the Kuramoto-Sivashinsky equation with dispersive effects found by Papageorgiou, Maldarelli & Rumschitzki, Phys. Fluids A 2(3), 1990, pp. 340-352, to a similar system with time periodic coefficients. The distinct regimes of slow and moderate flow are considered and the corresponding evolution is derived. Certain solutions are described analytically in the neighborhood of the first bifurcation point by use of multiple scales asymptotics. Extensive numerical experiments, using dynamical systems ideas, are carried out in order to evaluate the effect of the oscillatory pressure gradient on the solutions in the presence of a constant pressure gradient.

  20. 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

  1. Interfacial friction in low flowrate vertical annular flow

    International Nuclear Information System (INIS)

    During boil-off and reflood transients in nuclear reactors, the core liquid inventory and inlet flowrate are largely determined by the interfacial friction in the reactor core. For these transients, annular flow occurs at relatively modest liquid flowrates and at the low heat fluxes typical of decay heat conditions. The resulting low vapor Reynolds numbers, are out of the data range used to develop the generally accepted interfacial friction relations for annular flow. In addition, most existing annular flow data comes from air/liquid adiabatic experiments with fully developed flows. By contrast, in a reactor core, the flow is continuously developing along the heated length as the vapor flowrate increases and the flow regimes evolve from bubbly to annular flow. Indeed, the entire annular flow regime may exist only over tens of L/D's. Despite these limitations, many of the advanced reactor safety analysis codes employ the Wallis model for interfacial friction in annular flow. Our analyses of the conditions existing at the end-of-reflood in the PERICLES tests have indicated that the Wallis model seriously underestimates the interfacial shear for low vapor velocity cocurrent upflow. To extend the annular flow data base to diabatic low flowrate conditions, the DADINE tests were re-analyzed. In these tests, both pressure drop and local cross-section averaged void fractions were measured. Thus, both the wall and interfacial shear can be deduced. Based on the results of this analysis, a new correlation is proposed for interfacial friction in annular flow. (authors). 5 figs., 12 refs

  2. SIMULATE-3 core model for nuclear reactor training simulators

    International Nuclear Information System (INIS)

    This paper describes the adaptation of the Studsvik nuclear reactor analysis code, SIMULATE-3, to nuclear reactor training simulation. This adaption to real-time applications permits training simulation to be performed using the same 'engineering grade' core model used for core design, loading optimisation, safety analysis, and plant technical support. Use of SIMULATE-3R in training simulation permits simple initialisation of simulator core-models (without need for tuning) and facilitates application of cycle-specific core models. SIMULATE-3R permits training simulation of reactor cores with the accuracy normally associated with engineering analysis and enhances the simulator's 'plant analyser' functions. (author)

  3. Hollow-core fibers for high power pulse delivery

    DEFF Research Database (Denmark)

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

    2016-01-01

    We investigate hollow-core fibers for fiber delivery of high power ultrashort laser pulses. We use numerical techniques to design an anti-resonant hollow-core fiber having one layer of non-touching tubes to determine which structures offer the best optical properties for the delivery of high power......-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...... 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...

  4. 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.

  5. 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

  6. Design of radiation shields in nuclear reactor core

    International Nuclear Information System (INIS)

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

  7. 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.-

  8. Development of thermal hydraulic analysis code for nuclear reactors with annular fuels and assessment of the KAIST DNB-type theoretical critical heat flux model

    International Nuclear Information System (INIS)

    The development of thermal hydraulic analysis code for Gas-Cooled Reactors (GCRs) and for annular fuel and its application to various types of nuclear reactors, and the assessment of the Korea Advanced Institute of Science and Technology (KAIST) Departure from Nucleate Boiling (DNB)-type theoretical Critical Heat Flux (CHF) model for rod bundles with non-uniform axial power shapes were investigated. Thermal hydraulic characteristics of thorium-based fuel assemblies with annular seed pins were analyzed using Thermal-Hydraulic analysis code for Annular Fuel (THAF) combined with Multichannel Analyzer for steady states and Transients in Rod Arrays (MATRA), and compared with those of existing thorium-based assemblies. This study investigates the possibilities of using annular fuel pins in a pressurized water reactor with emphasis on coolant flow distribution and heat transfer fraction in internal and external sub-channels. MATRA and THAF showed good agreements for the pressure drops at the internal sub-channels. Mass fluxes were high in inner sub-channels of the seed pins due to the grid form losses in the outer sub-channels. About 43% of heat generated from the seed pin flowed into the inner sub-channel. The remaining heat flowed into the outer sub-channel. The inner to outer wall heat flux ratio was approximately 1.2. Maximum temperatures of annular seed pins were slightly above 500 .deg. C. Minimum DNB Ratios (MDNBRs) of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Temperatures and enthalpies were higher in the inner sub-channels due to the fact that inter-channel mixing cannot occur in the inner sub-channels. A thermal-hydraulic analysis code for annular fuel-based Liquid Metal Reactors (LMRs) has been developed. About 41% of the heat generated from the fuel pin flowed into the inner sub-channel and the rest into the outer sub-channel. The inner to outer wall heat flux ratio was equal to approximately 1.44. A new 37

  9. Thermo-hydraulic simulations of the experimental fast reactor core

    International Nuclear Information System (INIS)

    A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author)

  10. Nuclear Safety Research Reactor (NSRR) as a facility for reactor safety research and its modification for the future test plan

    International Nuclear Information System (INIS)

    The NSRR is a modified TRIGA-ACPR (annular core pulse reactor), and attained the initial criticality in May, 1975. It was built for studying reactor fuel behavior under a reactivity-initiated accident condition. The reactor is installed in a pool of 3.6 m width, 4.5 m length and 9 m depth, and water above the reactor core serves as a radiation shield. The reactor core contains 149 driver fuel rods, 6 regulating rods, 2 safety rods and 3 transient rods. An arbitrary reactivity up to 4.67 $ can be set up almost instantaneously in the reactor core. The pulse power generation is terminated by the large negative reactivity induced by prompt temperature feedback without inserting the control rods. This is brought about by an excellent property of the driver fuel which contains 12 wt.% U-ZrH enriched to 20 wt.% U-235. As a unique feature, the NSRR is equipped with a big experimental cavity through the center of the reactor core. It has the diameter of 220 mm, and is called loading tube. It is branched into a vertical loading tube and an offset loading tube. The characteristics of the pulse operation in the NSRR, the outline of fuel irradiation experiment, the future test plan and the modification of the NSRR are described. (Kako, I.)

  11. Materials problems in magnetically confined pulsed fusion reactors

    International Nuclear Information System (INIS)

    It is noted that materials in fusion power reactors must function satisfactorily under conditions of intense high energy neutron flux heat loads, and temperatures ranging from cryogenic to about 13000K. The competition between fatigue and thermal creep will occur in all pulsed fusion reactors, but with the additional possibility of radiation creep being the dominant deformation mode. The response of structural support members to the effects of neutron irradiation, in pulses, of cyclic temperature transients, and therefore cyclic thermal stresses, and of elevated temperature must be evaluated for each different type of pulsed fusion reactor having different combinations of cyclic stress, steady stress, temperature, flux level, and metal choice. (U.S.)

  12. The seismic assessment of fast reactor cores in the UK

    International Nuclear Information System (INIS)

    The design of the UK Commercial Demonstration Fast Reactor (CDFR) has evolved over a number of years. The design has to meet two seismic requirements: (i) the reactor must cause no hazard to the public during or after the Safe Shutdown Earthquake (SSE); (ii) there must be no sudden reduction in safety for an earthquake exceeding the SSE. The core is a complicated component in the whole reactor. It is usually represented in a very simplified manner in the seismic assessment of the whole reactor station. From this calculation, a time history or response spectrum can be generated for the diagrid, which supports the core, and for the above core structure, which supports the main absorber rods. These data may then be used to perform a detailed assessment of the reactor core. A new simplified model of the core response may then be made and used in a further calculation of the whole reactor. The calculation of the core response only, is considered in the remainder of this paper. One important feature of the fast reactor core, compared with other reactors, is that the components are relatively thin and flexible to promote neutron economy and heat transfer. A further important feature is that there are very small gaps between the wrapper tubes. This leads to very strong fluid-coupling effects. These effects are likely to be beneficial, but adequate techniques to calculate them are only just being developed. 9 refs, figs

  13. Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

  14. Lifetime embrittlement of reactor core materials

    International Nuclear Information System (INIS)

    Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion- generated hydrogen and to neutron irradiation damage. Results are presented on the effects of fast fluence on the fracture toughness of wrought Zircaloy-2, Zircaloy-4, and hafnium; Zircaloy-4 to hafnium butt welds; and hydrogen precharged beta treated and weld metal Zircaloy-4 for fluences up to a maximum of approximately 150 x 1024 n/M2 (> 1 Mev). While Zircaloy-4 did not exhibit a decrement in KIC due to irradiation, hafnium and butt welds between hafnium and Zircaloy-4 are susceptible to embrittlement with irradiation. The embrittlement can be attributed to irradiation strengthening, which promotes cleavage fracture in hafnium and hafnium-Zircaloy welds, and, in part, to the lower chemical potential of hydrogen in Zircaloy-4 compared to hafnium, which causes hydrogen, over time, to drift from the hafnium end toward the Zircaloy-4 end and to precipitate at the interface between the weld and base-metal interface. Neutron radiation apparently affects the fracture toughness of Zircaloy-2, Zircaloy-4, and hafnium in different ways. Possible explanations for these differences are suggested. It was found that Zircaloy-4 is preferred over Zircaloy-2 in hafnium-to- Zircaloy butt-weld applications due to its absence of a radiation- induced reduction in KIC plus its lower hydrogen absorption characteristics compared with Zircaloy-2

  15. Operational safety experience at 14 MW TRIGA research reactor from INR Pitesti, Romania

    International Nuclear Information System (INIS)

    The safe operation of TRIGA-14 MW Core and Annular Pulsed TRIGA Core in the assembly of Research Reactor in Pitesti, Romania for 27 years is presented from historical perspective as well in the light of evolving safety experience. The accomplishment of safety objectives and responsibilities of operating organization is described and sustained with practical examples including management responsibilities, resources of management, performance indicators, measurement analysis and monitoring. Further improvement of safety of Research Reactor trough a large refurbishment and modernization program under way is also presented in the paper. (author)

  16. Physics and behaviour during a ULOF of an innovative heterogeneous annular FBR core

    International Nuclear Information System (INIS)

    The major conclusions: • The reduction of the Na void worth is a way allowing a strong improvement of the dynamic behavior in very severe ULOF transient (10 s halving time), possibly allowing to avoid Na boiling; • 1st order effects: Na density ( 0); • 2nd order effects: - Mass flow gaggling scheme (as a function of the core neutronics); - Other feed-back effects: diagrid, driveline feed-back. → Very strong impact of uncertainties: Thermalhydraulic models & codes, drive-line feed-back modeling; • Methodology for feed-back coefficient calculation (example: in this calculation the Na density effect is linearized from nominal to 100% void, anticonservative in case of no Na boiling); • Core neutronics: nuclear data, models. → Even in case of no Na boiling, the critical events will be: • Fuel cladding and S/A wrapper behavior at very high temperature; • Upper core structures behavior

  17. 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)

  18. 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)

  19. Design for reactor core safety in nuclear power plants

    International Nuclear Information System (INIS)

    This Guide covers the neutronic, thermal, hydraulic, mechanical, chemical and irradiation considerations important to the safe design of a nuclear reactor core. The Guide applies to the types of thermal neutron reactor power plants that are now in common use and fuelled with oxide fuels: advanced gas cooled reactor (AGR), boiling water reactor (BWR), pressurized heavy water reactor (PHWR) (pressure tube and pressure vessel type) and pressurized water reactor (PWR). It deals with the individual components and systems that make up the core and associated equipment and with design provisions for the safe operation of the core and safe handling of the fuel and other core components. The Guide discusses the reactor vessel internals and the reactivity control and shutdown devices mounted on the vessel. Possible effects on requirements for the reactor coolant, the reactor coolant system and its pressure boundary (including the pressure vessel) are considered only as far as necessary to clarify the interface with the Safety Guide on Reactor Coolant and Associated Systems in Nuclear Power Plants (IAEA Safety Series No. 50-SG-D13) and other Guides. In relation to instrumentation and control systems the guidance is mainly limited to functional requirements

  20. Impulse power stochastically fluctuations of the modernized fast pulsed reactor IBR-2

    International Nuclear Information System (INIS)

    Full text : Stochastically fluctuations of the power of the IBR-2 reactor have been quite significant, they affect the dynamics of the reactor, the process of regulation, starting on the work of the experimental equipment. On the other hand, the presence of large fluctuations in power at the IBR-2M has had its advantages. Investigation of stochastic fluctuations has allowed to estimate some physical parameters of the nuclear reactor core, for example, the mean lifetime of prompt neutrons in the reactor, source of spontaneous neutrons and absolute power of the reactor. The main results of the investigation impulse power stochastically fluctuations of the IBR-2 periodic pulsed reactor after modernization have been presented. It has been shown that the experimental results have been close to the calculated ones

  1. 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

  2. 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

  3. Fast current pulse amplifier for neutron flux monitoring system of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The neutron flux monitoring system (NFMS) for Prototype Fast Breeder Reactor (PFBR) measures the neutron power and the reactivity changes in the core in all the states such as shut down, fuel handling, reactor startup, intermediate and power ranges using high temperature cylindrical fission chambers, four section fission counter and high temperature boron coated counter. Fast Current Pulse Amplifier has been developed to use in NFMS of PFBR that amplifies single/four numbers of input current pulses independently, discriminates and electronically wire - OR them to give differential pulse output along with the Campbell output. The paper describes the design, development of integrated single/Quad channel fast current pulse amplifier based on in-house developed ASIC, Hybrid IC, in built test features, LV and HV supplies. (author)

  4. 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.

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

    International Nuclear Information System (INIS)

    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. 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.

  7. 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 ...

  8. Evaluations of two-phase natural circulation flow induced in the reactor vessel annular gap under ERVC conditions

    International Nuclear Information System (INIS)

    Highlights: ► Two-phase natural circulation flow induced in insulation gap was investigated. ► Half-scaled non-heating experiments were performed to evaluate flow behavior. ► The loop-integrated momentum equation was formulated and solved asymptotically. ► First-order approximate solution was obtained and agreed with experimental data. - Abstract: The process of two-phase natural circulation flow induced in the annular gap between the reactor vessel and the insulation under external reactor vessel cooling conditions was investigated experimentally and analytically in this study. HERMES-HALF experiments were performed to observe and quantify the induced two-phase natural circulation flow in the annular gap. A half-scaled non-heating experimental facility was designed by utilizing the results of a scaling analysis to simulate the APR1400 reactor and its insulation system. The behavior of the boiling-induced two-phase natural circulation flow in the annular gap was observed, and the liquid mass flow rates driven by the natural circulation loop and the void fraction distribution were measured. Direct flow visualization revealed that choking would occur under certain flow conditions in the minimum gap region near the shear keys. Specifically, large recirculation flows were observed in the minimum gap region for large air injection rates and small outlet areas. Under such conditions, the injected air could not pass through the minimum gap region, resulting in the occurrence of choking near the minimum gap with a periodical air back flow being generated. Therefore, a design modification of the minimum gap region needs to be done to facilitate steam venting and to prevent choking from occurring. To complement the HERMES-HALF experimental effort, an analytical study of the dependence of the induced natural circulation mass flow rate on the inlet area and the volumetric air injection rate was performed using a loop integration of the momentum equation. The loop

  9. The axisymmetric long-wave interfacial stability of core-annular flow of power-law fluid with surfactant

    Science.gov (United States)

    Sun, Xue-Wei; Peng, Jie; Zhu, Ke-Qin

    2012-02-01

    The long wave stability of core-annular flow of power-law fluids with an axial pressure gradient is investigated at low Reynolds number. The interface between the two fluids is populated with an insoluble surfactant. The analytic solution for the growth rate of perturbation is obtained with long wave approximation. We are mainly concerned with the effects of shear-thinning/thickening property and interfacial surfactant on the flow stability. The results show that the influence of shear-thinning/thickening property accounts to the change of the capillary number. For a clean interface, the shear-thinning property enhances the capillary instability when the interface is close to the pipe wall. The converse is true when the interface is close to the pipe centerline. For shear-thickening fluids, the situation is reversed. When the interface is close to the pipe centerline, the capillary instability can be restrained due to the influence of surfactant. A parameter set can be found under which the flow is linearly stable.

  10. Falling the fuel assembly in core mesh of reactor

    International Nuclear Information System (INIS)

    Accident reflecting drop of a fuel assembly (FA) in core mesh during the overload operations in the INP AS RUz research reactor is observed. Calculations and analysis of the accident situation were carried out for the reactor cores formed from fully high enriched IRT-3M type fuel (36% enrichment on '235U), the first mixed core consisting from 16 IRT-3M and 4 IRT-4M with low enriched fuel (19.7% enrichment on 235U), and the core fully formed from low enriched fuel. (authors)

  11. A neutronic study on advanced sodium cooled fast reactor cores with thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    Highlights: • SFR burner core configurations are explored and analyzed for effective use of thorium blankets. • Thorium blankets can significantly improve SFR burner core performances. • No recycling or partial recycling of Th blankets with multi-batches is very effective. - Abstract: In this paper, new design concepts of sodium cooled fast reactor (SFR) cores having thorium blanket are suggested for pursuing effective burning of TRU (transuranics) nuclides from LWR spent fuels and their neutronic performances are analyzed. Several core configurations having different arrangements of thorium blankets are explored to improve the core performances and safety-related parameters including sodium void worth which is one of main concerns on safety of SFR cores. Specifically, axial and radial thorium blankets are considered for two type cores. The first one is the typical annular type cores having two different fuel regions where axial thorium blankets are placed in the axially central regions while the second one is the single fuel region cores having central non-fuel region where the axial blanket and radial blankets are considered. Also, the effects of the recycling options and fuel management schemes of the used thorium blanket on the core performances are analyzed. The core performance analyses show that thorium blankets with no recycling option and multi-batch fuel management schemes are very effective to improve the core performances including burnup reactivity swing, sodium void worth and TRU consumption rate

  12. Core management and full core conversion status of the Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the old 250 kW TRIGA-MARK II reactor. The reconstructed reactor reached its initial criticality in November 1983 and attained it nominal power of 500 kW in February 1984. The DNRR uses Russian fuel assemblies (FAs) type VVR-M2. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 HEU FAs. The 11 new HEU FAs were added in the core periphery, at previous beryllium element locations. After reloading the working configuration of reactor core consisted of 100 HEU FAs. Second reloading for Dalat Nuclear Research Reactor was realized in March 2002. The 4 new HEU FAs were added in the core periphery, at previous beryllium element locations. After reloading the working configuration of reactor core consisted of 104 HEU FAs. The third fuel reloading by shuffling of HEU FAs was executed in June 2004. The shuffling of 16 HEU FAs with highest burn up in the centre of the core and 16 HEU FAs with low burn up in the core periphery was done. The working configuration of reactor core kept unchanged of 104 HEU FAs. The fourth fuel reloading was executed in November 2006. The 2 new HEU FAs were loaded in the core periphery, at previous locations of wet irradiation channel and dry irradiation channel. After reloading the working configuration of reactor core consisted of 106 HEU FAs. Contracts for reactor core conversion between USA, Russia, Vietnam and the International Atomic Energy Agency for Nuclear fuel manufacture and supply for DNRR and Return of Russian-origin non-irradiated highly enriched uranium fuel to the Russian Federation have been realized in 2007. The 35 fresh HEU FAs were sent back to Russian Federation. The 36 new LEU FAs from Russian Federation have been received. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory and Vietnam

  13. 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

  14. 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

  15. Plastic-dynamic analysis on shock absorber of reactor core

    International Nuclear Information System (INIS)

    The plastic-dynamic process under the condition of impact is studied for the shock absorbing device. The safety of the reactor core and vessel can be ensured by reasonably selecting the dimensions to lessen the dynamic loading factor

  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. Core Seismic Tests for a Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2007-01-15

    This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

  18. Core conversion effects on the safety analysis of research reactors

    International Nuclear Information System (INIS)

    The safety related parameters of the 5 MW Democritus research reactor that will be affected by the scheduled core conversion to use LEU instead of HEU are considered. The analysis of the safety related items involved in such a core conversion, mainly the consequences due to MCA, DBA, etc., is of a general nature and can, therefore, be applied to other similar pool type reactors as well. (T.A.)

  19. 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.)

  20. 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.

  1. Reactor vessel model flow tests for 145-fuel assembly core

    International Nuclear Information System (INIS)

    Hydraulic tests on a one-sixth-scale model of a two-loop pressurized water reactor with 145 fuel assemblies are described. Core inlet and outlet flow distributions and reactor vessel pressure drop were investigated. The core inlet flow distribution was developed to be independent of the flow conditions in the inlet annulus. A flow distribution system, consisting of several flow splitters in the inlet annulus and a spherical plate flow distributor in the lower head region, was developed to obtain a symmetric and stable core inlet flow distribution. A minimum core inlet flow factor of 0.99 was established in the core. Reactor vessel unrecoverable pressure drops were measured on the model to predict losses that will occur in the prototype

  2. CANDU 6 reactor core physics and site physicist role

    International Nuclear Information System (INIS)

    The CANDU reactor is fuelled on-line. There is thus an on-going need for fuel and core management which is supported by an on-site Reactor Physics group. The author outlines the role of the on-site Physics group at the Point Lepreau Generating Station. This role covers Production, Technical as well as Safety and Compliance aspects

  3. The core design of the advanced power reactor plus (APR+)

    International Nuclear Information System (INIS)

    Advance Power Reactor Plus (APR+), a pressurized water reactor and an improved nuclear power reactor based on the Advanced Power Reactor 1400 MWe (APR1400) in Korea, has been developed with 18-month cycle operation strategy from its initial core. The APR+ core power is 4290 MWth which corresponds to a 1500 MWe class nuclear power plant. The reactor core consists of 257 fuel assemblies. Comparing with APR1400 core design, 16 fuel assemblies are added. Its cycle length is expected about 450 EFPD directly from initial core, although most of previous other plants had been started according to their annual or 15-month cycle operation schedule at their initial core and gone to 18-month after third - fourth cycle. In order to reduce the peaking power, fuel pin configurations of the assembly, are optimized by using some low enriched fuel pins and gadolinia bearings. APR+ core has been met the requirements as well as the above cycle length requirement; 1) peaking factor, 2) Negative MTC(Moderator Temperature Coefficient), 3) sufficient shutdown margin, 4) convergent Xenon stability Index. The maximum rod burnup and the discharge fuel assembly burnup are also satisfied those of the limit. It is expected to acquire the standard design approval by the end of 2012 by the Korean nuclear regulatory. (authors)

  4. 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)

  5. 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

  6. Aspects of long pulse commercial tokamak reactor design

    International Nuclear Information System (INIS)

    Possibilities of very long burn pulse operation in commercial tokamak reactors are analyzed. Long pulse operation could reduce or eliminate lifetime limits due to fatigue in the first wall, blanket and structural material and make the reactor more attractive from the standpoint of the electric utility. Very long pulse lengths can be achieved with tokamaks using inductive current drive if the major radius is increased to accomodate a large ohmic transformer. 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

  7. Global stability analysis of pressurized water reactor core nonlinear system

    International Nuclear Information System (INIS)

    Determining the global stability of a pressurized water reactor (PWR) core nonlinear system is the problem to be solved. In the paper, the core nonlinear system was modeled and the linearized model of the system was obtained via the small perturbation method. According to the distributing situation of the core nonlinearity measure in the power level range based on the equilibrium manifold, seven linear models corresponding to seven power levels respectively were chosen as local models of the core and the set of seven local models was used to approximately substitute the core system. The global stability of the PWR core nonlinear system was analyzed by utilizing Lyapunov stability theory. The calculated result shows that the core nonlinear system is globally and asymptotically stable. The modeling method of the core is effective in analyzing the global stability of a PWR core nonlinear system. (authors)

  8. Axial heterogeneous core concept applied for super phoenix reactor

    International Nuclear Information System (INIS)

    Always maintaining the current design rules, this paper presents a parametric study on the type of axial heterogeneous core concept (CHA), utilizing a core of fast reactor Super Phenix type, reaching a maximum thermal burnup rate of 150000 M W d/t and being managed in single batch. (author)

  9. 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)

  10. Development of a core follow calculational system for research reactors

    International Nuclear Information System (INIS)

    Over the last few years a comprehensive Pressurized Water Reactor and Materials Testing Reactor core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments. 10 refs., 1 tab., 1 fig

  11. 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)

  12. European ERANOS formulaire for fast reactor core analysis

    International Nuclear Information System (INIS)

    ERANOS code scheme was developed within the European collaboration on fast reactors. It contains all the functions required to calculate a complete set of core, shielding and fuel cycle parameters for LMFR cores. Nuclear data are taken from recent evaluations (JEF2.2) and adjusted on integral experiments (ERALIB1). Calculational scheme uses the ECCO cell code to generate cross section data. Whole core calculations are carried out using the spatial modules BISTRO (Sn) and TGVNARIANT (nodal method). Validation is based on integral and power reactor experiments. Integral experiments are also used for adjustment of nuclear data

  13. 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)

  14. 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

  15. 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)

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

    International Nuclear Information System (INIS)

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

  17. Vessel core seismic interaction for a fast reactor

    International Nuclear Information System (INIS)

    This report deals with the analysis carried out in collaboration between ENEA and NIRA for optimizing the iterative procedure applied for the evaluation of the effects of the vessel core dynamic interaction for a fast reactor in the case of a earthquake. In fact, as shown in a previous report the convergence of such procedure was very slow for the design solution adopted for the PEC reactor, i.e. with a core restraint plate located close to the top of the core elements. This study, although performed making use of preliminary data (the same of the cited previous report) demonstrates that the convergence is fast if a suitable linear core model is applied in the first iteration linear calculations carried out by NIRA, with an intermediate stiffness with respect to those corresponding to the two limit models previously assumed and increased damping coefficients. Thus, the optimized iterative procedures is now applied in the PEC reactor block seismic verification analysis

  18. 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

  19. The mechanical integrity of fuel pin cladding in a pulsed-beam accelerator driven subcritical reactor

    International Nuclear Information System (INIS)

    Highlights: ► We develop the PTS-ADS code to study transients in ADSR cladding. ► We study thermal response in an ADSR cladding to pulsed beam operation. ► We perform thermal fatigue analysis. ► The cladding mechanical integrity can be assumed unaffected by repetitive temperature variations due to pulsed beam operation. - Abstract: The Accelerator Driven Subcritical Reactor (ADSR) is one of the reactor designs proposed for future nuclear energy production. Interest in the ADSR arises from its enhanced and intrinsic safety characteristics, as well as its potential ability to utilize the large global reserves of thorium and to burn legacy actinide waste from other reactors and decommissioned nuclear weapons. The ADSR concept is based on the coupling of a particle accelerator and a subcritical core by means of a neutron spallation target interface. One of the candidate accelerator technologies receiving increasing attention, the Fixed Field Alternating Gradient (FFAG) accelerator, generates a pulsed proton beam. This paper investigates the impact of pulsed proton beam operation on the mechanical integrity of the fuel pin cladding. A pulsed beam induces repetitive temperature changes in the reactor core which lead to cyclic thermal stresses in the cladding. To perform the thermal analysis aspects of this study a code that couples the neutron kinetics of a subcritical core to a cylindrical geometry heat transfer model was developed. This code, named PTS-ADS, enables temperature variations in the cladding to be calculated. These results are then used to perform thermal fatigue analysis and to predict the stress-life behaviour of the cladding.

  20. Uncertainty Evaluation for Core Thermal Power in a Research Reactor

    International Nuclear Information System (INIS)

    The Jordan Research and Training Reactor (JRTR) also has three separated and independent channels of the neutron detectors to measure the core power. To calibrate these detectors, the thermal power of Primary Cooling System (PCS) which cools down the heat generated in reactor core is used as calibration reference. The core thermal power can be estimated by the measured values of the mass flow rate, core inlet temperature, and core outlet temperature of coolant in the PCS. In general, the uncertainty of the core thermal power is required to be controlled below a certain value. To meet this requirement, the uncertainty of core thermal power should be evaluated based on the uncertainty of the measured parameters. In this paper, the uncertain evaluation is conducted with variation of the uncertainty of the measured parameters such as mass flow rate, core inlet temperature, core outlet temperature. In addition, the numbers of inlet and outlet temperature are considered to get a higher allowable uncertainty of temperature sensors. The core thermal power uncertainty has been valuated according to measuring parameters such as mass flow rate, temperatures, and number of RTDs. In this parametric study, allowable uncertainties for measuring devices have been obtained to guarantee 5% of the core thermal power uncertainty

  1. Uncertainty Evaluation for Core Thermal Power in a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sunil; Seo, Kyoung-Woo; Kim, Seong-Hoon; Chi, Dae-Young; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The Jordan Research and Training Reactor (JRTR) also has three separated and independent channels of the neutron detectors to measure the core power. To calibrate these detectors, the thermal power of Primary Cooling System (PCS) which cools down the heat generated in reactor core is used as calibration reference. The core thermal power can be estimated by the measured values of the mass flow rate, core inlet temperature, and core outlet temperature of coolant in the PCS. In general, the uncertainty of the core thermal power is required to be controlled below a certain value. To meet this requirement, the uncertainty of core thermal power should be evaluated based on the uncertainty of the measured parameters. In this paper, the uncertain evaluation is conducted with variation of the uncertainty of the measured parameters such as mass flow rate, core inlet temperature, core outlet temperature. In addition, the numbers of inlet and outlet temperature are considered to get a higher allowable uncertainty of temperature sensors. The core thermal power uncertainty has been valuated according to measuring parameters such as mass flow rate, temperatures, and number of RTDs. In this parametric study, allowable uncertainties for measuring devices have been obtained to guarantee 5% of the core thermal power uncertainty.

  2. Training reactor deployment. Advanced experimental course on designing new reactor cores

    International Nuclear Information System (INIS)

    Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

  3. Physics design of an ultra-long pulsed tokamak reactor

    International Nuclear Information System (INIS)

    A pulsed tokamak reactor driven only by inductive current drive has recently revived, because the non-inductive current drive efficiency seems to be too low to realize a steady-state tokamak reactor with sufficiently high energy gain Q. Essential problems in pulsed operation mode is considered to be material fatigue due to cyclic operation and expensive energy storage system to keep continuous electric output during a dwell time. To overcome these problems, we have proposed an ultra-long pulsed tokamak reactor called IDLT (abbr. Inductively operated Day-Long Tokamak), which has the major and minor radii of 10 m and 1.87 m, respectively, sufficiently to ensure the burning period of about ten hours. Here we discuss physical features of inductively operated tokamak plasmas, employing the similar constraints with ITER CDA design for engineering issues. (author) 9 refs., 2 figs., 1 tab

  4. Core Management of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    The Dalat nuclear research reactor (DNRR) is a pool-type research reactor which was reconstructed from the 250 kW TRIGA-MARK II reactor. The reconstructed reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. The DNRR uses Russian fuel assemblies, type WWR-SM. The first fuel reloading was executed in April 1994 after more than ten years of operation with 89 fuel assemblies. Research on core management of DNRR with the purpose of maintaining safe operation and effective utilization of reserve fuel assem- blies is being carried out at the Nuclear Research Institute. Calculations of fuel burn-up for the Dalat nuclear research reactor are carried out based on the cell calculation program WIMS and two diffusion calculation programs HEXAGA and HEXNOD. Experimental measurement of fuel burn-up for the Dalat nuclear research reactor was realized by a measurement method of long-life isotopes from fission products. Optimum second fuel reloading and future refuelling for DNRR have been gained. A second fuel reloading for the Dalat nuclear research reactor was realized in March 2002. After reloading the working configuration of the reactor, the core consisted of 104 fuel assemblies. Research results for future refuelling for DNRR show that with 36 reserve fuel assemblies, the reactor will be operated for at least 17 851 h at nominal power since the second fuel reloading. (author)

  5. Nodal expansion method for reactor core calculations

    International Nuclear Information System (INIS)

    To perform realistic space dependent reactor dynamics analyses in large power reactor with all asymmetric material, control and shutdown devices, a full three dimensional calculation model is essential. A code FEMINA (Flux Expansion Method In Nodal Analysis) implementing a higher order nodal scheme employing a nodal flux expansion method in 3D is being developed. In this report the first part of this code viz., the theory of the static version and its validation with well known benchmark problems are described. The code has been found to be quite accurate as well as fast. It is available on DEC 10'', CYBER 170/730 and ND 540 computers. (author)

  6. Reactor Physics and Core Design Issues: India

    International Nuclear Information System (INIS)

    In ADS, since reactor control system is not required to maintain criticality, it is possible to increase burnup i.e. to extract more energy from a given mass of fuel till such time that the keff of the system falls to a value below which it is no more economical to maintain the fission power merely by increasing accelerator current. An interesting argument in support of ADS-based thorium utilization emanates from possibility of starting such reactor without a seed fissile species. On the basis to these possibilities, development of calculation codes and some investigative simulations of ADS operation with them were carried out, which are presented in the following section

  7. MTR (Materials Testing Reactors) cores fuel management. Application of a low enrichment reactor for the equilibrium and transitory core calculation

    International Nuclear Information System (INIS)

    This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun number of movements by rechange. The methodology proposed allows to determine the best options through conceptual relations, prior to a detailed calculation with the core code, reducing the test number with these codes and minimizing in this way CPU cost. The way to better systematized search of transient cores from the first one to the equilibrium one is presented. (Author)

  8. A Core design study on the fuel displacement options for an effective transition between breakeven and TRU burning fast reactors

    International Nuclear Information System (INIS)

    A core design study to convert a breakeven core into a TRU burner is performed for a 600 MWe rated fast reactor. No change in the core and subassembly layouts is assumed, which only allows geometry variations within the fuel rods. Investigated alternatives are to use variable cladding thicknesses, smearing fraction adjustments and annular fuel rod concepts with a central liner of a variable diameter consisting of void, Zr, B4C, Al, W, etc. The variable cladding thickness concept could not be employed due to too high a clad inner wall temperature. A smearing fraction adjustment below a typical fraction of 75% leads to a moderate TRU burning and reduced sodium void worth, but to a relatively high burnup swing. Placing a central non-fuel rod with the fuel arranged in an annular ring affects the core performance and reactivity coefficients, depending on whether it is a moderator or an absorber. In general, candidate materials of high atomic numbers contribute to large positive sodium void worths, but enhanced negative expansion effects. Among the light elements, vanadium reveals a favourable performance with a comparable TRU burning and a reduced sodium void worth, suggesting this material can be regarded as a solid substitute for sodium. (authors)

  9. 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

  10. Seismic study on high temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    The resistance against earthquakes of a high temperature gas-cooled reactor (HTGR) core with block-type fuel is not yet fully ascertained. Seismic studies must be made if such a reactor plant is to be installed in the areas with frequent earthquakes. The experimental and analytical studies for the seismic response of the HTGR core were carried out. First, the fundamental behavior, such as the softening characteristic of a single stacked column (which is piled up with blocks) and the hardening characteristic with the block impact were clarified from the seismic experiments. Second, the displacement and the impact characteristics of the two-dimensional vertical core and the two-dimensional horizontal core were studied from the seismic experiments. Finally, analytical methods and computer programs for the seismic response of HTGR cores were developed. (author) 57 refs

  11. 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)

  12. 3D computer visualization and animation of CANDU reactor core

    International Nuclear Information System (INIS)

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

  13. Electricity production from a pulsed tokamak fusion reactor

    International Nuclear Information System (INIS)

    A study has been undertaken to investigate the use of a possible pulsed fusion reactor to supply the national grid. Detailed models of the individual components of a 1200 MWe reactor plant were developed, including the reactor blanket, boiler and turbine generator. Using a drum boiler as a thermal energy store, full output could be maintained for reactor off-periods up to only 40 seconds, compared with an expected off-period for a pulsed tokamak fusion reactor of up to 300 seconds. Two possible solutions to this mis-match problem are considered, the first involving an externally fired superheater and reheater, which would allow the off-period to be extended to 105 seconds, and a second involving an auxiliary boiler, which would allow an indefinite off-period. Under these conditions, the plant and operating costs are estimated to be higher than the estimated cost of incorporating non-inductive current drive into a tokamak, and therefore the study suggests that it would be advantageous to develop a continuously operating tokamak fusion reactor, although other possible solutions relevant to the pulse operation should be further investigated. (Author)

  14. 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.

  15. 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.)

  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. In-core fuel management for pebble-bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Milian Perez, Daniel; Rodriguez Garcia, Lorena; Garcia Hernandez, Carlos; Milian Lorenzo, Daniel, E-mail: dperez@instec.cu, E-mail: cgh@instec.cu, E-mail: dmilian@instec.cu [Higher Institute of Technologies and Applied Sciences, Havana (Cuba); Velasco, Abanades, E-mail: abanades@etsii.upm.es [Department of Simulation of Thermo Energy Systems, Polytechnic University of Madrid (Spain)

    2013-07-01

    In this paper a calculation procedure to reduce the power peak in the core of a Very High Temperature pebble bed Reactor is presented. This procedure combines the fuel depletion and the neutronic behavior of the fuel in the reactor core, modeling once-through-then-out cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times, obtaining the asymptotic fuel-loading pattern. The procedure consists in several coupled computational codes, which are used iteratively until convergence is reached. The utilization of the MCNPX 2.6e, as one of these computational codes, is validated through the calculation of benchmarks announced by IAEA (IAEA-TECDOC-1249, 2001). To complete the verification of the calculation procedure a base case described in (Annals of Nuclear Energy 29 (2002) 1345-1364), was performed. The procedure has been applied to a model of Pebble Bed Modular Reactor (200 MW) design. (author)

  18. 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...

  19. IBR-2: A periodically operating pulsed reactor for neutron research

    International Nuclear Information System (INIS)

    An extensive description of the IBR-2 fast neutron pulse reactor is given. The operation and construction of the reactor, the control and safety systems and the cooling system are described in details. In the booster mode of operation a linear induction accelerator is applied as injector. The shielding problems and the experimental possibilities are also discussed. The particular problems in the design, construction and operation of such reactors are reviewed. The general characteristics of the IBR-2 as a neutron source for spectrometry as well as a list of the experimental apparatus under construction are given. (R.J.)

  20. Pulse reactor system for nuclear pumped laser using low enriched uranium

    International Nuclear Information System (INIS)

    Nuclear Pumped Laser (NPL) is a laser technology in which laser oscillation is done by energy of nuclear reaction. It is expected that NPL by nuclear fission make it possible to broaden the use of nuclear energy, which is now limited to electricity generation and the use of nuclear heat. It is often call as Reactor Pumped Laser (RPL), but in this paper, it is called as NPL simply. In NPL (RPL), fission fragments by nuclear fissions fly through Ar-Xe gas medium in laser cells, and the kinetic energy of the fission fragments do pumping of the laser oscillation. One of the designs of nuclear reactor system for NPL is coupling of high enriched uranium metal pulse cores and laser oscillation cells, whose inside surface is coated with high enriched uranium. Laser oscillation experiments have been performed using this type of system in IPPE, Russia, and it has been shown the laser oscillation was possible. The use of high enriched uranium is limited in the view point of non-proliferation. For research and even for practical use of NPL, reactor system using low enriched uranium is needed. In the paper, the possibility of pulse reactor system for NPL using low enriched uranium is discussed. For the neutronic analysis, continuous energy Mote Carlo code MVP was used with JENDL-3.3 nuclear data library. By the analysis, it became clear that it was possible to design a reactor system with 20% enriched uranium by making the size of pulse cores large and by making the position and number of the cores proper. The system is possible to give enough energy to the laser oscillation cells for laser pumping. It is also shown that this system has more flatten power distribution in laser oscillation cells, which is desirable for laser pumping. For the detail analysis of laser pumping, kinetic analysis in pulse operation is needed. For the analysis, time dependent neutron coupling factors were needed between each reactor region. For the calculation, modified MVP Mote Carlo code was

  1. 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.

  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. 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

  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. A study of passive safety conditions for fast reactor core

    International Nuclear Information System (INIS)

    A study has been made for passive safety conditions of fast reactor cores. Objective of the study is to develop a concept of a core with passive safety as well as a simple safety philosophy. A simple safety philosophy, which is wore easy to explain to the public, is needed to enhance the public acceptance for nuclear reactors. The present paper describes a conceptual plan of the study including the definition of the problem a method of approach and identification of tasks to be solved

  6. Core burnup characteristics of high conversion light water reactor, (1)

    International Nuclear Information System (INIS)

    In order to evaluate core burnup characteristics of a high conversion light water reactor (HCLWR) with tight pitched lattice, core burnup calculation was made using two dimensional diffusion method. The volume ratio of moderator to fuel is about 0.8 in the reactor (HCLWR-J1) under study. The burnup calculations were carried out under the assumption of three batch and out-in fuel loading from the first cycle to the equilibrium cycle. A detailed evaluation was made for discharge burnup, conversion ratio, power distribution, and reactivity coefficients and so on. (author)

  7. 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.

  8. Physics research for molten salt reactor with different core boundaries

    International Nuclear Information System (INIS)

    Background: Unlike the traditional solid fuel reactor with fixed boundary conditions, the inlet and outlet pipe and the core of molten salt reactor fuel is connected so that the flowing liquid fuel can travel freely between the pipe and core. Purpose: This article has made systematic study of the influence of different molten salt fuel regions on reactor physics, including the top and bottom fuel of the core vessel and the pipe fuel. The core physics was researched under different boundary conditions, and the region of the effective core was indicated subsequently. Methods: MSRE was taken as the reference reactor and the calculation was completed based on the Monte Carlo Code MCNP. Results: Results show that the fuel in the top and bottom of vessel impacts on keff and energy spectrum obviously. The influence of outlet pipe on keff was negligible when pipe radius less than 25 cm, and the perturbation of outlet pipe on the keff could be neglected when its length more than 20 cm. Conclusions: Results provide rational theory for the design of MSR and the development of computation code. (authors)

  9. 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.

  10. 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)

  11. Fast-core thermal-blanket breeder reactor

    International Nuclear Information System (INIS)

    A preliminary assessment of the performance expected from a specific type of FCTB reactor, consisting of a gas-cooled fast system for the core and natural-uranium light-water thermal system for the blanket is reported. Both the core and the blanket use the 238U-Pu fuel cycle. When all the neutrons leaking out of the core reach the blanket, the blanket-to-core power ratio is estimated to be about 1.3. By reducing its water-to-fuel volume ratio below 1.5, the light water blanket can be designed to have a higher ksub(eff), while maintaining an equilibrium fissile fuel content. Compared with conventional FBRs, having the same power output, the FCTB reactor considered offers the following advantages: a lower fissile fuel content, easier and safer control, no need for Pu separation. (B.G.)

  12. Polarized neutron reflectometry at the IBR-2 pulsed reactor

    Science.gov (United States)

    Aksenov, V. L.; Nikitenko, Yu. V.

    2007-05-01

    Polarized neutron reflectometry as a method for investigating layered nanostructures and its implementation at the IBR-2 pulsed reactor (Dubna, Russia) are described. The experimental data illustrating the studies of magnetic layered nanostructures and the development of the method of polarized neutron reflectometry on the polarized neutron spectrometer are presented. The directions of further development of the method of polarized neutron reflectometry are analyzed.

  13. 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

  14. Hydraulic characteristics of the N Reactor core and reactor cooling system

    International Nuclear Information System (INIS)

    In conjunction with the NUSAR program, the need was recognized for well substantiated pressure drop correlations for the N Reactor core to support in-depth safety analysis consistent with currently-available technology. Additionally, it was considered desirable to reconfirm the hydraulic characteristics of the reactor coolant system in the light of improved understanding of the hydraulic features of the current reactor fuel loading. The report summarizes the results of laboratory tests and analysis accomplished to meet the above objectives

  15. Assessment of TRIGA pulsing reactor safety without loss of coolant

    International Nuclear Information System (INIS)

    The fuel temperature, fuel can pressure and cladding stresses in the TRIGA fuel elements at various transient states has in part been measured and semiempirically estimated in cases where no experiments could be carried out. To the end of obtaining more reliable and up-to-date data and reactor parameters, reactor power and fuel temperatures were measured both at steady and transient states, including pulsing up to 280 MW. A novel and more realistic model for the fuel temperature at pulsing was also presented and used in the assessments. Based on the present assessment with the water coolant present and a maximum excess reactivity of (Δksub(max)/β = ) 4 $, no reactivity induced transients can bring about any undue hazards to the reactor or to the surroundings. (author)

  16. Neutronic decoupling characteristics of large fast reactor cores

    International Nuclear Information System (INIS)

    It was made clear by critical experiments (called JUPITER) that nuclear characteristics of large-sized fast reactor cores were quite different from those of small-sized cores. For instance, radial neutron flux distributions are significantly changed by perturbations, control rod reactivity interaction effects are large, etc. These phenomena are interpreted to be commonly caused by the spatial neutronic decoupling of large core. The changeability and instability of flux distribution, which might cause a power peaking and a flux tilting, is a new problem for the development of large core. The paper shows measured results of static decoupling characteristics, interprets them physically, and considers a reactor physics system of large core. The paper also investigates a nuclear core design method, in which the decoupling characteristics are taken into account. A neutronic stability is a new requirement for the nuclear design of large core, which the design policy is to promote and secure together with the performance and the safety. The more tightly coupled the core, i.e., the smaller the degree of decoupling, the better the neutronic stability. (author)

  17. 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)

  18. Preliminary core design calculations for the ACPR Upgrade

    International Nuclear Information System (INIS)

    The goal of the Annular Core Pulse Reactor (ACPR) Upgrade design studies is to define a core configuration that provides a significant increase in pulse fluence and fission energy deposition. The reactor modification should provide as flat an energy deposition profile for experiments as feasible. The fuels examined in this study were UO2-BeO (5-15 w/o UO2), UC-ZrC-C (200-500 mg U/cc) and U-ZrH1.5. The basic core concept examined was a two region core, - a high heat capacity inner core region surrounded by an outer U-ZrH1.5 region. Survey core calculations utilizing 1D transport calculations and cross sections libraries derived from the ORNL-AMPX code examined relative fuel loadings, fuel temperatures, reactivity requirements and pulse performance improvement. Reference designs for all candidate fuels were defined utilizing 2D transport and Monte Carlo calculations. The performance implications of alternative core designs were also examined for the UO2-BeO and UC-ZrC-C fuel candidates. (author)

  19. Experimental research of reactor core flooding

    International Nuclear Information System (INIS)

    The results are presented of experiments performed with the aim of finding the influence of the method of fixing the thermocouples for measuring the distribution of temperature to the wall of fuel pin simulator. This influence was found for the purpose of emergency core flooding. First experimental results on the effect of nitrogen dissolved in the water on the velocity of the cooling wave are given. These experiments were carried out under the following conditions: initial temperature in pin centre 300 to 600 degC, velocity of water at the inlet into the measuring section 3.5 to 20 cm/s, and atmospheric pressure in the model. (author)

  20. 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)

  1. 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)

  2. 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)

  3. Seismic behavior of a fuel assembly in the reactor core

    International Nuclear Information System (INIS)

    A general approach to the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced from earthquake. The dynamic responses such as fuel assembly deflected shapes and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed

  4. Design Requirements of an Advanced HANARO Reactor Core Cooling System

    International Nuclear Information System (INIS)

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

  5. Cooling of core debris within the reactor vessel lower head

    International Nuclear Information System (INIS)

    Under severe-accident conditions, the most crucial action for recovery from the accident state is to cool the core debris and prevent or terminate attack on the remaining fission product barriers. One means of preventing attack on the containment structures is to retain the core debris within the reactor vessel. The Three Mile Island Unit 2 (TMI-2) accident demonstrated that this could be accomplished by water resident within the reactor vessel combined with injection on a continual basis to quench the debris and remove decay heat over the long term. Some accident situations could result in the transport of molten core debris to the lower plenum, as occurred in TMI-2, the boil-off of water in the lower plenum, and an inability to add water to the reactor coolant system (RCS). Even in this extreme set of circumstances, sufficient cooling may be available to prevent failure of the reactor pressure vessel (RPV) lower head and thereby retain the core debris within the vessel. Experiments were performed in support of Commonwealth Edison's Zion individual plant examination and accident management programs that demonstrate nucleate boiling heat removal rates from the outer surface of a simulated RPV lower head surrounded by typical reflective insulation used in nuclear power plants

  6. Cooling of core debris within the reactor vessel lower head

    International Nuclear Information System (INIS)

    Under severe accident conditions, the most crucial action for recovery from the accident state is to cool the core debris and prevent or terminate attack on the remaining fission product barriers. One means of preventing attack on the containment structures is to retain the core debris within the reactor vessel. Some accident situations could result in the transport of molten core debris to the lower plenum, as occurred to some extent (∼20 tonnes) during the TMI-2 accident, boiloff of water in the lower plenum, and an inability to add water to the reactor coolant system (RCS). In this extreme set of circumstances, sufficient external reactor pressure vessel (RPV) cooling may be available to prevent failure of the RPV lower head and, thereby, retain the core debris within the vessel. Containment configurations like Zion would result in substantial accumulation of water around the lower parts of the reactor vessel for most accident sequences. The experiments which were performed in support of the Commonwealth Edison individual plant examination and accident management programs, are heat transfer tests designed to demonstrate that nucleate boiling is the dominant heat removal process from the outer surface of a simulated RPV lower head surrounded by typical reflective insulation used in nuclear power plants

  7. Nuclear detectors for in-core power-reactors

    International Nuclear Information System (INIS)

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

  8. 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

  9. The uranium-zirconium hydride pulsed reactor and its use in science and technology

    International Nuclear Information System (INIS)

    The performance of the first Chinese pulsed reactor is described briefly. The second reactor being built in China has a large prompt negative temperature coefficient of reactivity and uses uranium-zirconium hydride alloy as fuel element. Therefore its most outstanding features are its 'inherent safety' fairly high pulsed power capacity. The pulsed reactor is now widely used in science and technology. (orig.)

  10. 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 ...

  11. 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.

  12. Analysis of the seismic response of a fast reactor core

    International Nuclear Information System (INIS)

    This report deals with the methods to apply for a correct evaluation of the reactor core seismic response. Reference is made to up-to-date design data concerning the PEC core, taking into account the presence of the core-restraint plate located close to the PEC core elements top and applying the optimized iterative procedure between the vessel linear calculation and the non-linear ones limited to the core, which had been described in a previous report. It is demonstrated that the convergence of this procedure is very fast, similar to what obtained in the calculations of the cited report, carried out with preliminary data, and it is shown that the cited methods allow a reliable evaluation of the excitation time histories for the experimental tests in support of the seismic verification of the shutdown system and the core of a fast reactor, as well as relevant data for the experimental, structural and functional, verification of the core elements in the case of seismic loads

  13. 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.

  14. Computer code for nuclear reactor core thermal reliability calculation

    International Nuclear Information System (INIS)

    RASTENAR program was described for computing heat-engineering reliability of cores in nuclear reactors operating under stationary conditions. The following factors of heat-engineering reliability were found to be computable: rated critical margin; limiting critical margin; probability of initiation of critical heat removal in channel (inferior conditions of heat transfer); probability that no channel would be subject to critical heat removal; and reactor power reserve coefficient. The probability that no channel in the core would experience critical heat removal when boiling during operation of the reactor at fixed power level was taken for the principal quantitative criterion. The structure and limitations of the program were described together with the computation algorithm. The program was written for an M-220 computer

  15. 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)

  16. REX 2000 core : a new material testing reactor project

    International Nuclear Information System (INIS)

    REX 2000 is a new research reactor project entirely dedicated to technological irradiations, which should be located on the CEA site of CADARACHE. It will be aimed at satisfying the future needs for the validation of new concepts of nuclear materials and fuels, and will take over and replace the present experimental reactors, which are 30 to 40 years old. The fundamental studies started by the CEA in 1993, on future irradiation needs expected in 2005, lead to the design of a reactor which will essentially meet the needs of PWRs, without forgetting the other fields such as FBRs, fusion... The current reactor project is based on a light water open pool concept, with a thermal power of 100 MW, in about 150 l, and characterized by an in-core-central hole. It reaches neutronic flux levels twice those of present French reactor fluxes. It allows many irradiations in the central loop under high fast neutron flux, in order to accelerate the aging of materials and analyze their behaviour. It also enables the achievement of power transient tests under high thermal neutron flux gradients. These performances are obtained with high forced flow rates and upward flow in the core, in order to preserve the operating flexibility of the reactor. This leads to the design of a specific assembly design. (author)

  17. Core and transition fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    A core and a transition fuel assembly for a nuclear reactor are described. They have a first fuel assembly including structure for laterally spacing parallel and coextending fuel rods positioned at preselected core elevations and also a second fuel assembly including lateral spacing structure at preselected core elevations at least one of which is different than the elevations of the spacing structure of the first fuel assembly. The transition fuel assembly is positioned between the first and second assemblies and includes lateral spacing structure positioned at each core elevation where the first and second fuel assemblies have a spacing structure. The transition fuel assembly ensures that contact among the fuel assemblies of the core is through the spacing structures. 9 claims

  18. Cooling of core debris within the reactor vessel lower head

    International Nuclear Information System (INIS)

    Under severe accident conditions, the most crucial action for recovery from the accident state is to cool the core debris and prevent or terminate attack on the remaining fission product barriers. One means of preventing attack on the containment structures is to retain the core debris within the reactor vessel. The TMI-2 accident demonstrated that this could be accomplished by water resident within the reactor vessel combined with injection on a continual basis to quench the debris and remove decay heat over the long term. Some accident situations could result in the transport of molten core debris to the lower plenum, as occurred in TMI-2, the boiloff of water in the lower plenum, and the inability to add water to the reactor coolant system (RCS). Even in this extreme set of circumstances, sufficient cooling may be available to prevent failure of the reactor pressure vessel (RPV) lower head and thereby retain the core debris within the vessel. Containment configurations like Zion would result in substantial accumulation of water around the lower parts of the reactor vessel for most accident sequences. For some PWR containments, there could be substantial water accumulation around the reactor vessel and the hot and cold legs. If this water could directly contact the carbon steel vessel surface and RCS piping, substantial energy could be removed from the primary system and in particular the RPV lower head. Experiments discussed in this paper demonstrate nucleate boiling heat removal rates from the outer surface of a simulated RPV lower head surrounded by typical reflective insulation used in nuclear power plants

  19. 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.

  20. Computer based core monitoring system for an operating CANDU reactor

    International Nuclear Information System (INIS)

    The research was performed to develop a CANDU-6 Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong nuclear power plant unit 1. The CCMS uses Reactor Fueling Simulation Program(RFSP, developed by AECL) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from Digital Control Computer(DCC) for the purpose of producing basic input data. The CCMS has two modules; CCMS server program and CCMS client program. The CCMS server program performs automatic and continuous core calculation and manages overall output controlled by DataBase Management System. The CCMS client program enables users to monitor current and past core status in the predefined GUI(Graphic-User Interface) environment. For the purpose of verifying the effectiveness of CCMS, we compared field-test data with the data used for Wolsong unit 1 operation. In the verification the mean percent differences of both cases were the same(0.008%), which showed that the CCMS could monitor core behaviors well

  1. Reference Monte Carlo calculations of Maria reactor core

    International Nuclear Information System (INIS)

    The reference Monte Carlo calculations of MARIA reactor core have been carried to evaluate accuracy of the calculations at each stage of its neutron-physics analysis using deterministic codes. The elementary cell has been calculated with two main goals; evaluation of effects of simplifications introduced in deterministic lattice spectrum calculations by the WIMS code and evaluation of library data in recently developed WIMS libraries. In particular the beryllium data of those libraries needed evaluation. The whole core calculations mainly the first MARIA critical experiment and the first critical core after the 8-year break in operation. Both cores contained only fresh fuel elements but only in the first critical core the beryllium blocks were not poisoned by Li-6 and He-3. Thus the MCNP k-eff results could be compared with the experiment. The MCNP calculations for the cores with beryllium poisoned suffered the deficiency of uncertainty in the poison concentration, but a comparison of power distribution shows that realistic poison levels have been carried out for the operating reactor MARIA configurations. (author)

  2. Development and assessment of advanced reactor core protection system

    International Nuclear Information System (INIS)

    An advanced core protection system for a pressurized water reactor, Reactor Core Protection System (RCOPS), was developed by adopting a high performance hardware platform and optimal system configuration. The functional algorithms of the core protection system were also improved to enhance the plant availability by reducing unnecessary reactor trips and increasing operational margin. The RCOPS consists of four independent safety channels providing a two-out-of-four trip logic. The reliability analysis using the reliability block diagram method showed the unavailability of the RCOPS to be lower than the conventional system. The failure mode and effects analysis demonstrated that the RCOPS does not lose its intended safety functions for most failures. New algorithms for the RCOPS functional design were implemented in order to avoid unnecessary reactor trips by providing auxiliary pre-trip alarms and signal validation logic for the control rod position. The new algorithms in the RCOPS were verified by comparing the RCOPS calculations with reference results. The new thermal margin algorithm for the RCOPS was expected to increase the operational margin to the limit for Departure from Nucleate Boiling Ratio (DNBR) by approximately 1%. (author)

  3. 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

  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. Performance of piezoresistive and piezoelectric sensors in pulsed reactor experiments

    International Nuclear Information System (INIS)

    Pulsed reactor-based experiments require radiation tolerant sensors that do not perturb the device under test, or allow a radiation-induced signal to mask the true sensor output. Several commercial off-the-shelf accelerometers, pressure transducers, and acoustic emission sensors were subjected to multiple high-power reactor pulses. A piezoresistive accelerometer capable of operation to at least 44 kGy and 8.7 x 1015 n/cm2 is identified, and a piezoresistive pressure transducer that is resistant to about half that radiation level is selected. Further, two piezoelectric acoustic emission sensors employing lead metaniobate are also found to function to 55 kGy and 1.1 x 1016 n/cm2.

  6. 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.

  7. Water regime and core state in WWER reactors

    International Nuclear Information System (INIS)

    Some features of the ammonia-potassium water regime used in the WWER-reactors are considered. This regime makes it possible to suppress water radiolysis and simultaneously obtain a high pH value of the coolant, which is necessary in operation with boron control of the reactor power. Absence of oxygen and significant deposits on the zirconium alloy claddings of the fuel elements affect favourably the corrosion condition of the core. The thickness of the corrosion film on the fuel cladding at the end of their lifetime is as low as 3μm and the hydrogen content remains at the initial level

  8. 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.

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

    International Nuclear Information System (INIS)

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

  10. Pulse Coding Techniques for Reactor Safety Logic Systems

    International Nuclear Information System (INIS)

    Reactor safety logic systems for use on large nuclear power stations must provide an extremely high reliability for shutting down the plant for genuine fault conditions combined with the conflicting requirement of relative freedom from unnecessary shutdowns due to minor faults, particularly in the logic system. Most systems employ redundancy - often in the form of three or more instruments measuring each parameter - together with a majority vote technique. The latter frequently takes the form of danger signals from two out of three instruments measuring the same parameter being necessary to cause shut-down. The use of solid-state logic elements such as transistors has given rise to a rapid advancement of techniques. The paper deals with the majority vote part of such a system and the problem of being able to test that this part of the system is functioning correctly. A number of pulse coding techniques have been developed in which the input from each instrument is a coded pattern of pulses rather than a continuous train of pulses. Under normal conditions these pulse code patterns combine in the majority logic network to produce a continuous train of pulses at the output. If a single input disappears or a fault in the network develops, then the output pulse pattern changes giving information on the fault but not causing plant shut-down, A majority of inputs missing gives zero output and results in shut-down. Information is given on two different types of system that have these properties. (author)

  11. High performance light water reactor core design studies

    International Nuclear Information System (INIS)

    The High Performance Light Water Reactor (HPLWR) is the European version of the various supercritical water cooled reactor proposals. The paper presents the activity of KFKI-AEKI in the field of neutronic core design within the framework of the European funded project: 'HPLWR Phase 2'. As the coolant density along the axial direction shows remarkable change, coupled neutronic-thermohydraulic calculations are essential which take into account the heating of moderator in the special water rods of the assemblies. A parametrized diffusion cross section library was prepared for the HPLWR assembly with the MULTICELL neutronic transport code. The parametrized cross sections are used by the KARATE program system, which was verified by comparative Monte Carlo calculations. Methodical core calculations have been carried out. Without thermal insulation in the second superheater of the three pass core, the temperature of moderator water exceeds the pseudocritical temperature resulting in poor moderation. Applying heat insulators helps to avoid the extreme heat up of moderator, so it provides better moderation for neutrons. The increase of the local multiplication in the superheaters favorably affects the power distribution and keff of the core. Preliminary loadings of the HPLWR core were assessed, which contain insulated assemblies with Gd burnable absorbers. The initial reactivity excess was compensated by control rods. The reactivity coefficients of HPLWR with respect to temperatures show substantial negative feedback at normal operating conditions. (author)

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

    International Nuclear Information System (INIS)

    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 UF6 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 233U born from thorium. Fission product removal was continuous. Newly born 233U 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 233U. 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

  13. Survey of dust production in pebble bed reactor cores

    International Nuclear Information System (INIS)

    Highlights: → We review potential sources of the graphite dust found in the German pebble bed reactors. → Available literature on graphite wear coefficients in pebble bed core-like conditions is reviewed. → Limited conclusions and remaining open questions are discussed. - Abstract: Graphite dust produced via mechanical wear from the pebbles in a pebble bed reactor is an area of concern for licensing. Both the German pebble bed reactors produced graphite dust that contained activated elements. These activation products constitute an additional source term of radiation and must be taken under consideration during the conduct of accident analysis of the design. This paper discusses the available literature on graphite dust production and measurements in pebble bed reactors. Limited data is available on the graphite dust produced from the AVR and THTR-300 pebble bed reactors. Experiments that have been performed on wear of graphite in pebble-bed-like conditions are reviewed. The calculation of contact forces, which are a key driving mechanism for dust in the reactor, are also included. In addition, prior graphite dust predictions are examined, and future areas of research are identified.

  14. 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)....

  15. 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)

  16. 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.

  17. Cyclic gaseous core reactors for space nuclear power applications

    International Nuclear Information System (INIS)

    Extensive theoretical-experimental investigations have been performed at the University of Florida on cyclic gaseous core reactors. Neutronics-energetics analyses have led to a basic scientific understanding of the behavior associated with conceptual operation of these devices. Thermal-physical properties studies have resulted in the identification of desirable working fluids and UF6-materials interaction studies have identified a number of potential problems as well as corresponding potential solutions. The results of these research efforts indicate that the cyclic gaseous core reactor is a versatile and promising nuclear energy concept that has attractive features for space power generation. These include low critical mass, high fuel utilization, high output temperature and good thermal efficiency, wide operating ranges, excellent control and safety characteristics, and adaptability to a wide variety of different energy conversion systems

  18. Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Krass, A.W.

    2005-12-19

    This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.

  19. Thermal hydraulic performance assessment of dual-cooled annular nuclear fuel for OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang-Hwan, E-mail: shinch@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Chun, Tae-Hyun, E-mail: thchun@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Oh, Dong-Seok, E-mail: dsoh1@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); In, Wang-Kee, E-mail: wkin@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer A thermal hydraulic performance of a 12 Multiplication-Sign 12 annular fuel array is evaluated. Black-Right-Pointing-Pointer The subchannel analysis code for the dual-cooled annular fuel, MATRA-AF is validated. Black-Right-Pointing-Pointer We evaluate the sensitivity for geometry tolerances and operating parameter. Black-Right-Pointing-Pointer We decide the essential design parameters to uprate the power generation by dual-cooled annular fuel. Black-Right-Pointing-Pointer A thermal margin amount accommodating a 20% power-uprate seems viable. - Abstract: An internally and externally cooled annular fuel was proposed for an advance PWR, which can endure substantial power uprating. KAERI is pursuing the development for a reloading of power uprated annular fuel for the operating PWR reactors of OPR-1000. In this paper, the characteristics and verification of the MATRA-AF are described. The thermal hydraulic performance of a 12 Multiplication-Sign 12 annular fuel is calculated for the major design parameters and its performance is compared against the reference 16 Multiplication-Sign 16 cylindrical fuel assembly. In particular, the enhancements of the thermal hydraulic performance of dual-cooled annular fuel are estimated for the 100% normal power reactor core. The purpose of this study is to estimate a normal power for OPR-1000 with dual-cooled annular fuel, and ultimately to assess the feasibility of 120% core power. The parametric study was carried out for the fuel rod dimension, gap conductance, thermal diffusion coefficients, and pressure loss of the spacer grids. As a result of the analysis on the nominal power, annular fuel showed a sufficient margin available on DNB and fuel pellet temperature relative to cylindrical fuel. The margin amount seems accommodating a 20% power-uprate seems viable.

  20. RASPLAV, Refine accident management strategies during a reactor core meltdown

    International Nuclear Information System (INIS)

    Description: OECD RASPLAV Project. The RASPLAV project aimed to refine accident management strategies during a reactor core meltdown; it was completed in June 2000. Little is known about the complex interactions that take place during a core meltdown, so one of the RASPLAV project's primary goals was to develop an understanding of this process. The information gathered during tests at the Kurchatov Institute have allowed scientists to develop models of a core meltdown. These models can be used in the design of new reactors and in refining the accident procedures for existing ones. Two aspects of the issue were considered. First, for existing reactors, where external cooling may not be practicable, the process and time sequence before melt-through were studied. This was to help develop management strategies for severe accidents. Secondly, for future and some existing reactor designs, the project aimed to determine the heat transfer conditions under which cavity flooding can be a viable accident management option. The project was run in two successive phases. The RASPLAV Phase-2 project investigated the progression of a severe accident and in particular the thermal loading imposed by a corium pool on the lower head of a Light Water Reactor (LWR) vessel. It followed an earlier Phase-1 project dedicated mainly to the build-up of the experimental and analytical infrastructure. The project objectives were to obtain relevant data on the physical and thermal behavior of the corium in large-scale tests, to derive thermal-physical property data for various molten core materials, and to investigate the effects of stratification of molten materials. The programme of work involved the use of the large facilities available at the Kurchatov Institute in Russia. Four large-scale tests were carried out and were complemented by a series of smaller-scale experiments, all involving the use of materials representative of power reactor cores. Experiments with these test materials in

  1. Thermal Analysis of Air-Core Power Reactors

    OpenAIRE

    Zhao Yuan; Jun-jia He; Yuan Pan; Xiao-gen Yin; Can Ding; Shao-fei Ning; Hong-lei Li

    2013-01-01

    A fluid-thermal coupled analysis based on FEM is conducted. The inner structure of the coils is built with consideration of both the structural details and the simplicity; thus, the detailed heat conduction process is coupled with the computational fluid dynamics in the thermal computation of air-core reactors. According to the simulation results, 2D temperature distribution results are given and proved by the thermal test results of a prototype. Then the temperature results are used to calcu...

  2. 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

  3. A reactor core on-line monitoring program - COMP

    International Nuclear Information System (INIS)

    A program named COMP is developed for on-line monitoring PWRs' in-core power distribution in this paper. Harmonics expansion method is used in COMP. The Unit 1 reactor of Daya Bay Nuclear Power Plant (Daya Bay NPP) in China is considered for verification. The numerical results show that the maximum relative error between measurement and reconstruction results from COMP is less than 5%, and the computing time is short, indicating that COMP is capable for online monitoring PWRs. (authors)

  4. 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)

  5. Fast reactor calculational route for Pu burning core design

    International Nuclear Information System (INIS)

    This document provides a description of a calculational route, used in the Reactor Physics Research Section for sensitivity studies and initial design optimization calculations for fast reactor cores. The main purpose in producing this document was to provide a description of and user guides to the calculational methods, in English, as an aid to any future user of the calculational route who is (like the author) handicapped by a lack of literacy in Japanese. The document also provides for all users a compilation of information on the various parts of the calculational route, all in a single reference. In using the calculational route (to model Pu burning reactors) the author identified a number of areas where an improvement in the modelling of the standard calculational route was warranted. The document includes comments on and explanations of the modelling assumptions in the various calculations. Practical information on the use of the calculational route and the computer systems is also given. (J.P.N.)

  6. Computer simulation of pulsed-neutron experiments performed on the Fort St. Vrain high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Computer simulations were used to determine the optimum source location, detector location, and pulse rate prior to performing pulsed-neutron experiments on the 330-MW Fort St. Vrain high-temperature gas-cooled reactor (HTGR). The simulation procedure involved calculation of the amplitudes, decay constants, and modal shapes of the first few kinetic modes in the general expansion of the time response of the neutron flux following each pulse. The kinetic modes were calculated by the eigenfunction expansion method in two-dimensional geometry assuming two energy groups and six delayed-neutron precursors. The major limitation in the calculation is the use of two-dimensional core models, i.e., the assumption of separation of variables. For most power reactors on which pulsed-neutron experiments might be performed, this limitation should not be serious

  7. Assessment of Core Failure Limits for Light Water Reactor Fuel under Reactivity Initiated Accidents

    International Nuclear Information System (INIS)

    Core failure limits for high-burnup light water reactor UO2 fuel rods, subjected to postulated reactivity initiated accidents (RIAs), are here assessed by use of best-estimate computational methods. The considered RIAs are the hot zero power rod ejection accident (HZP REA) in pressurized water reactors and the cold zero power control rod drop accident (CZP CRDA) in boiling water reactors. Burnup dependent core failure limits for these events are established by calculating the fuel radial average enthalpy connected with incipient fuel pellet melting for fuel burnups in the range of 30 to 70 MWd/kgU. The postulated HZP REA and CZP CRDA result in lower enthalpies for pellet melting than RIAs that take place at rated power. Consequently, the enthalpy thresholds presented here are lower bounds to RIAs at rated power. The calculations are performed with best-estimate models, which are applied in the FRAPCON-3.2 and SCANAIR-3.2 computer codes. Based on the results of three-dimensional core kinetics analyses, the considered power transients are simulated by a Gaussian pulse shape, with a fixed width of either 25 ms (REA) or 45 ms (CRDA). Notwithstanding the differences in postulated accident scenarios between the REA and the CRDA, the calculated core failure limits for these two events are similar. The calculated enthalpy thresholds for fuel pellet melting decrease gradually with fuel burnup, from approximately 960 J/gUO2 at 30 MWd/kgU to 810 J/gUO2 at 70 MWd/kgU. The decline is due to depression of the UO2 melting temperature with increasing burnup, in combination with burnup related changes to the radial power distribution within the fuel pellets. The presented fuel enthalpy thresholds for incipient UO2 melting provide best-estimate core failure limits for low- and intermediate-burnup fuel. However, pulse reactor tests on high-burnup fuel rods indicate that the accumulation of gaseous fission products within the pellets may lead to fuel dispersal into the coolant at

  8. 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

  9. Fission product release out of the core of a pebble bed reactor in core heatup accidents

    International Nuclear Information System (INIS)

    This report presents the analysis of fission product release from the core of a pebble-bed high temperature reactor during hypothetical accidents. First the models describing fission product transport are discussed, and on the basis of these models a computer code is developped. This code includes the diffusion of fission products from particles and through the graphite, and the sorption of metallic fission product elements on graphite as well as the plateout of metallic fission product elements in the top- and bottom reflectors. In addition a review of the necessary empirical input data is given. Then the cesium release of a single fuel element at high temperatures is calculated, and the results are compared with experimental data. Furthermore calculations of the fission product release from the core of a 500 MW(th) high temperature reactor during core heatup accidents are made, and the influence of the most important parameters is described. (orig.)

  10. 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

  11. 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.

  12. 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.

  13. Preliminary core characterization for an ELSY-oriented demonstrative reactor

    International Nuclear Information System (INIS)

    An Italian effort has been initiated for the investigation of a concept for a small-size Advanced Lead Fast Reactor which aims at being oriented to the European Lead-cooled SYstem (ELSY). A demonstration reactor is expected to prove the viability of technology for use in a future commercial power plant, construction and operation, with the purpose of attesting the general strategy to use, to the largest extent. This paper aims at outlining the approach followed in establishing the main neutronic features by exploring the logical sequence that underlies a preliminary core characterization and by highlighting design rationales and mutual interdependencies among parameters. The first step towards the conceptual design of an ELSY-oriented reactor has been defining clearly what 'oriented to ELSY' means, i.e. determining those issues of ELSY to investigate/validate in such a reactor and the objectives to reach. As a result, some major objectives have been identified and further scope analyses have been carried out to in order to determine suitable core configurations in reference to a short-term and a medium/long-term schedule. (authors)

  14. Development of an automated core model for nuclear reactors

    International Nuclear Information System (INIS)

    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

  15. 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)

  16. 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)

  17. Core reactor with a reactor core which can be controlled by control rods

    International Nuclear Information System (INIS)

    In order to move and position the control rods, the piston-cylinder units of the drives, where there are stepped sections with narrow and wide gaps opposite one another, are connected via a single control pipe each to a valve block. The valves for controlling several units are combined there and are connected to a common pump for the coolant. The valve block with all the control pipes carrying the coolant is accommodated in a sealed metal component in the reactor pressure vessel, so there can be no leaks to the outside. The invention is particularly suitable for reactors of small power based on the boiling water principle. (orig./HP)

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

    International Nuclear Information System (INIS)

    . 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

  19. Pebble bed reactor with one-zone core

    International Nuclear Information System (INIS)

    The claim deals with measures to differentiate the flow rate and to remove spherical fuel elements in the core of a pebble bed reactor. Hence the vertical rate of the fuel elements in the border region is for example twice as much as in the centre. A central funnel-shaped outlet on the floor of the core container over which a conical body is placed with its peak pointing upwards, or also the forming of several outlets can be used to adjust to a certain exit rate for the fuel elements. The main target of the invention is a radially extensively constant coolant outlet temperature at the outlet of the core which determines the effectiveness of the connected heat exchanger and thus contributes to economy. (orig./PW)

  20. Evaluation of reactor structural function during core drop accident

    International Nuclear Information System (INIS)

    Background: A core drop accident causes the core barrel and supporting structure to drop and impact on the bottom of the reactor vessel. Purpose: To ensure the scram functionality is maintained. Methods: The stress-strain curves applicable to the material are approximated by Ludwik's expression. Considering the core deadweight, buoyancy force and heat expansion, material train ratios for both cold and hot conditions are calculated. Results: Vessel impact load (8294482N in cold condition and 6064537N in hot condition) is determined, which remains within the vessel design specification limits. The calculated drop length (47.44 mm in cold condition and 27.63 mm in hot condition) is less than the fuel pin full-diameter engagement length, so the fuel assembly top nozzle will remain engaged during a core drop. Finally the compressive assemblies are evaluated, which would not buckle due to the core drop accident loads. Conclusions: A core drop will not affect the scram function. (authors)

  1. 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

  2. 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

  3. Oscillating annular liquid membranes

    International Nuclear Information System (INIS)

    The response of annular liquid membranes (e.g. used as protection systems in laser fusion reactors) to sinusoidal mass flow rate fluctuations at the nozzle exit is analyzed as a function of the amplitude and frequency of the axial velocity fluctuations at the nozzle exit and thermodynamic compression of the gas enclosed by the membrane. The pressure of the gases enclosed by the annular membrane and the axial distance at which the annular membrane merges on the symmetry axis are periodic functions of time which have the same period as that of the mass flow rate fluctuations at the nozzle exit. They are also nearly sinusoidal functions of time for small amplitudes of the mass flow rate fluctuations at the nozzle exit, and exhibit delay and lag times with respect to the sinusoidal axial velocity fluctuations at the nozzle exit. The delay and the lag times are functions of the amplitude and frequency of the mass flow rate fluctuations at the nozzle exit and the polytropic exponent. The amplitudes of both the pressure of the gases enclosed by the annular liquid membrane and the convergence length increase and decrease, resp., as the amplitude and frequency of the mass flow rate fluctuations at the nozzle exit, resp., are increased. They also increase as the polytropic exponent is increased. (orig.)

  4. The VVER Core Physics, Reactor Dosimetry, and Shielding Researches in the LR-0 Reactor

    International Nuclear Information System (INIS)

    Zero-power water reactor LR-0 was created by the Nuclear Research Institute Rez, Nuclear Machinery Skoda, and RRC 'Kurchatov Institute' for researches of neutron parameters of the WWER type power reactors core, fuel storages, and-first of all-for researches in the reactor pressure vessel and internals dosimetry. Suitable geometrical conditions and flexible technical arrangements of the LR-0 facility enabled to carry out the wide experimental program on several full-scale models (mock-ups) of the WWER-440 and WWER-1000 reactors. The tasks of that experiments were the measurements of the neutron (from thermal energy up to 10 MeV) and gamma (from 0.1 up to 10 MeV) spectra and integral parameters of neutron and gamma fields in the different representative points of the mock-ups from the core to the outer pressure vessel surface and the biological shielding (including channel for ex-reactor ionizing chamber), as well as the measurement of spatial power distribution in the core. Fast neutron (energy from 0.5 to 10 MeV) and gamma spectra were measured in several representative points of the mock-ups by the two-parameter spectrometer with the cylindrical stilbene scintillation detectors. Measurements in the thermal and epithermal neutron region were carried out with the activation method using a broad set of activation monitors and with the 3He(n,p) counter. Activation measurements with threshold fast neutron detectors enlarge also the proton-recoil spectra measurements, such activation measurements were carried out especially in cases, when a spectrometer couldn't be put in the necessary position. The core fission rate distribution was obtained by means of gamma-scanning of the fuel pins. The calculations were carried out by different methods (deterministic and Monte Carlo). Experimental and calculation results in the core, internals, pressure vessel and shielding are reviewed and compared. (Authors)

  5. 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

  6. 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)

  7. 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

  8. Dynamic behavior of the Fast Reactor cores: the Symphony program

    International Nuclear Information System (INIS)

    A fast reactor core is schematically constituted of Fuel Assemblies and Neutronic Shields, immersed in the primary coolant (sodium) which circulates inside the assemblies. Two main physical phenomena have a strong influence on the dynamic behavior of this system: the impacts between the beams and the interactions with the fluid. The impacts between the beams limit the relative displacements. The fluid leads to “inertial effects”, with globally lower vibration frequencies, and “dissipative effects”, with higher damping. Symphony is an important research program on the seismic behaviour of the fast reactor cores, developed from 1993 to 1998 at the CEA Saclay, with both experimental and theoretical parts. The experiments are at a representative scale, with Fuel Assemblies (or FA) and Neutronic Shields (or NS). Test are made “in air” (without fluid) and “in water”, to study the influence of the fluid (the sodium). A numerical model has been built for the interpretation of the tests. The interpretation of the tests is made by using a simple and efficient numerical method, based on the Euler equations for the fluid and homogenization techniques, which yields low computational costs. Impacts between the beams are taken into account also. The gaps between the feet and the grid plate lead to high damping for the beams if the gaps are important. The fluid leads to a strong coupling between the FA and the NS in the whole core, and limits the relative displacement. (author)

  9. Inspection of PUSPATI TRIGA Reactor (RTP) core and control rod

    International Nuclear Information System (INIS)

    The 1 MW PUSPATI TRIGA Reactor (RTP), located at Malaysian Nuclear Agency has been operated since its first criticality on 28 June 1982. The RTP uses uranium zirconium hydride fuel enriched to about 20% of U-235. The RTP has four control rods made up of boron carbide where three are fuel-followers and one is an air-follower. The aluminium cylindrical core can accommodate up to 127 fuel elements while the reflector surrounding it is made from high purity graphite. Since, the reactor power is relatively small, natural convection is used for cooling. Light water is used both as a coolant and as well as a moderator. Visual inspection of the core, fuel and control rods are carried out routinely to ascertain their integrity. An underwater camera and boroscope was used to visually inspect the top grid plate of the core as well as the control rods. No visible defect was detected at the top grid plate however, two of the fuel-follower control rods had blemishes on its surface. This paper will describe the findings of the visual inspection as well as corrective actions taken. (author)

  10. Bat algorithm for the fuel arrangement optimization of reactor core

    International Nuclear Information System (INIS)

    Highlights: • For the first time, bat algorithm has been developed for the core pattern optimization problem. • BANEC results present the strength of BA in gaining semioptimized LPs consuming suitable run time. • Numerical results reveal the acceptable performance and reliability of BA for the LPO problem. - Abstract: In this paper, we develop a novel optimization algorithm, Bat Algorithm (BA), in order to implement in the Loading Pattern Optimization (LPO) of nuclear reactor core. For performing the fuel management optimization, we define a fitness function considering the multiplication factor maximizing and power peaking factor minimizing objectives simultaneously. For this purpose, we prepared a computer program i.e. Bat Algorithm Nodal Expansion Code (BANEC) in order to gain the possible maximum fitness value for the LPO operation. Fuel arrangement optimization using BANEC has been performed for two PWR test cases including KWU and BIBLIS reactors. Numerical results of BANEC confirm that the BA has a great strength to obtain a semioptimized core pattern as respect to considered objective functions during suitable consuming run time. At last, the results show that BA is a very promising algorithm for LPO problems and has the potential to use in other nuclear engineering optimization problems

  11. 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

  12. 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.

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

    International Nuclear Information System (INIS)

    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.

  14. 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

  15. 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

  16. Reactor vessel and core two-phase flow ultrasonic densitometer

    International Nuclear Information System (INIS)

    A local ultrasonic density (LUD) detector has been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) for the Loss-of-Fluid Test (LOFT) reactor vessel and core two-phase flow density measurements. The principle of operating the sensor is the change in propagation time of a torsional ultrasonic wave in a metal transmission line as a function of the density of the surrounding media. A theoretical physics model is presented which represents the total propagation time as a function of the sensor modulus of elasticity and polar moment of inertia. Separate effects tests and two-phase flow tests have been conducted to characterize the detector. Tests show the detector can perform in a 3430C pressurized water reactor environment and measure the average density of the media surrounding the sensor

  17. 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)

  18. 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

  19. 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)

  20. Update on the pulsed reactor IBR-2 and its instruments at Dubna

    International Nuclear Information System (INIS)

    The IBR-2 reactor is the world's most powerful pulsed neutron source. The advantages of the IBR-2 reactor may be most fully revealed at cold neutron wavelengths. This contribution presents data on the parameters of the reactor today and in the near future. A brief description of the reactor instrumentation for physical research is given. (orig.)

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

    International Nuclear Information System (INIS)

    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

  2. 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.

  3. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-04-01

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

  5. RACC-PULSE, Neutron Activation in Fusion Reactor System

    International Nuclear Information System (INIS)

    1 - Description of program or function: CCC-0388/RACC was specifically developed to compute the radioactivity and radioactivity-related parameters (e.g. afterheat, biological hazard potential, etc.) due to neutron activation within Inertial Fusion Energy and Magnetic Fusion energy reactor systems. It can also be utilized to compute the radioactivity in fission, accelerator or any other neutron generating and neutron source system. This new designated RACC-PULSE is based on CCC-0388 and has the capability to model irradiation histories of varying flux levels having varying pulse widths (on times) and dwell periods (off times) and varying maintenance periods. This provides the user with the flexibility of modeling most any complexity of irradiation history beginning with simple steady state operating systems to complex multi-flux level pulse/intermittent operating systems. 2 - Method of solution: The solution method implemented within the RACC-PULSE code is a matrix based method which relies on the evaluation of the Matrix Exponential for the pulse period (on period), dwell period (off time) and post shutdown periods. For the pulsed and dwell periods, the Matrix Exponential was evaluated using the squaring and scaling technique outlined in a review article by Molar and Van Loan entitled Nineteen Dubious Ways to Compute the Exponential of a Matrix. A balanced binary tree method utilized for parameter storage in information systems was employed to evaluate the linear chains constructed for the post shutdown period. The RACC-Pulse code retains the capability of modeling the standard slab, cylinder, sphere and torus geometries in multi-dimensions as well as the point or zero-dimension geometry for Monte Carlo code interfacing. It provides easy interfacing with many of the standard multigroup, multidimensional neutron/photon transport code systems currently employed by the fusion community and implemented on the UNICOS Cray 2 System at NERSC. An auxiliary code is provided

  6. Plant with nuclear reactor, in particular a thermal reactor

    International Nuclear Information System (INIS)

    The reactor core of the plant has tubular and vertically movable control rods moved by a flow of coolant under pressure. Each control rod surrounds a similarly tubular guide rod, stationary relative to the reactor core, leaving an annular slot-like space therebetween. The inside of each guide rod forms a first pressure chamber supplied with the coolant under pressure. The upper end of each control rod is closed and has a vertical shaft that extends into the inside of the guide rod and forms therewith a second annular slot-like space. At least one first restriction is provided in the first annular slot-like space and at least one second restriction is provided in the second annular slot-like space. A second pressure chamber is formed between both restrictions. The coolant supplied to the guide rod thus returns to the pressure vessel surrounding the reactor core through the second annular slot-like space, the second pressure chamber and the first annular slot-like space. Controlling means are provided, with which pressure thrusts can be generated if necessary in the coolant within the first pressure chamber. (author) 5 refs., 10 figs

  7. 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)

  8. Reactor core of a gas-cooled high-temperature reactor

    International Nuclear Information System (INIS)

    In order to increase the outlet temperature of the coolant (helium) leaving the reactor core of a gas-cooled high-temperature nuclear reactor and thereby to improve its thermal efficiency there is proposed to design the geometry of the fuel elements or the fuel element units and/or their main dimensions non-uniformly. Those fuel elements whose geometry causes a larger pressure drop of the coolant gas are to be arranged towards the outlet side of the hot coolant. (GL) 891 GL/GL 892 MKO

  9. 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

  10. Granuloma annulare.

    Science.gov (United States)

    Gupta, Diptesh; Hess, Brian; Bachegowda, Lohith

    2010-01-01

    We present a case of a 77-year-old, diabetic male with a 20-year history of a migratory erythematous, asymptomatic, generalized, nonscaly, and nonitchy rash that started over the dorsum of his left hand. On examination, there were multiple annular erythematous plaques, distributed symmetrically and diffusely over his torso and arms, with central clearing and no scales. A punch biopsy of the skin helped us to arrive at the diagnosis of a generalized granuloma annulare (GA). GA is a benign, self-limiting skin condition of unknown etiology that is often asymptomatic. The cause of this condition is unknown, but it has been associated with diabetes mellitus, infections such as HIV, and malignancies such as lymphoma. These lesions typically start as a ring of flesh-colored papules that slowly progress with central clearing. Lack of symptoms, scaling, or associated vesicles helps to differentiate GA from other skin conditions such as tinea corporis, pityriasis rosea, psoriasis, or erythema annulare centrifugum. Treatment is often not needed as the majority of these lesions are self-resolving within 2 years. Treatment may be pursued for cosmetic reasons. Available options include high-dose steroid creams, PUVA, cryotherapy, or drugs such as niacinamide, infliximab, Dapsone, and topical calcineurin inhibitors. PMID:20209383

  11. 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...

  12. Comparison between a steady-state fusion reactor and an inductively driven pulse reactor

    International Nuclear Information System (INIS)

    In the present report, a comparison is made between tokamak reactors of steady state operation -SSTR- and pulse operation. The former design uses neutral beams as a current driver to realize steady state operation. The latter is inductively operated basic tokamak with burn time of one hour to a half day. This time is determined by dimensions of the central solenoid coil and these dimensions also determine the basic design concept of the pulse tokamak. The dimension includes effect of fatigue due to pulse operation. Performance as a power plant is evaluated with a schematic design of heat transport and power generation system. Heat accumulation in the primary coolant loop is studied in order to make up for a dwell time of a pulse reactor. It is shown that large heat accumulator is necessary to suppress a drop in output during the dwell time. The dwell time has an optimum length with respect to the dwell time. Comparison of fusion plant with other energy source reveals that reduction of the size is essential in order that the fusion is competitive with other sources. (author)

  13. Seismic behaviour of LMFBR reactor cores. The SYMPHONY program

    International Nuclear Information System (INIS)

    As part of a comprehensive program on the seismic behaviour of the LMFBR reactor cores, the SYMPHONY experimental program, performed at the CEA Saclay, is carried out from 1993 up to now. LMFBR reactor cores are composed of fuel assemblies and neutronic shields, immersed in sodium (the primary coolant) or water (for the experimental tests). The main objective of the seismic studies is to evaluate the assembly motions, with consequences on the reactivity and the control rod insertability, and to verify the structural integrity of the assemblies under the impact forces. The experimental program has reached its objectives. Tests have been performed in a satisfying way. Instrumentation allowed to collect displacements, accelerations, and shock forces. All the results constitute a comprehensive base of valuable and reliable data. The interpretation of the tests is based on beam models, taking into account the Fluid Structure Interaction, and the shocks between the assemblies. Theoretical results are in a quite good agreement with the experimental ones. The interpretation of the hexagonal tests in water pointed out very strong coupling between the assemblies and lead to the development of a specific Fluid Structure Interaction, taking into account not only inertial effects, but dissipative effects also. (author)

  14. 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)

  15. A pattern recognition method for nuclear reactor core surveillance

    International Nuclear Information System (INIS)

    An automatic surveillance principle based on a pattern recognition approach is described. The recognition is performed among a set of normal kind of working classes obtained by an unsupervised training. A recognition likehood criterion is used as abnormality decision rule. No hypothesis is made concerning the notion of abnormality. It is defined in opposition with a training data base assumed to be an exhaustive summary of normal behaviours. A core vibration surveillance method for a pressurized water reactor is developed. One of the aims of this study is the early detection of failures. The method allows on line applications and is able to give quick surveillance decisions (about one minute). The sensors are the four ex-core neutron ionization chambers. The extracted features are the characteristics of the neutron noise power spectrum densities resonances. The system is very flexible and adaptable to the statistical properties of the data. The abnormality decision rule severity is adjustable by a sensitivity parameter. The algorithms are heuristic but need only the definition of a small number of parameters. These parameters are optimized either by automatic procedures or by explicit experimental methods. The system is applied on data from a French 900 MW nuclear power reactor. A false alarm probability estimation is carried out by letting the surveillance process work on unknown normal data. Abnormality detection probability is calculated with several kinds of simulated abnormalities by adding a parasit frequency or a white noise. The experimental results obtained by the surveillance system validate the retained principle

  16. 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

  17. 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)

  18. Determination of temperature distributions in fast reactor core coolants

    International Nuclear Information System (INIS)

    An analytical method of determination of a temperature distribution in the coolant medium in a fuel assembly of a liquid-metal-fast-breeder-reactor (LMFBR) is presented. The temperature field obtained is applied for a constant velocity (slug flow) fluid flowing, parallel to the fuel pins of a square and hexagonal array assembly. The coolant subchannels contain irregular boundaries. The geometry of the channel due to the rod adjacent to the wall (edge rod) differs from the geometry of the other channels. The governing energy equation is solved analytically, assuming series solutions for the Poisson and diffusion equations, and the total solution is superposed by the two. The boundary conditions are specified by symmetry considerations, assembly wall insulation and a continuity of the temperature field and heat fluxes. The initial condition is arbitrary. The method satisfies the boundary conditions on the irregular boundaries and the initial condition by a least squares technique. Computed results are presented for various geometrical forms, with ratio of rod pitch-to-diameter typical for LMFBR cores. These results are applicable for various fast-reactors, and thus the influence of the transient solution (which solves the diffusion equation) on the total depends on the core parameters. (author)

  19. Integral experiment on effects of large opening in fusion reactor blanket on tritium breeding using annular geometry

    International Nuclear Information System (INIS)

    An experiment involving a simulated blanket with an opening has been performed using the line source and annular blanket system developed under the JAERI/USDOE collaborative programme in order to examine the effects of the opening on neutronics parameters such as the tritium-breeding ratio. The annular test assembly was rectangular in shape and consisted of a lithium oxide blanket covered with graphite and SS304 which simulated the graphite armour plate and first wall in a fusion device. A large opening (376mm x 425.5mm) was made in the middle of the test blanket. This opening simulated a neutral beam injector.Tritium production rates and reaction rates were measured inside the blanket. Neutron spectra and reaction rates were also measured on the surfaces of both sides and without the opening of the inner cavity. The opening decreased the number of low energy neutrons contained in the cavity and especially decreased 6Li tritium production by 10% inside the blanket at the opposite side of the opening. The Monte Carlo code GMVP using the JENDL-3 nuclear data library predicted the measured nuclear parameters in the test blankets, such as the tritium production rate, to within 10% accuracy. (orig.)

  20. Quality assurance of emergency core cooling system in nuclear reactors

    International Nuclear Information System (INIS)

    Ensuring integrity of nuclear fuel clad is most important from radiation safety point of view. The paper provides introduction to experimental and theoretical methods for evaluation of rewetting velocity. Quality Assurance (QA) checks on Emergency Core Cooling System (ECCS) in Nuclear Reactors are very important to ensure accurate coolant flow introduction and minimum radiation hazard during Loss of Coolant Accident (LOCA). In depth knowledge of acceptable rate of temperature rise of fuel subsequent to LOCA and having fully reliable method to ensure that the same will not exceed the set limit is testimony of safe reactor operation. Spread of radioactive contamination and resultant radiation exposure from above contamination in nuclear reactors depends heavily on size and shape of split or rupture in clad. Suggests that a plant operating with 0.125 percent pin hole in fuel clad defects showed in general, upto five-fold increase in contamination level and resultant whole body radiation exposure rates in some areas of the plant when compared to a sister plant with high integrity fuel. The checks on ECCS will protect environment and public from radiation exposure to remarkable extent. (author)

  1. 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)

  2. Modeling of reflood of severely damaged reactor core

    International Nuclear Information System (INIS)

    The TMI-2 accident and recently Fukushima accident demonstrated that the nuclear safety philosophy has to cover accident sequences involving massive core melt in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although severe accidents are low likelihood and might be caused only by multiple failures, accident management is implemented for controlling their course and mitigating their consequences. In case of severe accident, the fuel rods may be severely damaged and oxidized. Finally, they collapse and form a debris bed on core support plate. Removal of decay heat from a damaged core is a challenging issue because of the difficulty for water to penetrate inside a porous medium. The reflooding (injection of water into core) may be applied only if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ from original rod bundle geometry and will resemble to the severe damaged core observed in TMI-2. The higher temperatures and smaller hydraulic diameters in a porous medium make the coolability more difficult than for intact fuel rods under typical loss of coolant accident conditions. The modeling of this kind of hydraulic and heat transfer is a one of key objectives of this. At IRSN, part of the studies is realized using an European thermo-hydraulic computer code for severe accident analysis ICARE-CATHARE. The objective of this thesis is to develop a 3D reflood model (implemented into ICARE-CATHARE) that is able to treat different configurations of degraded core in a case of severe accident. The proposed model is characterized by treating of non-equilibrium thermal between the solid, liquid and gas phase. It includes also two momentum balance equations. The model is based on a previously developed model but is improved in order to take into account intense boiling regimes (in particular

  3. Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

    International Nuclear Information System (INIS)

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO2 as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO2 is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year

  4. Study of Homogeneous Core Assemblies Using Pulsed Neutron Sources

    International Nuclear Information System (INIS)

    The pulsed neutron technique was there applied to the study of light-water-moderated homogeneous core assemblies, using fissile materials in solution form it is easy to achieve large variations in geometrical buckling and moderation ratio. In the initial series of experiments, the fuel consisted of U235 or U238 in the form of uranyl nitrate. The solution concentrations used varied from 44 to 326 g of uranium per litre. Adoption of a coherent series of cross-sections made it possible to deduce, from the variations in the prompt neutron decay constant as a function of geometrical buckling, data on the non-leakage probability and the slowing-down area. We adopted effective cross-sections calculated on the basis of assimilating light water to a secondary differential thermalizer. Interesting comparisons are made possible by the use of two fissile materials with markedly different η and resonance capture values. In a second series of experiments, devoted mainly to safety measures, we were able to deduce the maximum permissible concentrations in various containers at processing plants by measuring the prompt neutron decay constants in weak plutonium nitrate solutions. (author)

  5. 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

  6. 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)

  7. 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)

  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. New results from pulse tests in the CABRI reactor

    International Nuclear Information System (INIS)

    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

  10. 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.

  11. Forming the BN-600 reactor core model using GEFEST code fuel archive for SYNTES code

    International Nuclear Information System (INIS)

    The article describes the first stage of forming SYNTES code simulation model of the BN-600 reactor core, i.e. organization of transfer of the existing model of the core from GEFEST code fuel archive to a temporal database

  12. 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

  13. 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)

  14. Intelligent system for conceptural design of new reactor cores

    International Nuclear Information System (INIS)

    The software system IRDS has been developed at Japan Atomic Energy Research Institute to support the conceptual design of a new type of reactor core in the fields of neutronics, thermohydraulics, and fuel behavior. IRDS involves various analysis codes, database, and man-machine interfaces that efficiently support a whole design process on a computer. The main purpose of conceptual design is to decide an optimal set of basic design parameters. Designers usually carry out many parametric survey calculations and search a design window (DW), which is a feasible parameter range satisfying design criteria and goals. An automatic DW search function is installed to support such works. The man-machine interface based on menu windows will enable nonspecialists to use various analysis codes easily

  15. Detectors for hydrodynamical processes in the reactor core

    International Nuclear Information System (INIS)

    The method described in this report is based on noise analytical measurements of electrical conductivity fluctuations of the cooling water stream in the reactor core. The conductivity fluctuations have their origin in steam and air bubbles and in cooling water mixing effects in regard to temperature and ionisation by gamma and neutron fields. The fluctuations are transformed into voltage signals by two electrodes in direction of the cooling water stream and then crosscorrelated. From the known distance of the two electrodes and the shift of the crosscorrelation function one can compute the velocity of the cooling medium and the bubbles. Void fractions were also determined with this detection device in out of pile experiments. (author)

  16. Analysis of core calculation schemes for advanced water reactors

    International Nuclear Information System (INIS)

    This research thesis addresses the analysis of the core control of sub-moderated water reactors with plutonium fuel and varying spectrum. Firstly, a calculation scheme is defined, based on transport theory for the three existing assembly configurations. It is based on the efficiency analysis of the control cluster and of the flow sheet shape in the assembly. Secondly, studies of the assembly with control cluster and within a theory of diffusion with homogenization or detailed assembly representation are performed by taking the environment into account in order to assess errors. Thirdly, due to the presence of a very efficient absorbent in control clusters, a deeper physical analysis requires the study of the flow gradient existing at the interface between assemblies. A parameter is defined to assess this gradient, and theoretically calculated by using finite elements. Developed software is validated

  17. 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.)

  18. 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

  19. CEDM Controller for a Linear Pulse Motor by using Pulse Width Modulation Method in Integral Reactor

    International Nuclear Information System (INIS)

    Integral Reactor SMART is under development at KAERI. The design characteristics of SMART are radically different from those employed in currently operating loop type PWR in Korea. The reliability and accuracy of Control Rod Drive Mechanism are very important to the reactor safety and the design of the Plant Protection System. The SMART CEDM designed for fine-step movement consists of a linear pulse motor, reed switch type sensor with top and bottom limit switches which also act as Control Element Assembly(CEA) Position indicator, The linear pulse motor is a four phase synchronous DC electric machine with inner stator and output stator in coolant medium inside a strong housing. The objective of this paper is to introduce and to explain the CEDM controller CEDM Controller is being developed with a new design concept and digital technology to reduce the Operating Error and improve the systems' reliability and availability. And Switched Mode Power Supply is also being developed with digital hardware technology. This paper involves the test details and result

  20. 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

  1. 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

  2. Core reactivity estimation in space reactors using recurrent dynamic networks

    International Nuclear Information System (INIS)

    A recurrent Multi Layer Perceptron (MLP) network topology is used in the identification of nonlinear dynamic systems from only the input/output measurements. This effort is part of a research program devoted in developing real-time diagnostics and predictive control techniques for large-scale complex nonlinear dynamic systems. The identification is performed in the discrete time domain, with the learning algorithm being a modified form of the Back Propagation (BP) rule. The Recurrent Dynamic Network (RDN) developed is applied for the total core reactivity prediction of a spacecraft reactor from only neutronic power level measurements. Results indicate that the RDN can reproduce the nonlinear response of the reactor while keeping the number of nodes roughly equal to the relative order of the system. As accuracy requirements are increased, the number of required nodes also increases, however, the order of the RDN necessary to obtain such results is still in the same order of magnitude as the order of the matematical model of the system. There are a number of issues identified regarding the behavior of the RDN, which at this point are unresolved and require further research. Nevertheless, it is believed that use of the recurrent MLP structure with a variety of different learning algorithms may prove useful in utilizing artifical neural networks (ANNs) for recognition, classification and prediction of dynamic systems

  3. Localization of Vibrating Noise Sources in Nuclear Reactor Cores

    International Nuclear Information System (INIS)

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

  4. Abstracts of reports of the scientific-technical seminar on pulsed research reactors

    International Nuclear Information System (INIS)

    Pulsed graphite reactor IGR was 40 on May 13, 1998. The reactor was constructed to study physical processes occurring in atomic reactors. The IGR creation was initiated by I.V. Kurchatov, the remarkable Soviet physicist and academician. During these years hundreds of atomics units were tested, more than two thousand tests were conducted using IGR reactor. There were also obtained experimental results acknowledged in the world and that will always be of present interest. In relation to this remarkable event, the scientific-technical seminar 'Pulsed Research Reactors'. Application, experimental Researches and results' was held from 9 to 11 of June, 1998

  5. A liquid-metal reactor core demonstration experiment using HT-9

    International Nuclear Information System (INIS)

    The use of the ferritic/martensitic HT-9 alloy as the cladding and duct material for the attainment of the high fuel burnup levels critical to the viability of an economical liquid-metal reactor fuel system. The CDE, a partial core loading of fuel and blanket assemblies in the US Department of Energy's Fast Flux Test Facility, has successfully attained its irradiation exposure goal of 3 yr. Consisting of ten fuel and six blanket assemblies in a heterogeneous core configuration, the CDE has clearly demonstrated the capability of the advanced fuel and blanket designs to attain high burnups and fast fluences. Each CDE fuel assembly consisted of 169 large-diameter fuel pins comprising mixed-oxide annular fuel pellets in sealed HT-9 cladding tubes. Each CDE blanket assembly consisted of 91 large-diameter pins comprising solid depleted uranium dioxide pellets in sealed HT-9 cladding tubes. The maximum-exposure CDE fuel assembly reached a peak pellet burnup of 163,900 MWd/ton metal (M) and a peak fast fluence (E > 0.1 MeV) of 23.3 x 1022 n/cm2. The maximum-exposure CDE blanket assembly reached a peak pellet burnup of 43 100 MWd/ton M and a peak fast fluence (E . 0.1 MeV) of 22.8 x 1022 n/cm2. Lead test fuel assemblies built to CDE specifications continue their successful irradiation and have attained burnups of > 238,000 MWd/ton M with accumulated fast fluences (E > 0.1 MeV) of > 38 x 1022 n/cm2. In-core measurements of HT-9 ducts and withdrawal loads of the assemblies indicate that duct distortion will not be a factor that limits the lifetime of the fuel or blanket assemblies. Comparison of the measured and predicted coolant outlet temperatures from the peak CDE fuel and blanket assemblies indicate the irradiation of the CDE has proceeded as planned. The CDE represents a tremendous success in demonstrating the lifetime capabilities of this advanced oxide system using the HT-9 ferritic alloy for structural materials

  6. Fast reactor core monitoring by analysis of temperature noise

    International Nuclear Information System (INIS)

    The study shows, with the results obtained, how it is possible to approach the problem of diagnosis with a technique based on the use of algorithms for statistical pattern recognition was justifiable. The results presented here, with a view to their use for fast breeder reactor core surveillance, are very encouraging, the most important point being the data representation. For this study, it was difficult to find the most suitable parameters for characterizing the various simulated core states, however, despite this handicap, the classification algorithm provided quite acceptable results. The second point concerns the characterization of a system's evolution. The criterion defined was chosen for adaptation to our algorithm. One acertained that it was possible to characterize evolution on the basis of this criterion as long as the rejected points were not too far from the known learning sets. Under these circumstances, the advantage in characterizing evolution in that the changes in evolution occur when the rejected points have a tendency to agglomerate in a small area of space could be seen. This phenomenon thus makes it possible to forsee whether the creation of a new class is possible. Where the rejected points are far away from the known learning sets, the criterion used proved to be too sensitive and the characterization of evolution was less satisfactory

  7. 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

  8. A Burnup Analysis of PBMR-400MWth Reactor Core

    International Nuclear Information System (INIS)

    The purpose of this study is to analyze the burnup characteristics of 400MWth PBMR using Monte Carlo method. In the world, the deterministic method is widely used to model such that system but it still has a disadvantage which is not flexible in simulating the burnup cycle. Although this method applies some techniques to increase the accuracy of calculation results but it is necessary to model this system by a suitable computer code that can verify and validate the results of the deterministic method. A method which uses a Monte Carlo technique for simulating the burnup cycle was performed. A reactor physics computer code uses in this method is MONTEBURN 2.0 which accurately and efficiently computes the neutronic and material properties of the fuel cycle. MONTEBURN is a fully automated tool that links the MCNP Monte Carlo transport code with a radioactive decay and burnup code ORIGEN. In this model, the calculations are based on a detailed core modeling using MCNP. The fuel pebble is thoroughly modeled by introducing unit cell modeling for the graphite matrix and fuel kernels in the pebble. For the burnup model, a start-up core was studied with considering the movement of pebbles. By shifting down one layer at each discrete time step and inserting fresh fuel from the top, this cyclic calculation is continued until equilibrium burnup cycle is achieved. In this study, the time dependence of multiplication factor keff, the spatial dependence of flux profile, power distribution, burnup, and inventory of isotopes in the start up process are analyzed. The results will provide the basis data of the burnup process and be also utilized as the verified data to validate a compute code for PBMR core analysis which will be developed in near future

  9. Thermal core design of the high performance light water reactor

    International Nuclear Information System (INIS)

    The High Performance Light Water Reactor (HPLWR) is a SCWR concept, operated at an inlet pressure of 25 MPa with a core outlet temperature of 500 deg. C. A thermal core design for this reactor has been worked out by a consortium of Euratom member states within the 6th European Framework Program. Aiming at peak cladding temperatures of less than 630 deg. C, including uncertainties and allowances for operation, the coolant is heated up in three steps with intermediate coolant mixing to eliminate hot streaks. Each fuel assembly is built from 40 fuel pins with 8 mm diameter and a pitch of 9.4mm, housed in a thermally insulated assembly box. Additional moderator water is foreseen in water rods inside each assembly and in gaps between the assembly boxes. With a thermal power of 2300 MW, a net electric power of 1000 MW shall be achieved, resulting in a net efficiency of 43.5%. This concept has been studied with neutronic, thermal-hydraulic and structural analyses to assess its feasibility, which will be summarized in this paper. Coupled neutronic / thermal-hydraulic analyses by Maraczy et al. with the 2 group diffusion code KARATE and the one-dimensional code SPROD are defining an initial distribution of fuel enrichment, the positioning of the control rods, and the use of the burnable Gd absorbers to reach the envisaged power distribution. An equilibrium cycle analysis is showing radial form factors and the discharge burn-up. Different from conventional reactors, the radial power profile is intended to be non-uniform, with the highest power in the first heat up step in the core center and the lowest power in the second superheater step to result in the same peak cladding temperatures in each region. Sub-channel analyses by Himmel et al. performed for different radial power gradients demonstrate the excellent coolant mixing inside assemblies thanks to the wire wrap spacers used in this design. Coolant mixing above and underneath the core has been studied by Wank et al

  10. 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.

  11. 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

  12. 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)

  13. Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

    International Nuclear Information System (INIS)

    Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

  14. McCARD for neutronics design and analysis of research reactor cores

    International Nuclear Information System (INIS)

    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 research reactor, and YALINA subcritical facility. (authors)

  15. 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.

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

    International Nuclear Information System (INIS)

    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: ► We demonstrated the first use of an electro-optic device to trace reactor pulses in real-time. ► We examined the changes in photodiode current for different reactivity insertions. ► Created a linear best fit line from the data set to predict peak pulse powers.

  17. 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.

  18. Neutron measurements at the TRIGA reactor Ljubljana for core inventory verification

    International Nuclear Information System (INIS)

    Safeguards inspections are periodically made in nuclear facilities as a consequence to the Nonproliferation Treaty. The inspection methods are permanently being improved and should not cause serious interference with the reactor operation. Therefore, Core Inventory Verifier (CIVR) is being developed as an indirect quantitative method for verification of the core inventory and detection of the declared operation of research reactors. The CIVR method measures the kinetic behavior of the reactor as well as the neutron flux and its energy distribution at several points inside or outside the core. Measured data taken during inspection is compared with the set of reference data determined previously. The inspection result ''nothing has changed'' indicates that all declared nuclear material really exists in the core. TRIGA reactors are one of the target groups for the CIVR method. The TRIGA reactor Ljubljana was chosen as a reference facility. First results of test series at the TRIGA reactor Ljubljana will be presented in this paper.(author)

  19. 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...

  20. Experimental and analytic investigation of the ITU TRIGA Mark-II reactor core

    International Nuclear Information System (INIS)

    Experimental and analytical studies have been performed to determine the temperature distribution as a function of reactor power in the TRIGA Mark-II reactor at the Istanbul Technical University (ITU). The lumped parameter model with four governing equations was used in the analytical model. Based on the mathematical model, a computer code has been developed for calculating fuel and coolant temperatures in the reactor core. The calculated results for fuel and coolant temperature in the reactor core for different reactor power levels have been compared with the experimental data. Agreements between experiment and results from the computer code are fairly good. (orig.)

  1. 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)

  2. Minimum core configuration with IRT-3M fuel in the VR-1 reactor

    International Nuclear Information System (INIS)

    The present paper shortly describes advances of the RERTR program in the Czech Republic. The minimum core configuration B2 with 9 fuel elements IRT-3M and Beryllium reflector was performed on the training reactor VR-1 Sparrow. The paper presents results of reactor calculations and experimental measurements on the core configuration B2, their evaluation as well as the operation experiences with the Russian fuel assemblies IRT-3M on the reactor VR-1. (author)

  3. Effects of lateral separation of oxidic and metallic core debris on the BWR [Boiling Water Reactor] MK I containment drywell floor

    International Nuclear Information System (INIS)

    In evaluating core debris/concrete interactions for a BWR MK I containment design, it has been common practice to assume that at reactor vessel breach, the core debris is homogeneous and of low viscosity so that it is uniformly distributed radially on the drywell floor. In a recent study performed by the NRC-sponsored Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory, calculations indicate that at reactor vessel bottom head failure, the debris temperature is such that the metallic components (Zr, Fe, Ni, Cr) are completely molten while the oxidic components (UO2, ZrO2, FeO) are completely frozen. Thus, the frozen oxides are expected to remain within the reactor pedestal while the molten metallic species radially separate from the frozen oxidic species, flow through the opening in the reactor pedestal, and spread over the annular region of the drywell floor between the pedestal and the containment shell. This report assesses the impact on calculated debris gas release and the production and release of fission product-laden aerosols for two different cases of debris distribution: uniform distribution, and the laterally separated case of 95% oxides-5% metals inside the pedestal and 5% oxides-95% metals outside the pedestal. The computer codes used in this assessment are CORCON-MOD 2, MARCON 2.1B and VANESA

  4. Numerical analysis of flow distribution at the reactor core inlet of Qinshan phase-II reactor

    International Nuclear Information System (INIS)

    To improve the thermal hydraulic performance of pressurized water reactor (PWR), it is of great importance to obtain reliable flow distribution data at the core inlet. Through computational fluid dynamics (CFD) analysis the flow field in a 1/4 scale model of the 600 MW PWR was worked out numerically. The sensitivity analysis focused on factors such as the lower plenum geometry and 500, 880, 900 m3/h of flow rate for an inlet nozzle, respectively. The results provided a deep understanding of the flow behavior concerning the pressure vessel of the PWR. Numerical results indicated that flow distribution at the core inlet is not sensitive to the inlet nozzle flow rate under two-loop operation mode. Moreover, flow characteristics inside the pressure vessel under single loop operation mode differ significantly from those under two-loop operation mode. A dimensionless flow distribution subfactor of 0.05 is in good agreement with the prototype design. The validity of applying CFD methods in flow distribution for nuclear reactor is verified. The analysis results are useful for the thermal hydraulic design of the PWR. (authors)

  5. 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

  6. 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

  7. ULOF transient behaviour of metal-fuelled fast breeder reactor cores as a function of core size and perturbation methods

    International Nuclear Information System (INIS)

    Highlights: • Metal fuel FBR safety can be assessed by its response to unprotected transients. • Safety during unprotected loss of flow accident (ULOF) is important for FBR cores. • ULOF analyses are carried out as a function of core size and perturbation method. • Smaller metal cores are found to be safer with respect to the ULOF accidents. • 1st order perturbation method gives conservative results in an ULOF accident. - Abstract: The safety behaviour of metal-fuelled fast breeder reactor cores may be assessed by their transient behaviour during anticipated unprotected transients. Out of such transients, unprotected loss of flow accident (ULOFA) has been recognized as an event important for determining reactor safety due to the positive sodium void coefficient of reactivity and the remote possibility of complete power failure as initiator. Reactor safety under ULOFA condition is particularly based on the inherent feedbacks, which is calculated using the removal worths and Doppler constants. As the removal worth is a strong function of reactor size, ULOF analyses are carried out in three different reactor size viz. 120 MWe, 500 MWe and 1000 MWe. The study reveals that smaller metal cores are safer than larger cores with respect to the ULOF accidents in the pre-disassembly phase. The present study also shows that the use of exact perturbation based reactivity worths introduce no significant changes in the safety behaviour of metal fuel reactor compared to that with the use of first order perturbation worths in pre-disassembly phase. The first order approximation is found to be valid as the expansion of materials in the core during ULOFA is small before the core enters the disassembly phase

  8. ULOF transient behaviour of metal-fuelled fast breeder reactor cores as a function of core size and perturbation methods

    Energy Technology Data Exchange (ETDEWEB)

    Riyas, A., E-mail: rias@igcar.gov.in [111B, CDO, Reactor Physics Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Mohanakrishnan, P. [Adjunct Professor, Manipal University, Manipal (India)

    2014-10-15

    Highlights: • Metal fuel FBR safety can be assessed by its response to unprotected transients. • Safety during unprotected loss of flow accident (ULOF) is important for FBR cores. • ULOF analyses are carried out as a function of core size and perturbation method. • Smaller metal cores are found to be safer with respect to the ULOF accidents. • 1st order perturbation method gives conservative results in an ULOF accident. - Abstract: The safety behaviour of metal-fuelled fast breeder reactor cores may be assessed by their transient behaviour during anticipated unprotected transients. Out of such transients, unprotected loss of flow accident (ULOFA) has been recognized as an event important for determining reactor safety due to the positive sodium void coefficient of reactivity and the remote possibility of complete power failure as initiator. Reactor safety under ULOFA condition is particularly based on the inherent feedbacks, which is calculated using the removal worths and Doppler constants. As the removal worth is a strong function of reactor size, ULOF analyses are carried out in three different reactor size viz. 120 MWe, 500 MWe and 1000 MWe. The study reveals that smaller metal cores are safer than larger cores with respect to the ULOF accidents in the pre-disassembly phase. The present study also shows that the use of exact perturbation based reactivity worths introduce no significant changes in the safety behaviour of metal fuel reactor compared to that with the use of first order perturbation worths in pre-disassembly phase. The first order approximation is found to be valid as the expansion of materials in the core during ULOFA is small before the core enters the disassembly phase.

  9. 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

  10. Dimensional changes in elements of the BN-600 reactor core

    International Nuclear Information System (INIS)

    The spread of the data concerning dimensional changes of the components of the BN-600 core is typical for most of the construction materials but cannot be explained exclusively by their nonuniform operational parameters. The spread is caused by nonuniformity of the composition, structure, and properties of the materials within the effective specification requirements. The nonuniformity manifests itself from one lot to another and also within a single lot and, possibly, even in individual finished articles. Embrittlement of the materials during their irradiation is still another important factor which makes it impossible to obtain a burnup in excess of 10% of the heavy atoms. Proper functioning of the fuel assemblies is also limited by the stress caused by irradiation during operation as a consequence of the combined deformation of a fuel-element bundle and the hexagonal box. This holds particularly for fuel assemblies having boxes of weakly swelling ferritic-martensitic steel and fuel cladding tubes made from austenitic steel. It seems that this is the main reason for the fact that proper functioning of standard fuel assemblies with fuel cladding tubes made from cold-worked steel and boxes which is restricted to a burnup of 11-13% of the heavy atoms. When the combined deformation occurs, 05Kh12N2M is the material to be preferred for the boxes. Increasing the service life of the core components of fast reactors must involve work on improving the materials to obtain well-reproducible properties as these determine the initial state of the steel

  11. Control Rod Reactivity Measurements in the Aagesta Reactor with the Pulsed Neutron Method

    International Nuclear Information System (INIS)

    An extensive series of control rod measurements was made in the Aagesta reactor during the low power experimental period following the first criticality. This report describes the part of these investigations made with the pulsed neutron method, comprising nearly 300 measurements. The main objective was the determination of control rod reactivity worths for different rods and groups of rods, but some supplementary measurements were also made, e.g. a determination of the prompt neutron decay constant for the delayed critical condition and four different cores. The cores consisted of 20, 32, 68, and 140 fuel elements respectively, and measurements were made at room temperature and with the moderator level close to critical for each core, and for the 140-element core also with full moderator height and at the temperatures 140 deg C and 215 deg C. Both fully and partly inserted control rod groups were investigated. The measurements at critical water level give directly the control rod reactivity worths, whereas those with full water height give the shut-down reactivity. A comparison was made between measured reactivity worths for a number of rod groups and those calculated with the HETERO code. The prompt neutron decay constant at delayed criticality α0=β/l, for the full core at 215 deg C was found to be 9.60 ± 0.30/sec, corresponding to l = 0.76 ± 0.02 msec. The shut-down reactivity with 16 coarse control rods in pos. A-D 22, 40-04, 44, 26 is -5% at 25 deg C and -13% at 215 deg C. The relative error is usually around 8% in the reactivity worths, originating mainly from the higher harmonics content in the measured curves

  12. 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.

  13. 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

  14. 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

  15. 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)

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

    International Nuclear Information System (INIS)

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

  17. Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

    International Nuclear Information System (INIS)

    Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

  18. Fault current limiter-predominantly resistive behavior of a BSCCO shielded-core reactor

    International Nuclear Information System (INIS)

    Tests were conducted to determine the electrical and magnetic characteristics of a superconductor shielded core reactor (SSCR). The results show that a closed-core SSCR is predominantly a resistive device and an open-core SSCR is a hybrid resistive/inductive device. The open-core SSCR appears to dissipate less than the closed-core SSCR. However, the impedance of the open-core SSCR is less than that of the closed-core SSCR. Magnetic and thermal diffusion are believed to be the mechanism that facilitates the penetration of the superconductor tube under fault conditions

  19. Aspects of unconventional cores for large sodium cooled power reactors; evaluation of a literature survey

    International Nuclear Information System (INIS)

    The report gives an overview of a literature study on the application of unconventional cores for sodium cooled fast reactors. Different types of unconventional cores (heterogeneous cores, pancake cores, moderated cores and others) are compared with conventional cores, which are characterized by a cylindrical geometry with two or three fissile zones surrounded by an axial and a radial blanket. The main parameters of interest in this comparison are the neutronic parameters sodium void and Doppler effect, the breeding properties and the steel damage. Consequences for the core safety and the overall plant design are also mentioned

  20. A nuclear analytical model for uranium zirconium hydride reactor core

    International Nuclear Information System (INIS)

    The nuclear analytical model and codes for the uranium zirconium hydride reactor are outlined. The criticality and control rods effeciency of abroad TRIGA reactor are obtained using this model and codes. The results are satisfactory

  1. Influence Of The Gas Multipurpose Reactor Core Conversion From Oxide To Silicide On The GAMMA Density

    International Nuclear Information System (INIS)

    In order to prepare the reactor core conversion from oxide to silicide, analysis of the gamma heat generation in the fuel plate and its influence on the gamma density in the reactor core using the GAMSET computer code have been done. The heat generation was evaluated for oxide (U3O8-Al) and silicide (U3Si2-Al) plate for different uranium loading. The calculation result shows that the heat generation in the silicide fuel plate contains 400 gram of U-235 per fuel element increase of 10.64% related to the normal oxide plate. This means that the gamma density in the reactor core will consequently decrease. Regarding this result, it can be concluded that the core conversion from oxide to silicide fuel with higher uranium loading will be followed by the heat generation increases in the fuel plate and the gamma density decreases in the reactor core

  2. Identification of a nuclear reactor core (VVER) using recurrent neural networks

    International Nuclear Information System (INIS)

    Recurrent neural networks (RNNs) in identification of complex nonlinear plants like nuclear reactor core, have difficulty in learning long-term dynamics. Therefore, in most papers in this area, the reactor core is used to identify just the short-term dynamics. In this paper we used a multi-NARX (nonlinear autoregressive with exogenous inputs) structure, including neural networks with different time steps and a heuristic compound learning method, consisting of off-line and on-line batch learnings. This multi-NARX was trained by an accurate 3-dimensional core calculation code. Network responses show that this procedure solves the difficulty in identification of complex nonlinear dynamic MIMO (multi-input multi-output) plants like nuclear reactor core, and can be used in fast prediction of nuclear reactor core dynamics behavior

  3. Dynamics model of the IBR-2M pulsed reactor for analysis of fast transition processes

    Science.gov (United States)

    Pepelyshev, Yu. N.; Popov, A. K.; Sumkhuu, D.; Sangaa, D.

    2015-05-01

    A nonlinear model of the IBR-2M pulsed reactor dynamics relating values of variables at discreet instants of time (when power pulses appear) is developed on the basis of the MATLAB program system. The tests of the model by simulating calculated processes in the IBR-2M reactor proved the correctness of the model. A tentative estimate of the transfer coefficient for the linear part of the automatic regulator is obtained.

  4. Comparative study of research reactor core utilizing LEU and mixed (LEU and HEU) fuels

    International Nuclear Information System (INIS)

    Two cores of a swimming pool type research reactor, PARR-1, comprising of i) Low Enriched Uranium (LEU) fuel only , ii) LEU fuel mixed with High Enriched Uranium (HEU) fuel, have been analyzed. This study aims to utilize the partially burnt HEU spent fuel elements from the spent fuel rack, with burnup much less than their designed discharge burnup limit, discharged from the reactor core at the time of dismantling the HEU core during the implementation of world wide core conversion project from HEU to LEU in mid 1980's. For this, some reactor physics characteristic parameters , important from reactor operation, control and safety point of view, have been calculated and compared for the above mentioned two cores. These results included, core criticality, excess reactivity, shutdown margin, integrated control rods' worth, flux/power distribution, power peaking factors and the reactivity feed back coefficients for both these cores. Reactor lattice and 3- dimensional core analysis codes, WIMS-D/4 and CITATION were employed for the calculations. For the mix-fueled core, excess reactivity is found to be on higher side, 617 pcm, and accordingly decrease in its shutdown margin is predicted as compared with the values for LEU core. This is due to the effectiveness of less burnt HEU fuel elements in the mix-fueled core. However, other parameters do not show any significant difference for both these cores, due to the location of less burnt HEU fuel element at the core periphery. These results provide the basis for the operation of the research reactor utilizing mixed fuel without affecting its performance from safety and utilization point of view. (author)

  5. CFD analysis of core melt spreading on the reactor cavity floor using ANSYS CFX code

    International Nuclear Information System (INIS)

    Highlights: ► Spreading of core melt on nuclear reactor cavity is calculated using ANSYS CFX. ► Thermal radiation and viscosity of liquid–solid mixture of the melt are modeled. ► The code is validated with FARO and VULCANO spreading experiments. ► Calculation of a full-scale cavity shows the spreading completes within a minute. - Abstract: In the very unlikely event of a severe reactor accident involving core melt and reactor pressure vessel failure, it is important to provide an accident management strategy that would allow the molten core material to cool down, resolidify and bring the core debris to a coolable state for Light Water Reactors (LWRs). One approach to achieve a coolable state is to quench the core melt after its relocation from the reactor pressure vessel into the reactor cavity. This approach typically requires a large cavity floor area on which a large amount of core melt spreads well and forms a shallow melt thickness for small thermal resistance across the melt pool. Spreading of high temperature (∼3000 K), low superheat (∼200 K) core melt over a wide cavity floor has been a key question to the success of the ex-vessel core coolability. A computational model for the melt spreading requires a multiphase treatment of liquid melt, solidified melt, and air. Also solidification and thermal radiation physics should be included. This paper reports the approach and computational model development to simulate core melt spreading on the reactor cavity using ANSYS-CFX code. Solidification and thermal radiation heat transfer were modeled in the code and analyses of the FARO and VULCANO spreading experiments have been carried out to check the validity of the model. The calculation of 100 tons of core melt spreading over the full scale reactor cavity (6 m × 16 m) showed that the melt spread was completed within a minute.

  6. 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...... power, duration, and bandwidth. The same conclusion is found for the peak power and energy of solitons formed beyond the point of maximal compression. Long-pass filtering of these solitons is shown to be a promising route to clean solitonlike output pulses with peak powers of several MW....

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

    International Nuclear Information System (INIS)

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

  8. 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.

  9. Hydrodynamic and heat transfer mathematical model for the pebble-bod high temperature reactor core

    International Nuclear Information System (INIS)

    Mathematical model for the pebble-bed reactor core in a two-dimensional approximation under the coolant natural circulation conditions is considered. The calculation region represents a reactor core with a lateral reflector and unloading channel through which counercurrent gas flow enters. As basic physical model is the porous body model. Thermal hydraulic reactor core calculation for the normal operation mode on the base of the suggested mathematical model is in perfect agreement with the results of the experiment and calculations obtained by other methods. The results of calculation under the coolant natural circulation regime are in qualitative agreement with the analytical estimations obtained by the perturbation method

  10. Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

    Science.gov (United States)

    Weinstein, H.; Lavan, Z.

    1975-01-01

    Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

  11. Design study of eventual core conversion for the research reactor RA

    International Nuclear Information System (INIS)

    Main options are specified for the future status of the 6.5 MW heavy water research reactor RA. Arguments pro and contra restarting the reactor are presented. When considering the option to restart the RA reactor, possibilities to improve its neutronic parameters, such as neutron flux values and irradiation capabilities are discussed, as well as the compliance with the worldwide activities of Reduced Enrichment for Research and Test Reactors (RERTR) program. Possibility of core conversion is examined. Detailed reactor physics design calculations are performed for different fuel types and uranium loading. For different fuel management schemes results are presented for the effective, multiplication factor, power distribution, fuel burnup and consumption. It is shown that, as far as reactor core parameters are considered, conversion to lower enrichment fuel could be easily accomplished. However, conversion to the lower enrichment could only be justified if combined with improvement of some other reactor attributes. (author)

  12. 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

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

    International Nuclear Information System (INIS)

    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

  14. 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.

  15. 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)

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

    International Nuclear Information System (INIS)

    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)

  17. 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

  18. 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.

  19. Simulation of emulsion copolymerization reactions in a continuous pulsed sieve-plate column reactor

    Directory of Open Access Journals (Sweden)

    C. Sayer

    2004-09-01

    Full Text Available This work addressed the viability of using a pulsed sieve-plate column reactor to carry out continuous vinyl acetate/butyl acrylate emulsion copolymerization reactions. A rigorous mathematical model of emulsion copolymerization reactions in a tubular reactor with axial dispersion was used for this purpose. Operational conditions were defined to attain high monomer conversions at the reactor outlet in a relatively short residence time and, at the same time, produce a copolymer with a more homogeneous composition.

  20. Simulation of emulsion copolymerization reactions in a continuous pulsed sieve-plate column reactor

    OpenAIRE

    Sayer C.; Giudici R.

    2004-01-01

    This work addressed the viability of using a pulsed sieve-plate column reactor to carry out continuous vinyl acetate/butyl acrylate emulsion copolymerization reactions. A rigorous mathematical model of emulsion copolymerization reactions in a tubular reactor with axial dispersion was used for this purpose. Operational conditions were defined to attain high monomer conversions at the reactor outlet in a relatively short residence time and, at the same time, produce a copolymer with a more homo...

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

    International Nuclear Information System (INIS)

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

  2. core calculations for ETRR-1 research reactor upgrading

    International Nuclear Information System (INIS)

    nuclear research reactors play an important role in supporting the nuclear energy program for most countries. research reactors are categorized according to the type of fuel, fuel enrichment, type of moderator and reflector, the power of the reactor and its application. most reactors initially operated at low power then an era began to up-rate the power by changing the fuel type, improving the thermal-hydraulic system performance and modifying the control system to comply with the new trends in research reactors and its applications. in this thesis, we carried out static calculation for the egyptian first research reactor ETRR-1 to evaluate its power upgrade possibility. firstly, we carried out cell calculation using WIMSD/4 code to study the variation of the infinite multiplication factor with the variation of fuel enrichment, lattice pitch and adding heavy water by increasing percentage to the ordinary water coolant

  3. 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.

  4. 14 MW INR-TRIGA research reactor core conversion - emergency preparedness challenges

    International Nuclear Information System (INIS)

    INR-Pitesti TRIGA research reactor is basically a pool type reactor with a special design in order to fulfil the requirements for material testing, power reactor fuel and nuclear safety studies. The safety evaluation involved a several design basis accidents. For training purposes, and to exercise our ability to conduct Level-3 PSA studies, a severe accident scenario involving 14-MW INR-TRIGA research reactor has been developed. In this scenario is assumed that a large part of the reactor hall roof or a heavy object escaped from the crane hook is dropped over the 14-MW TRIGA-SSR core, resulting in mechanical damage of the core. It is assumed, also, that no core melting is occurring, but only fuel-cladding rupture being involved for several 25-pins fuel bundles. The paper evaluates the radiological consequences, both early and late consequences, from the emergency preparedness point of view. (author)

  5. 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

  6. LOGOS. HX: a core simulator for high conversion boiling water reactors

    International Nuclear Information System (INIS)

    A three-dimensional physics simulator 'LOGOS. HX' has been developed for the designing analysis of high conversion boiling water reactor (HCBWR) cores. Its functions, calculational methods, and verification results will briefly be discussed. (author)

  7. 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)

  8. 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.

  9. Verification of computer codes for dynamic processes in nuclear reactors against experiments at loop facility of IGR-1 pulse reactor

    International Nuclear Information System (INIS)

    Basic principles of PRISDG and PRISET computer codes structure to analyze dynamic processes in nuclear reactors are presented. The codes were verified against experimental studies of dynamic processes related with flow-stop and power surge. The experimental data were obtained at loop facility of IGR-1 pulse reactor using fuel assemblies of IVV-2M research reactor. Accuracy of the codes is the same as the accuracy achieved in the experiments. Analysis could be performed at PS-2-type personal computers. Running time is not longer than several tens of minutes. (author)

  10. 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.)

  11. Conversion and evaluation of the THOR reactor core to TRIGA fuel elements

    International Nuclear Information System (INIS)

    The THOR reactor is a pool type 1 MW research reactor and has been operated since 1961. The original MTR fuel elements have been gradually replaced by TRIGA fuel elements since 1977 and the conversion completed in 1987. The calculations were performed for various core configurations by using computer codes, WIMS/CITATION. The computing results have been evaluated and compared with the core measurements after the fuel conversion. The analysis results are in good correspondence with the measurements. (author)

  12. Experience of fuel loading formation in the BN-600 reactor core

    International Nuclear Information System (INIS)

    Experience of the fuel loading formation in the BN-600 reactor core is analysed because the safety, reliability and stability of the core operation determine the power unit operation as a whole. For substantiation of the reactor operation safety every fuel loading is planned and realized by means of calculated data base. The GEFEST certified program complex is used as the permanent program for the BN-600 operation calculations

  13. 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)

  14. Complex degradation and ageing phenomena of research reactor core structural materials - experience at 14 MW TRIGA reactor from INR Pitesti

    International Nuclear Information System (INIS)

    The 14 MW TRIGA Research Reactor designed in the early '70s is a relative new research reactor with an operational experience of 30 years. The specific design of reactor core objectives, were to manufacture, build and operate a flexible structure which incorporate previous experience of pool type research reactors. Aluminum alloy 6061 and stainless steel are only materials used for core structural components, which are all easily remotely removable and replaceable by simple hand tools. Properties of those categories of materials were well characterized / known for many other reactors predecessors, and no special criteria or preliminary tests were performed. The mechanical core structure is presented in the paper and designed procedure for periodic testing and inspection is also described. In spite of well known materials properties, the behavior uncertainties of those materials in each reactor case may have special aspects related to design of components, manufacturing technologies, surface finishing and processing, quality control methods, price of specific components, complex conditions in core and vicinity, history of operation, inspection and verification of components, radioactive waste characterization at the end of life of components. Limited assessment of materials properties and suitability for certain application without considering the each individual component load, exposure and life time, may produce limited information on material itself, in fact the issue is the selection criteria for a standard material suitable for a certain application and consequent failure of components. The degradation and ageing are specific to components starting from design, manufacturing technology and expected life when the component should be replaced. The paper presents the practical experience on maintenance requirements specific to TRIGA core components and some techniques of material investigations available at Institute for Nuclear Research Pitesti Post Irradiation

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

    International Nuclear Information System (INIS)

    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) ∼ 1, provides getting quite high burn-up of the core fuel (Bmax ∼ 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)

  16. A comparison of pulsed and steady-state tokamak reactor burn cycles. Pt. 2

    International Nuclear Information System (INIS)

    Pulsed operation of a tokamak reactor imposes cost penalties due to such problems as mechanical fatigue and the need to periodically transfer large amounts of energy to various reactor components. This study focuses on lifetime limitations and capital costs of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas include: fatigue in pulsed poloidal field coils; out-of-plane bending fatigue in toroidal field coils; electric power supply costs; and noninductive current driver costs. A capital cost comparison is made for tokamak reactors operating under the four distinct operating cycles which have been proposed. Since high availability and a low cost of energy will be mandatory for a commercial fusion reactor, we can characterize improvements in physics and technology which will help achieve these goals for different burn cycles. A key conclusion is that steady-state operation is likely to result in the least expensive tokamak reactor (perhaps 20% cheaper than the best pulsed reactor), provided noninductive current drive efficiency can be increased roughly four-fold over present-day experimental results. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R.; Villa, M.; Bock, H.; Abele, H.; Steinhauser, G. [Vienna University of Technology-Atominstitut, Vienna (Austria)

    2011-07-01

    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)

  18. 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)

  19. Main configurations of the reactor core TRIGA Mark III of the ININ, during their operation

    International Nuclear Information System (INIS)

    The Reactor TRIGA Mark III is 43 years old since was put lay critical on November 8 of 1968 for the first time, along their operative life there have been 18 different configurations of the core, being three those more important: the first configuration with elements standard with an enrichment lightly minor than 20% in U-235, the second configuration that deserves out attention is when a mixed core was charged, composite of two different fuels as for their enrichment, the core consisted of 26 fuel elements Flip (of high enrichment approximately of 70%) more 3 control bars with follower of fuel Flip and 59 standard fuel elements, as those mentioned previously, finally is necessary to consider the recent reload of the reactor, with a compound core by fuel elements of low enrichment LEU 30/20. In this work the characteristics more important of the reactor are presented as well as of each one of the described cores. (Author)

  20. Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G. (Nuclear Engineering Division); (2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department)

    2012-04-04

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

  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. Safety and core design of large liquid-metal cooled fast breeder reactors

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  3. A Design Study on Experimental Power Reactor Core Fueled with UO2 CFP

    International Nuclear Information System (INIS)

    A neutronic study on core design of a 300 MWt EPR was performed. In this study the use of 4.8% enriched UO2 coated fuel particle was analyzed. The design was then compared to 5% enriched UO2 pin fueled EPR based on existing PWRs. Both reactors are operated with single batch refueling system with a cycle length of 3 years. The core physics parameters analyzed were : effective multiplication factor in a cycle, flux distributions and cycle burnup. The results of calculation showed that the core effective multiplication factor for reactor with fuel compact can be maintained at 1.2841 at beginning of cycle (BOC) and 1.0060 at end of cycle (EOC). As for the UO2 pin fueled reactor, the effective multiplication factor was 1.1927 at BOC and 1.0514 at EOC. The size of active core for the CFP fueled reactor were 320 cm in height and 320 cm in diameter. As for pin fueled reactor, the height was 200 cm and diameter was 180 cm. The results of calculations showed that neutron flux distribution was quite flat for both types of reactor designs, although the volume of CFP fueled reactor was 5 times as big as the pin fueled reactor

  4. Characteristics and uses of a 250 kW TRIGA reactor

    International Nuclear Information System (INIS)

    The 250 kW TRIGA Mark II reactor is a light water reactor with solid fuel elements in which the zirconium hydride moderator is homogeneously distributed between enriched uranium. Therefore the reactor has the large prompt negative temperature coefficient of reactivity, the fuel also has very high retention of radioactive fission products. The reactor core is a cylindrical configuration with an annular graphite reflector. The experimental facilities include a rotary specimen rack, a central incore radiation thimble, a pneumatic transfer system, and pulsing capability. Other experimental facilities include two radial and two tangential beam tubes, a graphite thermal column, and a graphite thermalizing column. At the steady state power of 250 kW the peak flux is 1x1013n/cm2s in the central test position. In addition, pulsing to about 2000 MW is usually provided giving peak fluxes of about 2x1016n/cm2sec. All TRIGA reactors produce a core-average thermal neutron flux of about 107n.v per watt. Only with very large accelerators could such a high neutron flux be achieved. In order to give an appreciation for the research conducted at research reactors, the types of research could be summarized as follows: thermal neutron scattering, neutron radiography, neutron and nuclear physics, activation analysis, radiochemistry, biology and medicine, and teaching and training. Typical applied research with a 250 kW reactor has been conducted in medicine in biology, archeology, metallurgy and materials science, engineering and criminology. It is well known that research reactors have been used routinely to produce isotopes for industry and medicine. In some instances, reactors are the preferred method of isotope production. We can conclude that the 250 kW TRIGA research reactor is a useful and wide ranging source of radiation for basic and applied research. The operation cost for this instrument is relatively low. (author)

  5. 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

  6. Pebble Bed Reactor: core physics and fuel cycle analysis

    International Nuclear Information System (INIS)

    The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes

  7. 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)

  8. 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

  9. 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

  10. 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

  11. 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.)

  12. 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

  13. 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)

  14. 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

  15. 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.

  16. 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

  17. Effective height of the core of the nuclear research reactor Dalat

    International Nuclear Information System (INIS)

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

  18. Polarization-maintaining fiber pulse compressor by birefringent hollow-core photonic bandgap fiber.

    Science.gov (United States)

    Shirakawa, Akira; Tanisho, Motoyuki; Ueda, Ken-Ichi

    2006-12-11

    Structural birefringent properties of a hollow-core photonic-bandgap fiber were carefully investigated and applied to all-fiber chirped-pulse amplification as a compressor. The group birefringence of as high as 6.9x10(-4) and the dispersion splitting by as large as 149 ps/nm/km between the two principal polarization modes were observed at 1557 nm. By launching the amplifier output to one of the polarization modes a 17-dB polarization extinction ratio was obtained without any pulse degradation originating from polarization-mode dispersion. A hybrid fiber stretcher effectively compensates the peculiar dispersion of the photonic-bandgap fiber and pedestal-free 440-fs pulses with a 1-W average power and 21-nJ pulse energy were obtained. Polarization-maintaining fiber-pigtail output of high-power femtosecond pulses is useful for various applications. PMID:19529631

  19. 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

  20. Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core, OFI (onset of flow instability) and DNB (departure from nucleate boiling)

  1. Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, I.H. [SDTP User Group, Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore 45650, Islamabad (Pakistan)], E-mail: ihbokhari@yahoo.co.uk; Mahmood, T.; Chaudri, K.S. [SDTP User Group, Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore 45650, Islamabad (Pakistan)

    2007-10-15

    Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core, OFI (onset of flow instability) and DNB (departure from nucleate boiling)

  2. 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

  3. 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

  4. Perfection of power release in WWER reactor core

    International Nuclear Information System (INIS)

    Axial processes (AP) during reactor transition regime were studied. An algorithm for supporting a stationary AP value was developed. It was realised at Khmelnitsky NPP, is connected with boron regulation and satisfies all regulatory requirements

  5. Safety critical reactor core central sub-assembly temperature monitoring and control system

    International Nuclear Information System (INIS)

    Full text: This paper describes diversified independent hard wired temperature monitoring and control system for 500 MWe prototype fast breeder reactor (PFBR). The detection of integrity of the subassembly plays a major role, because of high power density and compact core structure of PFBR fuel. To achieve this, a central sub-assembly temperature monitoring and control system (CSA TMCS) is provided for detection of transient over power, blockage of coolant etc. The central sub assembly temperature sensor assembly is provided with four numbers of K-type thermocouples. Out of these, three thermocouples are used for continuous monitoring and the fourth one is a hot stand by. CSATMCS consists of three identical units, each having independent signal conditioner, temperature monitor, set-point input, trip circuit, alarm generator and temperature display. Authorization is provided for changing the set-point. The trip circuit generates SCRAM, which is connected to 2/3- pulse coded safety logic (PCSL). This SCRAM and alarm signals are displayed in control room. Further, they are available through distributed data control system. Self-test diagnostic is available to check the healthiness of the system. The response time of the electronic system shall be 30 ms. The reliability of the temperature monitoring system shall be better than 0.5 spurious failure per year and 10-4 fail danger per demand. The temperature monitoring system complies with EMI/EMC standard IEC 801. The system is classified as safety class-I and it shall be categorized as seismic category-I system. Different design schemes are discussed in detail to achieve high reliability and simple design. The conditioned thermocouple output is also connected to core temperature monitoring system for the calculation of mean outlet temperature rise (δθm) across the core and temperature rise across each subassembly (δθi). This paper also discusses about the verification and validation with reactor subassembly

  6. Cosmic-ray muon radiography for reactor core observation

    International Nuclear Information System (INIS)

    Highlights: • Cosmic-ray muon radiography was evaluated by the HTTR. • Cosmic-ray muon radiography could be performed from the outside of the RPV. • Cosmic-ray muon radiography could be performed from the outside of the CV. • The major structures of the HTTR could be detected by Cosmic-ray muon radiography. • The detectable minimum volume is estimated at approximately 1 (m3) in the HTTR. - Abstract: One of the critical problems that have arisen from the accident at TEPCO’s Fukushima Daiichi nuclear power plant is the removal of fuel debris. For solving this problem, an examination of the internal reactors has been planned to identify the fuel debris. However, the high radiation dose around the reactors has necessitated the development of a remote sensing method that would enable observation of the internal reactors from the outside. In our study, we focused on a nondestructive inspection method by which cosmic-ray muons could be used to observe the internal reactor from outside the reactor pressure vessel (RPV) and containment vessel (CV). We conducted an observation test on the high-temperature engineering test reactor (HTTR) at the Japan Atomic Energy Agency to evaluate the applicability of the method to the internal visualization of a reactor. We also analytically evaluated the resolution of existing muon telescopes to assess their suitability for the HTTR observation, and were able to detect the major structures of the HTTR based on the distribution of the surface densities calculated from the coincidences measured by the telescopes. Our findings suggested that existing muon telescopes could be used for muon observation of the internal reactor from outside the RPV and CV

  7. Cores of production : reactors and radioisotopes in France

    OpenAIRE

    Adamson, Matthew

    2009-01-01

    This paper concerns the technologies used in radioisotope production in the French Atomic Energy Commission (the Commissariat à l’Energie Atomique) between 1946 and 1958. Particular attention is given to the various instruments used for the bombardment of isotopes, including accelerators and reactors, and their relationship with the CEA’s radioisotope preparation laboratories. Ultimately, the vast majority of bombardments took place in research reactors. These versatile machines, and the isot...

  8. Development and application of neutron transport methods and uncertainty analysis for reactor core calculations. Final report

    International Nuclear Information System (INIS)

    This report documents the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations''. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.

  9. CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

    International Nuclear Information System (INIS)

    The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

  10. Research progress and recommendations on reactor pressure vessel integrity under hypothetical core melt down accident

    International Nuclear Information System (INIS)

    Background: It is very important to ensure the integrity of the reactor pressure vessel under core melt down accident. The high-temperature creep failure is the main failure mode of the reactor pressure vessel under core melt down accident. Purpose: This paper is to present an overview of research status and progress on high-temperature creep behavior of reactor pressure vessel considering the hypothetical core melt down scenario. Methods: Emphasis is placed on accomplished achievements in creep tests, scale model experiments and numerical simulation, and the domestic newly research productions on high-temperature creep behavior of reactor pressure vessel structure integrity. Conclusions: This paper also discusses the limitations of existing researches and indicates future research directions, such as multi-axis tensile tests, analysis of three-dimensional coupling temperature field, scaled model tests, and so on. (authors)

  11. 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)

  12. Analysis of mechanical vibrations in the reactor core using noise techniques

    International Nuclear Information System (INIS)

    Using reactor noise techniques, a vibrating in-core instrument tube in the boiling water reactor could be detected. The normalized auto power spectral density (NPSD) of the current fluctuations of a min in-core chamber, fixed within the instrument tube, has been analysed before and after repair. There was a strong evidence of a peak at the resonance frequency of the vibration at 2.6 Hz in the NPSD. It vanished promptly after bypass-streaming at the power core plate, being the physical origin of the vibrating instrument tube, was rearranged. (author)

  13. In-core fuel management optimization of a Very High Temperature pebble-bed Reactor

    International Nuclear Information System (INIS)

    A new calculation procedure was developed to reduce the power peak in the core of a Very High Temperature pebble-bed Reactor. The procedure consists in several coupled computational codes, which are used iteratively until convergence is reached. This procedure combines the fuel depletion and the neutronic behavior of the fuel in the reactor core, modeling once-through-then-out cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times, obtaining the asymptotic fuel-loading pattern directly, without any intermediate loading pattern. (Author)

  14. Design factors affecting dynamic behaviour of fast reactor cores. UK review paper

    International Nuclear Information System (INIS)

    This paper summarises the consideration that has been given in the UK to the following factors that affect the dynamic behaviour of fast reactor cores: fuel design - Pu/u homogeneity, fuel expansion, fuel-clad gaps, uranium fraction. Structural response - CR supports, diagrid, sub-assembly bowing sodium expansion coefficients - low void cores including heterogenous cores. Calculational methods and models are outlined and some experimental results are discussed. (author)

  15. Development and validation of a fast reactor core burnup code - FARCOB

    Energy Technology Data Exchange (ETDEWEB)

    Mohanakrishnan, P. [Indira Gandhi Centre for the Atomic Research, CDO, Reactor Physics Division, Kalpakkam, TN 603 102 (India)], E-mail: mohana@igcar.gov.in

    2008-02-15

    A large fast breeder reactor requires the accurate estimation of power produced in different parts of the reactor core and blanket during any operating condition for a safe and economic operation through out reactor life time. A fast reactor core simulation code FARCOB based on multigroup diffusion theory has been developed in IGCAR for core simulation of PFBR reactor under construction. FARCOB uses centre mesh differencing scheme with triangular meshes in the X-Y plane. Steady state solution results match exactly with those of other reputed codes DIF3D and VENTURE for SNR-300 benchmarks. For burnup simulation, core is divided into radial and axial burnup zones and burnup equations are solved at constant power. Burnable fuel and blanket number densities are found and stored for each mesh, so that the user can alter burnup zones and core geometry after a burnup step. For validation, results of FARCOB has been compared with results of other institutes in two burnup benchmarks (ANL 1000 MWe benchmark and BN-600 hybrid core benchmark). It is found that FARCOB results match well with those of the other institutes.

  16. In core gamma dosimetry using thermoluminescence detectors (TLDs) in research reactor

    International Nuclear Information System (INIS)

    Since gamma flux co-exists with the neutrons in the reactor core of a research reactor, it becomes difficult to measure exclusively gamma dose rate. Whereas it is quite important to know the gamma dose rates while performing controlled experiments in a research reactor. With this urge experiments have been performed to measure gamma dose rate at central vertical port (CVP) of the University of Florida Training Reactor (UFTR) using thermoluminescence detectors (TLDs). It is demonstrated that among the commercially available TLDs, LiF/sub 2/ can be used to determine the in-core gamma dose rate at low rector power levels (a few hundred watts). The gamma dose rate depending upon reactor power at CVP of UFTR has been determined to be 0.66+- 0.09 Ghy/sup -1/W/sup -1/. Extrapolation of gamma dose higher power is discussed. (author)

  17. Overview on material relocation phenomena in liquid metal fast reactors as consequence of core disruption

    International Nuclear Information System (INIS)

    During core destruction of liquid metal cooled reactors materials relocation and their reactivity feedback effects determine the course of the resulting nuclear power excursion. The objective of the analysis of these relocation phenomena is twofold: 1. Evaluation whether the resulting power excursion might lead to a mechanical loading of the reactor vessel and roof structures which exceed load limits being able to violate the integrity of these structures; 2. Determination of resulting core materials distribution within the reactor vessel after achievement of permanent subcriticality for analysis of their long term in-place coolability. Dependinh on the core states established during the different phases of core destruction, different phenomena determine the materials relocation and their reactivity feedbacks. A qualitative overview of the most important material relocation phenomena and their impact on reactivity feedbacks is given. Specific aspects of fuel relocation phenomena in the initiation phase and in the late transition phase are discussed

  18. Development of cutting technique of reactor core internals by CO laser

    International Nuclear Information System (INIS)

    The CO laser is superior in the absorption characteristic to materials to the CO2 laser due to its shorter wavelength. In consideration of this characteristic Nuclear Power Engineering Corporation is studying this applicability sponsored by the Ministry of International Trade Industry of Japan to cutting of reactor core internals of commercial nuclear power plant. In decommissioning of reactor core internals it is necessary to cut stainless steel plates of 305 mm thick. The authors cut stainless steel plates of up to 310mm thick in air and those of up to 150 mm thick underwater with a 20kW class laser. Further, models simulating key structural elements of PWR core internals were cut and secondary products to clarify the applicability of the CO laser cutting to reactor core internals were evaluated. (author)

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

    International Nuclear Information System (INIS)

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

  20. Thermohydraulic investigation in justification of fast reactor core with liquid metal coolants

    International Nuclear Information System (INIS)

    Thermohydraulic analysis of fast reactor (FR) cores is a component of the complex of interrelated problems on FR parameters justification. These problems concern reactor physics, thermal mechanics, failure theory and other components. Thermohydraulic analysis includes determination of temperature mode of core elements, hydraulic characteristics of fuel assemblies, their nominal and maximal capacitance. The present state of the art of problem-oriented, pilot and applied thermohydraulic studies of FR cores with liquid metal coolants (LMC) is under consideration. The experimental and calculating data on thermal hydraulics of LMC reactors gathered in the Institute for Physics and Power Engineering is analyzed and generalized when comparing different calculating procedures, problem solution techniques. The methods and codes of numerical simulation of thermohydraulic processes in LMC FR core are considered. The problems and aims of further investigations are formulated

  1. 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)

  2. Digital signal processing of pulse counting and MSV measurement for in-core instrumentation

    International Nuclear Information System (INIS)

    A digital signal processing technique for pulse counting and mean-square-voltage (MSV) measurement were developed for start-up range neutron monitor (SRNM) used in BWR plants. The output pulse of fission detector is sampled at 40 MHz. From these sampled data, digital signal processing directly performs pulse counting and MSV measurement. This processing has the following two key features: (1) digital pulse counting technique, allowing rejection of the error counts induced by external noises, and (2) digital over-sampling technique, allowing MSV measurement to cover the ranges of the measurement required for SRNM. A real-time processing prototype apparatus was manufactured and tested at Toshiba Training Reactor (TTR). This apparatus can demonstrate a successful performance of the digital signal processing for SRNM and make it possible to accurately count only the detector output. (author)

  3. Neutronic Design of the First Core of the Replacement Research Reactor

    International Nuclear Information System (INIS)

    The paper describes the general neutronic characteristics of the first core of the replacement research reactor (RRR) for the Australian Nuclear Science and Technology Organisation (ANSTO). A compact core with 16 FA has been designed to fulfil all the very demanding neutronic requirements of the RRR facility. The contractual performance parameters must be verified for the equilibrium core; a very important design effort was carried out in the initial fresh core to have a similar performance. The description covers different aspects of the neutronic design: a detailed nuclear design of U3Si2 first core, the design calculation tools, together with a comparison of the first core performance against the core design criteria and the equilibrium core performance. (author)

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

    International Nuclear Information System (INIS)

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

  5. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

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

  7. Fast reactor 3D core and burnup analysis using VESTA

    Energy Technology Data Exchange (ETDEWEB)

    Luciano, N.; Shamblin, J.; Maldonado, I. [Nuclear Engineering Dept., Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)

  8. Pressurized water reactor core management at Electricite de France

    International Nuclear Information System (INIS)

    The large fraction of nuclear power plants in France has resulted in a number of constraints which have to be considered in core management. The most important are daily load following capacity, frequency control capacity, and shutdown timing optimization. In view of these constraints, optimization of fuel management consists in taking benefits of increased discharged burnup as available. Previous core management schemes were based on extended cycle with gadolinium poisons. Now, due to consumption forecast and operation conditions (outage time), decreased reload fractions (1/4 core) are preferred according to annual cycle length. Also, plutonium recycling is taken into account. (orig.)

  9. 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.

  10. 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.

  11. 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)

  12. 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.

  13. A five years experience of pulse columns extraction cycles for the reprocessing of fast breeder reactor fuels at the Marcoule pilot plant (SAP)

    International Nuclear Information System (INIS)

    The reprocessing of Phenix fast breeder reactor started at the MARCOULE PILOT PLANT in 1977 with the enriched UO2 first core (2.3 tons U) followed by several campaigns of UO2-PuO2 Phenix-core II (6.5 tons U-Pu). After a short description of the Pilot Plant, characteristics of the pulse columns extraction flow-sheets are presented. Pulse columns are used for extraction and scrubbing of uranium and plutonium and for uranium backwashing whilst plutonium stripping and U-Pu partition are carried out in mixer settlers with HAN and in-line electrolytic U IV generation. Performances of pulsed columns including recovery yields and decontamination factors are discussed: they show a good β γ decontamination can be reached with two cycles and partition carried out at the second cycle. (author)

  14. A five years experience of pulse columns extraction cycles for the reprocessing of fast breeder reactor fuels at the Marcoule pilot plant (SAP)

    International Nuclear Information System (INIS)

    The reprocessing of Phenix fast breeder reactor started at the Marcoule Pilot Plant in 1977 with the enriched UO2 first core (2.3 tons U) followed by several campaigns of UO2-PuO2 Phenix-core II (6.5 tons U-Pu). After a short description of the Pilot Plant, characteristics of the pulse columns extraction flow-sheets are presented. Pulse columns are used for extraction and scrubbing of uranium and plutonium and for uranium backwashing whilst plutonium stripping and U-Pu partition are carried out in mixer settlers with HAN and in-line electrolytic U IV generation. Performances of pulsed columns including recovery yields and decontamination factors are discussed: they show a good β γ decontamination can be reached with two cycles and partition carried out at the second cycle

  15. 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

  16. 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.

  17. 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

  18. 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

  19. 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

  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